ML18102A793
| ML18102A793 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 01/21/1997 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18102A792 | List: |
| References | |
| PROC-970121, NUDOCS 9701310343 | |
| Download: ML18102A793 (496) | |
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SGS EAL Technical Basis T.O.C.
Pg. 1 of 2 SALEM ECG TECHNICAL BASIS TABLE OF CONTENTS/SIGNATURE PAGE \O\
January 21, 1997 *
.. .. -* -~--DATE .
SECTION TITLE REV#. . - .. PAGES Table of Contents/Signature Page* 00 2 01121/97 T.O.C.
Introduction and Usage 00 3 01121197
- oo 6. . . . 01121/97 11 Glossary of Acronyms & Abbreviations
- 1.0 Fuel Clad Challenge *
- 00 4 01/21/97 2.0 RCS Challenge _ **oo.
~ . {
2 01/21/97 3.0 Fission Product.Barriers (Table) 3.1 Fuel Clad Barrier 00 20 01121/97 3.2 RCS Barrier 00 16 01/21197 3.3 Containment Barrier 00 24 01121/97 4.0 EC Discretion 00 . 8 . 01/21/97 5.0 Failure to SCRAM . 00 8 01/21/97 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release 00 42 01/21/97 .
6.2 Liquid Effluent Release 00 4 01/21/97 6.3 In - Plant Radiation Occurrences 00 6 01/21/97 6.4 Irradiated Fuel Event 00 10 01/21/97 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 00 11 01/21/97 7.2 *Loss of DC Power Capabilities 00 8 01/21197 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 00 10 01/21197 8.2 Loss of Overhead Annunciators 00 6 01/21197 8.3 Loss of Communications Capability 00 4 01121197 8.4 Control Room Evacuation 00 4 01/21/97 8.5 Technical Specifications 00 2 01/21197 9.0 Hazards - Internal/External 9.1 Security Threats 00 8 01121/97 9.2 Fire 00 6 01/21197 9.3 Explosion 00 4 01/21197 9.4 Toxic/Flammable Gases 00 11 01/21197 9.5 Seismic Event 00 4 01/21197 9.6 High Winds 00 5 01/21197 9.7 Flooding 00 4 01121197 9.8 Turbine FailureNehicle Crash/ 00 6 01121197 Missile Impact 9.9 River Level 00 4 01/21/97 SGS Rev. 00
SGS EAL Technical Basis T.O.C . . .
Pg. 2 of 2 SALEM ECG TECHNICAL BASIS TABLE OF CONTENTS/SIGNATURE PAGE January 21, 1997 ::: ~ I SECTION TITLE REV# PAGES DATE 10.0 Reserved for future use 11.0 Reportable Action Levels (RALs) 11.1 Technical Specifications 00 8 01/21/97 11.2 Design Basis/ Unanalyzed Condition 00 7 01/21/97 11.3 Engineered Safety Features (ESF) 00 4 01/21/97 11.4 Personnel Safety/Overexposure 00 8 01/21/97 llS Environmental 00 3 01/21/97 11.6 After-the-Fact 00 1 01/21/97 11.7 Security/Emergency Response 00 5 01/21/97 Capabilities 11.8 Public Interest 00 3 01/21/97 11.9 Accidental Criticality/ 00 9 01/21/97 Special Nuclear Material I
- Rad Material Shipments - Releases 11.10 Voluntary Notifications 00 2 01/21/97 SGS Rev. 00
SGS EALIRAL Technical Basis*
Section i Pg 1 of3 SALEM
.. ECG, TECHNICAL BASIS_
- ~ h; INTRODUCTION & USAGE I'. <-' *: : *
- 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 d~ubt ex!~~s a.t?out the intent or meaning of an EAL.
1.0 PURPOSE OF THE ECG TECHNICAL BASTS {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;
- analysis of the effect of the-event on fission product barriers
- escalation criteria guidance I .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 2.1 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.l.10.
SGS REV. 00
SGS EALIRAL Technical Basis Section i Pg 2 of3 2.1.1 INITIATING CONDITION (IC): A gen_eric 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 .
2.1.2 EAL NUMBER (EAL#): Each Emergency Action Level (EAL) has been
, assigned.:a unique alpha numeric identifier. Each digit of the EAL# has a specific ~eaning 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 MODE: Refers to the Operational Mode at Salem during which a
-particular IC/EAL is applicable. The Mode 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 affected 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 ~C based EAL examples given in NESP-007.
2.1.1 O
REFERENCES:
Provides a short list of the pertinent documents that are the basis for information included in this technical basis document.
SGS REV. 00
SGS EAL/RAL Technical Basis Section i
. Pg 3 of3 3.0 TECHNICAL BASTS (TB) USAGE , -
NOTE, Event classification should always.be madewith direct reference to the ECG.
If any numbering inconsistency or error should be discovered between the ECG
- *~ and TB, the EAL#s fro_m_the ECG are to be used.
3.1 Classification
USE the Salem 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 NOTE If there 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.
SGS REV. 00
ECG Section ii Pg. 1of6 SALEM.**.
- EVENT-CLASSIFICATION GUIDE Glossary of Acronyms & Abbreviations
..
- Section ii * *
\~
AMG Accident Assessment Advi~ry 'Group (Delaware)..... -.** .....
AB Auxiliary Building . .
AC Alternating Current . * : . .*.. .
.. 1* :* *. 1 ****:.
AFST. Auxiliary Feedwater Storag¢ '.failk AFW. Auxiliary Feedwater* * * * **
ALARA As Low As Reasonably Achievable ARM Area Radiation Monitor AS. Administrative Supervisor_ .....
ASAP As Soon As Possible * * * **
ASM Administrative Support Manager ATY!J Anticipated Transient With~ut T~p BIT Boron Injection .Tank ..
BKGD Background .*
BKR Breaker (electrical circuit)
BNE Bureau of Nuclear Engineering (NJDEPE)
CAS Central Alarm Station CCPM Corrected Counts per Minute CDE Committed Dose Equivalent CEDE Committed Effective Dose Equivalent CET Core Exit Thermocouple CFCU Containment Fan Coil Unit CFR Code of Federal Regulations CFST Critical Safety Function Tree CMl Primary Communicator (CR)
CM2 Secondary Communicator (CR)
CNTMT Containment (Barrier)
CP Control Point CPM Counts Per Minute CR Control Room CRD Control Rod Drive DC Direct Current DDE Deep Dose Equivalent .
DEi Dose Equivalent Iodine DEMA Delaware Emergency Management Agency SGS Rev. 00
ECG Section ii P,g. 2 of.6 DEP D~partment of Environmental Protection (NJ)
DID Dir~t Inward Dial (phone system)
DOE Departmentof Energy DOT Department of Transportation DPCC/DCR - Discharge Prevention, Containment, & Countermeasures/
Discharge Cleanup & Remqval Plan D}>M Decades per Minute ..
l:)PM Disintegrati.ons per Minute PRC~ 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 pmergency Operations Center (NJ & DE)
EOF Emergency Operations Facility EOP Emergency Operating Procedures EPA Emergency Preparedness Advisor EPA Environmental Protection Agency EPIP Emergency Plan Implementing Procedure EPM Emergency Preparedness Manager EPZ Emergency Planning Zone ERDS Emergency Response Data System ERF Emergency Response Facility 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 FHB Fuel Handling Building FPB Fission Product Barrier FRCC Functional Restoration Core Cooling FRCE Functional Restoration Containment Environment FRCI Functional Restoration Coolant Inventory SGS Rev. 00
ECG Section ii Pg. 3of6 FRERP Federal Radiological Emergency Response Plan FRHS Functional Restoration Heat Sink FRSM Functional Restoration Shutdown Margin FRTS Functional Restoration Thermal Shock FTS Federal Telecommunications System (NRC}
GE General Emergency HEPA High Efficiency Particulate. Absorbers HP Health Physics .
HVAC Heating, Ventilation & Air Conditioning HX Heat Exchanger IAW In Accordance With .
IC Initiating Condition ICMF Initial Contact Message Form IDLH Immediately Dangerous to Life and Health
- IR Intermediate Range I/S In Service ISOL Isolation 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 LPZ Low Population Zone MDA Minimum Detectable Amount MEA Minimum Exclusion Area MEES Major Equipment & Electrical Status (Form)
MET Meteorological MIMS Metal Impact Monitoring System MOU Memorandum of Understanding MRO Medical Review Officer MSIV Main Steam Isolation Valve MSL Main Steam Line SGS Rev. 00
ECG Section ii Pg. 4 of.6 NAWAS National Attack Warning Alert System NCO Nuclear Control Operator NDAB Nuclear Department Administration Building (TB2)
NEO Nuclear Equipment Operator NETS Nuclear Emergency Telecommunications System NFE Nuclear Fuels Engineer
&PB Normal Full Power Background NG Noble Gas NJSP New Jersey State Police NOAA . . National Oceanographic and Atmospheric Administration NR Narrow Range NR,C Nuclear Regulatory Commission NSP Nuclear Site Protection NSS Nuclear Shift Supervisor 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 OHA Overhead Annunciators OSB Operational Status Board (Form) osc Operations Support Center PAG Protective Action Guideline PAR Protective Action Recommendation PASS Post Accident Sample System PIM Public Information Manager PMP Pump PORV Power Operated Relief Valve PSIG Pounds per Square Inch Gauge PWST Primary Water Storage Tank PZR Pressurizer RAC Radiological Assessment Coordinator RAD Radiation RAL Reportable Action Level RC Reactor Coolant RCA Radiologically Controlled Area RCAM Repair and Corrective Action Mission RCP Reactor Coolant Pump SGS Rev. 00
ECG Section ii
- . Pg. 5of6 RCS Reactor Coolant System (Barrier) .
RHR Residual Heat Removal RM Recovery Manager RMO Recovery Management Organization RMS Radiation Monitoring System RPS Radiation Protection Supervisor RPS Reactor Protection System RSM Radiological Support Manager RVLIS Reactor Vessel Level.:tnstrumentation System RWST Refueling Water Storage Tank SAE Site Area Emergency SAM Severe Accident Management SAS Secondary Alarm Station (Security)
SAT Satisfactory SBO Station Blackout SCBA Self Contained Breathing Apparatus SCP Security Contingency Procedure SDE Shallow Dose Equivalent SDM Shutdown Margin SIG Steam Generator SGS Salem Generating Station SGTR Steam Generator Tube Rupture SI Safety Injection SJAE .. Steam Jet Air Ejector SNM Special Nuclear Material SNSS Senior Nuclear Shift Supervisor sos Systems Operations Supervisor (Security)
SPDS Safety Parameter Display System SRPT Shift Radiation Protection Technician SSCL Station Status Checklist SSE Safe Shutdown Earthquake SSM Site Support Manager SSNM Strategic Special Nuclear Material.
SUR Startup Rate T-COLD Temperature Cold (Leg)
T-HOT Temperature Hot (Leg)
TAF Top of Active Fuel TDR Technical Document Room TEDE Total Effective Dose Equivalent TPARD Total Protective Action Recommendation Dose T/S Technical Specifications SGS Rev. 00
ECG Section ii Pg. 6 of 6 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 USCff ---
- United States Coast Guard VDC Volts Direct Current VLV Valve WB Whole Body WR Wide Range SGS Rev. 00
SGS EAL/RAL Technical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity
.UNUSUAL EVENT- 1~1.1.a/1.1.1.b
\-6, IC Fuel Clad Degradation EAL Reactor*Coolant Activity >l µCi/gm Dose Equivalent 1-131for>48 Hours Reactor Coolant Activity (Dose Equivalent Iodine) exceeds limits of Technical Specification Figure 3 .4-1 MODE - 1, 2, 3, 4, 5, 6 BASIS Coolant Iodine activity in excess of Technical Specifications is considered to be a precursor of more serious problems. The Technical Specification Iodine limit reflects a degrading or degraded core condition. This level is above any possible short duration Iodine spikes under normal conditions.
Barrier Analysis This event does not reach the threshold for the loss of Fuel Clad Barrier, but does affect that barrier.
ESCALATION CRITERIA This event will be escalated to an Alert when Reactor Coolant activity exceeds 300 µCi/gm Dose Equivalent I-131 per EAL Section 3.1.2.
EAL - 1.1.1.a/1.1. l.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION An Unusual Event is only warranted when actual fuel clad damage is the cause of the elevated coolant sample (as determined by RCS sample analysis confirmation. The Technical Specification limit on RCS Activity of 100/ EµCi/gm was not included in this EAL because it specifically excludes Iodine Activity.
DEVIATION NUMARC requires this EAL to be applicable in all Modes of operation. Since there is no fuel in the Reactor vessel in Mode "Defueled", this EAL is not Applicable.
REFERENCES NUMARC NESP-007, SU4.2 Technical Specification Section 3.4.8 - Unit 1 Technical Specification Section 3.4.9 - Unit 2 EAL - 1.1.1.a/1.1.1.b Rev. 00
. Page 2 of 2
SGS EAL/RALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity UNUSUAL EVENT- 1.1.1.c IC. . .. Fuel Clad Degradation EAL Valid LetdownLine Monitor in Alarm (1R31A or 2R3 l)
MODE - 1, 2, 3, 4, 5, 6 BASIS The letdown monitoring system (1-R31A and 2-R3 l) detects the RCS radiation concentration that is attributable to the fission products that are produced in the reactor and escapes to the coolant. This indicator of elevated coolant activity would be one of the first indicators of a degrading core, and is considered to be a precursor of more serious problems. "Valid" means confirmed by other indications on related or redundant instrumentation.
Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect this barrier.
ESCALATION CRITERIA This event will be escalated to an Alert when RCS activity exceeds 300 µCi/gm Dose Equivalent Iodine 131 per EAL Section 3 .1.2.
DISCUSSION A valid Letdown Line Monitor alarm may indicate that the failed fuel level has reached I% due to an increased number of failed fuel elements or a fuel gap activity release. Sample results are not required prior to classification; however, other radiation monitors should be used to confirm this alarm to prevent inaccurate classification based on an instrument malfunction.
EAL - 1.1.1.c Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION NUMARC requires this EAL to be applicable in all l\fodes of operation. Since there is no fuel in the Reactor vessel in Mode "Defueled", this EAL is *not Applicable.
REFERENCES SGS-UFSAR Section 11.4 NUMARC ~SP-007, SU4.1 '. . ~
OP-AB.RC-'6002(Q
~alem Ul/U2 Radiation Monitoring System Manual EAL - 1.1.1.c Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 2.0 RCS Challenge.
2.1 RCS Leakage UNUSUAL EVENT - 2.1.1.a/2.1.1.b/2.1.1.c IC RCS Leakage EAL Reactor Coolant System Pressure Boundary Leakage_? 10 gpm Reactor Coolant System Unidentified Leakage > 10 gpm Reactor Coolant System Identified Leakage > 25 gpm MODE- 1, 2, 3, 4 BASIS This EAL addresses plant conditions where RCS leakage significantly exceeds limits imposed by Technical Specifications. A leak of such magnitude is consistent with an Unusual Event classification and should be declared immediately. Credit for the Technical Specification Action Statement time in deferring an Emergency Classification should only be given when the leakage exceeds Technical Specification limits but has not yet exceeded the Unusual Event threshold.
These EALs are included as Unusual Events as they may be precursors to a more serious event.
As such, it is considered to be a potential degradation of the level of safety of the plant. The unidentified or pressure boundary threshold value was chosen to be readily observable from the control room using normal indications. The identified leakage threshold value is set at a higher value due to its lesser significance compared to unidentified or pressure boundary leakage. Note that identified leakage includes Primary to Secondary leakage per Technical Specification
- definition.
Barrier Analysis This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier.
EAL - 2.1.1.a/2. l.1.b/2. l. l.c Rev. 00 Page 1 of 2
SGS EALIRAL Technical Basis ESCALATION CRITERIA This event will be escalated to an' Alert or higher classification based on a loss or potential loss of fission product barriers per EAL section 3;0 .
. DISCUSSION Utilizing the leak before break methodology, it is anticipated that there will be indication(s) of minor reactor coolant system boundary leakage prior to a fault escalating to a major leak or a system rupture. Detection of low levels of leakage while pressurized permits monitoring for catastrophic failure or rupture precursors.
DEVIATION None REFERENCES NUMARC NESP-007, SUS EOP-TRIP-1 EOP-LOCA-1 Technical Specifications Definition 1.15.c Technical Specifications 3A6.2 - Unit 1 Technical Specifications 3.4.7.2 - Unit 2 EAL - 2.1.1.a/2. l.1.b/2. l. l.c Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 ~ission Product Barriers 3.1 Fuel Clad Barrier 3.1.1 CRITICAL SAFETY STATUS.
3.1.1.a IC Potential Loss of Fuel Cl.ad Bartjer ~) POINTS EAL
~ CORECOOLINGP~LE PAW MODE_- 1, 2, 3, 4 BASIS Core Cooling PURPLE Path, as verified by EOP-CFST-1, indicates that subcooling has been lost and that some clad damage may occur.
- Barrier Analysis Fuel Clad Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
DEVIATION Salem Generating Station replaced the CFST "Orange Path" color designation with "Purple Path" due to the limitations imposed by the SPDS CRT's color gun configuration.
EAL - 3.1. l.a Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis REFERENCES NUMARC NESP-007, FCI EOP-CFST"'l EOP-TRIP-1 EAL - 3.1.1.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission'Product'Barriers 3.1 Fuel Clad Barrier 3~1.1 CRITICAL SAFETY FUNCTION STATUS'*
3.1.1.b IC Potential Loss of Fuel Clad Barrier= 3 POINTS EAL HEAT SINK RED PATH MODE- 1, 2, 3, 4 BASIS Heat Sink RED Path, as verified by EOP-CFST-1, indicates that Steam Generator dryout could occur.. A loss of Heat Sink poses an extreme challenge to the Fuel Clad. A barrier loss
- , classification should not be made if the Heat Sink RED Path is the result of procedurally required Auxiliary Feedwater flow control.
Barrier Analysis Fuel Clad and RCS Barriers have been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
DEVIATION None EAL - 3.1. l.b Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis REFERENCES NUMARC NESP-007, FCl EOP-CFST-1 EOP-TRIP-1 FRHS~*l ***'
EAL - 3.1.1.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.1- CRITICAL SAFETY FUNCTION STATUS 3.1.1.c IC Loss*ofFuel Clad Barrier= 4 POINTS
- EAL
~ CORECOOLINGREDPAIB MODE - 1, 2, 3, 4 BASIS Core Cooling RED Path, as verified by EOP-CFST-1, is definitive indication that the heat transfer from the fuel to the coolant has degraded leading to a fuel clad heatup, significant superheating and core uncovery.
Barrier Analysis Fuel Clad Barrier has been lost and the Primary Containment Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
DEVIATION None EAL - 3.1.1.c Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, FCl EOP-CFST-1 EOP-TRIP-1 FRCC-"l EAL - 3.1.1.c Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.2 PRIMARY COOLANT IODINE 'CONCENTRATION IC *. _Loss of Fuel Clad Barrier= 4 POINTS EAL Reactor Coolant ~ctivity > 300 µCi/gm Dose Equiv~dent I-131 MODE- 1, 2, 3, 4 BASIS A reactor coolant sample activity of greater than 300 µCi/gm Dose Equivalent Iodine-131 (DEI-131) was determined to indicate significant clad heating or mechanical stress and is indicative of the loss of the fuel clad barrier. This concentration is well above that expected for iodine spikes and corresponds to approximately 2.5% clad damage.
Barrier Analysis Fuel Clad Barrier has been lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION The actual value of 300 µCi/gm Dose Equivalent Iodine-131 (DEI-131) was determined based upon an engineering calculation which is not included with this EAL. This calculation was prepared by the Nuclear Fuels Group and is on file with Emergency Preparedness under file title DSI.6-00XX: "Verification of Emergency Action Levels for Event Classification" date 1/26/95.
DEVIATION None EAL- 3.1.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, FC2 Reg. Guide 1.109, Table E-9
_SGS-USFJ\R, Tablell.1-1 SGS-USFAR, T~ble 11.1~7 OP-AB.RC-0002(Q)
Calculation by Nuclear Fuels Group file title DS 1.6-00:XX "Verification of Emergency Action Levels for Event Classification" date 1/26/95.
EAL - 3.1.2 Rev. 00 Page 2 of 2
SGS EAL/RALTechnicaI Basis 3.0 Fission Product Barrie.rs 3.1 Fuel Clad Barrier J.1.3 CORE EXIT THERMOCOUPLES (CETS) 3.i.3.a IC P.otential Loss of Fuel Clad Barrier= 3 POINTS EAL 5 or more CETs > 700 °F MODE- 1, 2, 3, 4, BASIS The threshold value chosen is from the EOP-CFST-1 Core Cooling Status Tree and indicates a loss of core subcooling which could lead to clad damage.
Barrier Analysis Fuel Clad Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based upon the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the EOP Critical Safety Function Tree (CFST) monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG. Use of Core Exit.
Thermocouple (CET) temperature.to indicate loss of subcooling is equivalent to the CFST Core Cooling status codes.
DEVIATION Salem Generating Station replaced the CFST "Orange Path" color designation with "Purple Path" due to the limitations imposed by the SPDS CRT's color gun configuration.
EAL - 3.1.3.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, FC3 EOP-CFST-1 EOP-TRIP-1 EOP-Setpoint Doc (G.03)
EAL - 3.1.3.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.3 CORE EXIT THERMOCOUPLES (CETS) 3.1.3.b IC Loss of Fuel Clad Barrier= 4 POINTS.
EAL 15 or more CETs > 1200 °F MODE:. 1, 2, 3, 4
- BASIS Five Core Exit tliermocouple (CET) temperatures> 1200 Of indicates a significant superheating of the reactor coolant.
Barrier Analysis Fuel Clad Barrier has been lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The EAL threshold of> 1200 °F is equivalent to CFST Core Cooling RED Path.
DEVIATION None EAL - 3.1.3.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, FC3 EOP-CFST-1 EOP-TRIP-1 EOP-Setpoint Doc (G.04)
EAL - 3.1.3.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.4 - RX VESSEL LEVEL INDICATION SYSTEM (RVLIS) 3.1.4.a IC Potential Loss of Fuel Clad Barrier= 3 POINTS EAL RVLIS Full Range< 39% -
MODE- 1, 2, 3, 4 BASIS -
The threshold value ofRVLIS Full Range< 39% is chosen from the EOP-CFST-1 Core Cooling Status Tree. This value approximates the "Top of Active Fuel" which is a water level at which clad damage may be expected to occur.
Barrier Analysis Fuel Clad Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based upon the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the EOP Critical Safety Function Tree (CSFT) monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG. Use ofRVLIS to indicate reactor vessel water level is more specific than the CFST Core Cooling status codes. Full Range RVLIS indicates reactor vessel water level with no RCPs running. The intent of this EAL is to provide a RVLIS level which approximates core uncovery. The actual RVLIS level which indicates "Top of Active Fuel" is somewhat higher than 39%; however, 39% was adopted to be consistent with the CFST value.
EAL - 3.1.4.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION Salem Generating Station replaced the CFST "Orange Path" color designation with "Purple Path" due to the limitations imposed by the SPDS CRT's* color gun configuration.
REFERENCES NUMARC NESP-007, FC4 EOP-CFST-1 EOP-TRIP-1 EOP-Setpoint Doc (G.03)
- EAL - 3.1.4.a Rev. 00 Page 2 of 2
SGS EALIRAL Technical Basis 3.0 Fission Product Barriers
- 3.1 Fuel Clad Barrier 3.1.4 RX VESSEL LEVEL INDICATION SYSTEM (RVLIS) 3.1.4.b IC Potential Loss of Fuel Clad Barrier= 3 POINTS EAL RVLIS Dynamic Range Indicates ANY one of the following:
- 4 RCPs I/S < 44%
- 3 RCPs I/S < 30%
- 2 RCPs I/S < 20%
- 1 RCP I/S < 13%
MODE- *1, 2, 3, 4 BASIS The threshold values for RVLIS Dynamic Range levels with various combinations ofRCPs is chosen from the EOP-CFST-1 Core Cooling Status Tree. These values correspond to a 50%
void fraction which may result in clad damage.
Barrier Analysis Fuel Clad Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based upon the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the EOP Critical Safety Function Tree (CSFT) monitoring are integrated into this EAL. The CFSTs are conta~ned as a tab to the ECG. Use ofRVLIS to indicate reactor vessel water level is more specific than the CFST Core Cooling Purple Path status codes. Dynamic Range R VLIS indicates reactor vessel water level when at least 1 RCP is EAL - 3.14.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis running. The intent of this EAL is to provide aRVLIS level which approximates a 50% RCS void fraction. With this void fraction, a loss of all operating RCPs could lead to core uncovery.
DEVIATION
- Salem Generating Station replaced the CFST "Orange Path" color designation with "Purple Path" due to the limitations imposed by the SPDS CRT's color gun configuration.
REFERENCES NUMARC NESP-007, FC4 EOP-CFST-1 EOP-TRIP-1 EOP-Setpoint Doc (L.01)
EAL - 3.1.4.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 *Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.5 CONTAINMENT RADIATION.LEVELS IC : - Loss of Fuel Clad Barrier= 4 POINTS j_*1 .
EAL R44A or R44B > 300 R/hr MODE - .1, 2, 3, 4 BASIS The reading of300 R/hr on the containment high range monitor (R44A or R44B) indicates the loss of the Fuel Clad fission product barrier. The reading was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment at an RCS Activity of 300 µCi/gm Dose Equivalent Iodine 131. This value is much larger than Technical Specification allowed Iodine spikes and corresponds to fuel clad damage of approximately 2.5%.
Barrier Analysis Fuel Clad and RCS Barriers have been lost..
ESCALATION CRITERIA This event will be classified and/or escalated based upon the loss or potential loss of the Primary Containment barrier per EAL Section 3.0 DISCUSSION This calculation is based upon a concentration of300 µCi/gm Dose Equivalent Iodine 131 as it relates to R44 measured Dose Rate values. This calculation was prepared by the Nuclear Fuels Group and is on file with Emergency Preparedness under file title DSI.6-00XX: "Verification of Emergency Action Levels for Event Classification" date 1/26/95.
EAL - 3.1.5 Rev.00 Page 1 of 2
SGS EAL!RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, FCS Calculation by Nuclear Fuels file title DS 1.6-00:XX "Verification of Emergency Action Levels for Event Classification EAL - 3.1.5 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 Fission Product*Barriers 3.1 Fuel Clad Barrier 3.1.6 EMERGENCY COORDINATOR JUDGMENT 3.1.6.a/ 3.1.6.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 MODE- 1, 2, 3, 4 BASIS This EAL allows the Emergency Coordinator (EC) to address any factor not otherwise covered in the Fission Product Barrier Table to determine that the Fuel Clad barrier has been lost or potentially lost. A complete loss in the ability to monitor the Fuel Clad barrier should be considered a "Potential Loss" of that barrier.
Barrier Analysis The Fuel Clad Barrier has been lost or potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the loss or potential loss of additional barriers per EAL section 3.0.
DISCUSSION None DEVIATION None EAL - 3.1.6.a/ 3.1.6.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, FC7 EAL - 3 .1.6.a/ 3. 1.6.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.1 CRITICAL SAFETY FUNCTION STATUS 3.2.1.a IC Potential Loss of RCS Barrier= 3 POINTS EAL THERMAL SHOCK RED PATH MODE - 1, 2, 3, 4 BASIS Thermal Shock RED Path, as verified by EOP-CFST-1, indicates an excessive RCS cooldown has occurred and that RCS pressure and temperature conditions have resulted in significant Pressurized Thermal Shock concerns.
Barrier Analysis RCS Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
DEVIATION None EAL - 3.2.1.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, RC 1 EOP-CFST-1 EOP-TRIP-1 EAL - 3.2. l .a Rev. 00 Page 2 of 2
SGS EAl./RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier
.3.2.1 CRITICAL SAFETY FUNCTION STATUS 3.2.1.b IC Potential Loss of RCS Barrier= 3 POINTS EAL
~.HEAT SINK RED PATH MODE- 1, 2, 3, 4 BASIS Heat Sink RED Path, as verified by EOP-CFST-1, indicates that Stearn Generator dryout could occur. A loss of Heat Sink poses an e?{treme heat removal challenge to the RCS. A barrier loss classification should not be made ifthe Heat Sink RED Path is the result of procedurally required Auxiliary Feedwater flow control.
Barrier Analysis Fuel Clad and RCS Barriers have been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3. 0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
DEVIATION None EAL - 3.2. l .b Rev. 00 Page 1 of 2
SGS.EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, RCl EOP-CFST-1 EOP-TRIP-1 FRHS-1 EAL - 3.2.1.b Rev. 00 Page 2 of 2
SGS EAL(RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 1.2.2 RCS LEAK RA TE 3.2.2.a IC Potential Loss of RCS Barrier= 4 POINTS EAL One Centrifugal Charging Pump CANNOT maintain PZR level> 17% (as a result of RCS leakage).
MODE - l; 2, 3, 4 .
BASIS RCS leakage which results in an inability to maintain Pressurizer (PZR) Level with a normal charging lineup using one Centrifugal Charging Pump is indicative of an RCS inventory loss which would require initiation of Safety Injection (SI) and entry into EOP-TRIP-1 from OP-AB.RC-0001 (Q), Reactor Coolant System Leak.
Non-RCS leakage events (such as steam/ feedwater system breaks) where no mass is lost from the RCS should not be classified under this EAL.
Barrier Analysis RCS Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Significant leakage from the RCS will result in implementation ofOP-AB.RC-OOOl(Q). Actions required by this procedure will result in one Centrifugal Charging Pump in service, discharging to the charging header, and Letdown secured. If Pressurizer Level cannot be maintained stable or rising with this lineup established, a manual Safety Injection will be initiated. This EAL assumes EAL - 3 .2.2.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis that any event that would result in significant RCS mass loss will result in at least an ALERT declaration.
DEVIATION None REFERENCES NlJMARC NESP-007, RC2 EOP-TRIP-1 EOP-FRCE-1 EOP- Setpoint Doc (D.02)
OP-AB.RC-0001 (Q)
EAL - 3.2.2.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 *RCS Barrier 3.2.2 RCS ..LEAK RA TE-
- 3.2.2.b
.JC Loss of RCS Barrier= 4 POINTS EAL Subcooling is 0 °F as a result of RCS leakage MODE- 1, 2, 3, 4 BASIS This EAL attempts to classify a "Loss" of the RCS Barrier due to LOCA conditions. Non-RCS leakage events (such as steam/feedwater system breaks) where no mass is lost from the RCS should not be classified under this EAL. Subcooling equal to 0 °F is indication that leakage from the RCS boundary is greater than the available inventory control capacity. The loss of subcooling signifies that the inventory control systems are inadequate to maintain RCS press_ure and inventory against the mass loss through the leak.
Loss of subcooling due to, or as a result of, EOP directed operator actions do not require classification under this EAL.
Barrier Analysis RCS Barrier has been Jost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Status Tree (CFST) Monitoring are integrated irito this EAL. The CFSTs are contained as a tab EAL - 3.2.2.b Rev. 00 Page 1 of 2
- SGS EAL!RALTechnical Basis to the ECG. The intent of using CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the control room.
The EAL threshold of 0 °F is reached by CFST Core Cooling YELLOW or Continuous Action
. Summary (CAS) monitoring. It is not intende~ to use this EAL for Primary to Secondary leakage events since adequate injection capability should exist for all ranges of these events including Steam Generator Tube Rupture (SGTR).
~
EOP directed actions resulting in deliqerate subcooling reduction (e.g. during SGTR saturated recovery), steam/feedwater line breaks, or momentary reductions below 0°F that are recoverable
, (~J~* SI flow reduction sequence) should not be classified under this EAL.
DEVIATION None REFERENCES NUMARC NESP-007, RC2 EOP-CFST-1 EOP-TRIP-1 EOP-Setpoint Doc (R.01)
EAL - 3.2.2.b Rev. 00 Page 2 of 2
- SGS EALIRALTechnical Basis 3.0 Fission Pr<~duct Barriers 3.2 RCS_ Barrier
.3.2.3 STEAM GENERATOR TUBE.RUPTURE
- .3.2.3.a ,*...
IC Potential Loss of RCS Barrier= 3 POINTS "EAL O~e Centrifugal Charging Pump CANNOT maintain PZR level> 17% (as a result of a SGTR)
Control Room has determined that an SGTR has occurred MODE - 1, 2, 3, 4 BASIS This EAL is indicative of a Loss of RCS from a Steam Generator Tube Rupture (SGTR). Non-RCS leakage events (such as steam/feedwater system breaks) where no mass is lost from the RCS should not be classified under this EAL. The threshold values for determining a SGTR are those used in the EOP network. Inability to maintain Pressurizer (PZR) Level with a normal charging lineup is indicative of a SGTR that would require initiation of SI and entry into EOP-TRIP-1.
Barrier Analysis RCS Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION It is understood that this EAL is redundant to the RCS leakage EAL. Inclusion of this EAL ensures that significant SG tube leakage will be 'classified consistent with RCS leakage. Known SG tube leakage will result in implementation of OP-AB.SG-OOOI(Q). Actions required by this EAL - 3.2.3.a Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis procedure may result in a manual Safety Injection initiat_ion and entry into the EOP network. This EAL assumes that any SGTR that results in significant RCS mass loss will result in at least an ALERT classification.
For Ruptured SGs that are also faulted, further evaluation of the Containment Barrier is required.
-For faults that occur inside of Containment this "Potential Loss" EAL will serve as the correct classification as long as no Containment challenges occur. For faults which occur outside the Containment, the RCS SGTR "Loss" EAL must also be considered.
DEVIATION None REFERENCES NUMARC NESP-007, RC3 EOP-SGTR-1 S 1(2).0P-AB.SG-OOOl(Q)
EOP-S~tpoint Doc (D.02)
EAL - 3_2_3_a Rev_ 00 Page 2 of 2
SGS EALIRALTechnical Basis 3~0
- Fission -Product Barriers 3.2 RCS-Barrier*
3.2.3 STEAM GENERATOR-TUBE RUPTURE 3~2.3;b IC Loss of RCS Barrier= 4 POINTS .. *
Ruptured Steam Generator pressure is dropping in an uncontrolled manner or completely
- depressurized
.Prolonged, direct secondary leakage to the environment (e.g. steam breaks, feed breaks, stuck open safety or relief valves) NOTE: SEE 3.3.4.b MODE- 1, 2, 3, 4 BASIS This EAL is indicative of a loss of RCS inventory due to a Steam Generator Tube Rupture (SGTR) and the Ruptured SG is also Faulted outside Containment. The threshold values for determining that a SGTR exist are those used in the EOP network. This condition results in a prolonged, direct release of radioactive fission and activation products to the environment.
This EAL does not include SG depressurization events that are a direct result of EOP directed operator action. The term "dropping in an uncontrolled manner" is defined consistent with the EOP definition of a Faulted SIG. A "prolonged" release is defined as an unisolable rupture (steam breaks, feed breaks, stuck open safety or relief valves excluding minor valve leakage) of a steam or feed line outside of Containment, or a stuck open relief valve on the ruptured SG.
The term "direct secondary leakage to the environment" is intended to include all flowpaths of contaminated secondary coolant to the environment either directly or via systems which exhaust to the Plant Vent (e.g.; leakage to the Auxiliary Building ventilation system) with the following EAL - 3_2_3.b Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis exception: If the EOPs require steaming the ruptured SG to the main condenser, the condenser off-gas (Rl5) pathway is excluded from this EAL provided the release is both controlled and monitored.
Barrier Analysis _
__ RCS and Containment Barriers have been lost.
ESCALATION CRIJ'ERIA This event will be classified and/or escalated based on the potential loss or loss of the Fuel Clad Barner per EAL Section 3 .1.
DISCUSSION This "Loss" EAL addresses Ruptured SGs with an unisolable fault outside of Containment. This EAL is used in conjunction with-the Containment Barrier Bypass "Loss" EAL and will always result in a loss of the Containment Barrier. Ruptured SGs that are faulted inside the Containment are-excluded from this EAL. This EAL excludes classification based on a depressurization that
- results from an EOP induced cooldown of the RCS that does not involve prolonged release of contaminated secondary coolant from the affected SG to the environment. Releases which reach the environment via the Plant Vent should also be classified under this EAL.
DEVIATIONS None REFERENCES NUMARC NESP-007, RC3 EOP-SGTR-1 S 1(2).0P-AB.SG-OOOl(Q)
EOP-Setpoint Doc (D.02)
EAL - 3_2_3.b Rev_ 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barri~rs 3.2 RCS,Barrier 3.2.4 CONTAINMENT RADIATION LEVELS IC Loss of RCS Barrier= 4 POINTS EAL , *-c Valid Containment Radiation level which exceeds ANY one of the following Containment Rad_
Monitor values:
- R2>1R/hr
- R44A > 10 R/hr
- R44B > 10 R/hr MODE - 1, 2, 3, 4 BASIS A reading of>l R/hr on 130' Containment Area Rad Monitor R2 is the preferred method of classification under this EAL. The measurement scales on R2 range from 0.1 rnR/hr to 10 R/hr thus providing reasonable accuracy for this threshold value.
The term "valid" was added specifically for the Containment High Range R44 detectors as they are log scale detectors scaled only in R/hr and are extremely inaccurate at this low value. This reading is less than that specified for the loss of Fuel Clad Barrier since this EAL attempts to identify RCS leakage assuming RCS activity at the Technical Specification limit.
Classification under this EAL should not be made based upon crud burst evolutions or other non-RC S leakage events.
Barrier Analysis RCS Barrier has been lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the loss or potential loss of additional barriers per EAL Section 3.0.
EAL.- 3.2.4 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION ..
The R44A/B detectors were included in this EAL to ensure that classification of an RCS "loss" would occur for events which result in significant R/hr -readings on these high range detectors which "over scale" the R2 detector. It is understood that these detectors are incapable of accurately.reading 1 R/hr due to their log function (with 1 R/hr being the setpoint for coming "off the lower peg"). Therefore the EAL threshold value for these monitors has been increased to 10 R/hr which corresponds to the upper range of the R2 monitor.
The threshold value of 1 R/h{for the R2 monitor was calculated assuming an instantaneous release of the Reactor Coolant volume into the Primary Containment at a coolant concentration of 1.0 µCi/gm Dose Equivalent 1-131 (Technical Specification limit). This calculation was prepared by the Nuclear Fuels Group and is on file with Emergency Preparedness under file title DS 1.6-00XX "Verification of Emergency Action Levels for Event Classification" dated 1/26/95. This RAD monitor value is to be used as a backup indication to other systems designed to measure RCS leakage.
-DEVIATION None REFERENCES NUMARC NESP-007, RC4
- Calculation by Nuclear Fuels Group file title DS 1. 6-00XX "Verification of Emergency Action Levels for Event Classification" dated 1/26/95.
EAL - 3.2.4 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.2 *RCS Barrier 3.2.5 EMERGENCY COORDINATOR JUDGMENT 3.2.5.a/ 3.2.5.b IC Potential Loss ( *::::£ 3 POINTS) orLoss 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 MODE- 1, 2, 3, 4 BASIS This EAL allows the Emergency Coordinator (EC) to address any factor not otherwise covered in the Fission Product Barrier Table to determine that the RCS barrier has been lost or potentially lost. A complete loss in the ability to monitor the RCS barrier should be considered a "Potential Loss" of that barrier.
Barrier Analysis The RCS Barrier has been potentially lost or lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION None DEVIATION None EAL - 3.2.5.a/ 3.2.5.b Rev.00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, RC6
'*:j EAL - 3.2.5.a/ 3.2.5.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.1 CRITICAL SAFETY FUNCTION STATUS 3.3.1.a IC Potential 'Loss of Containment Barrier = 1 POINT EAL II CNTMT ENVIRONMENT RED PAIB MODE- 1, 2, 3, 4 BASIS Containment Environment RED Path, as verified by EOP-CFST-1, results from RCS barrier loss or a faulted SIG inside Containment and signifies that breach of the Primary Containment is imminent. For this condition, all Containment isolations, as well as automatic Containment Spray and CFCU "low speed" operation should be initiated before this threshold is reached.
Barrier Analysis Containment Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room.
Although the yield strength of the Primary Containment may be much higher that 4 7 psig, for the purposes of event classification, the barrier is considered potentially lost at that value. Thus, this EAL - 3.3.1.a Rev. 00 Page 1 o"f 2
SGS EALIRALTechnical Basis EAL is primarily a discriminator between a Site Area Emergency and a General Emergency, representing a potential loss of the third barrier.
DEVIATION None REFERENCES NUMARC NESP-007, PCl EOP-CFST-1 EOP-TRIP-1 EOP-FRCE-:1 EAL - 3.3.1.a Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3~3~1 CRITICAL SAFETY FUNCTION STATUS 3.3.l.b IC Potential '.Loss of Containment Barrier = 1 POINT EAL i CORE COOLING RED PAIB for.> 15 minutes MODE-.1, 2, 3; 4 BASIS Core Cooling RED Path, as verified by EOP-CFST-1, represents an imminent melt sequence which if not corrected could lead to Reactor Vessel failure and potential for Containment failure.
The 15 minutes is used as a threshold for indicating that operator actions have not been effective in restoring core cooling.
Barrier Analysis Fuel Clad Barrier has been lost, RCS and the Containment Barriers have been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the loss of an additional barrier per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using confirmed CFST status in this EAL is to simplify the identification of the .EAL threshold criteria monitored in the Control Room.
Severe accident analysis has concluded that functional restoration procedures can arrest core degradation within the Reactor Vessel in a significant fraction of the scenarios, and that the EAL - 3.3. l.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis likelihood of Containment failure in these scenarios is small. It is appropriate, therefore, to allow a reasonable period of time for the functional restoration procedures to arrest the core melt sequence. It should be apparent within 15 minutes ifthe procedures will be effective. The Emergency Coordinator should make the classification as soon as it is determined that the procedures
. have been, or will be, ineffective.
DEVIATION None REFERENCES NlJMARC NESP-007, PC6 EOP-CFST-1 EOP-TRIP-1 EAL - 3.3.1.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.2 CONTAINMENT PRESSURE 3.3.2.a
MODE - 1, 2, 3, 4 BASIS Hydrogen gas can be present in the Containment at the threshold level only as a result of an inadequate core cooling accident, substantial zirc-water reaction, and a breach of the RCS.
Containment H2 level above 4% signifies that an explosive mixture may exist.
Barrier Analysis Containment Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION A 4% mixture ofH2 with normal Containment atmosphere represents the deflagration lower limit.
Any subsequent ignition and bum of this level mixture releases a substantial amount of energy that must be absorbed by the Containment structure, which is already under stress due to the Loss of the RCS Barrier.
DEVIATION None EAL - 3 .3 .2.a Rev. 00 Page 1 of 2
SGS:-EAL!RALTechnical Basis REFERENCES NUMARC, NESP-007, PC2 EOP-TRIP-1 EOP-FRCE-1 EOP-Setpoint Doc (T .18)
EAL - 3.3.2.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers
- 3.3 Containment Barrier
.:3_.3*2 : CONTAINMENT PRESSURE 3.3.2.b
,JC *Potential Loss of Containment Barrier=. l POINT EAL CNTMT-Press. > 15 psig with EITHER one of the following:
- No CNTMT Spray AND < 5 CFCUs Running in "Low Speed"
- One CNTMT Spray Train I/SAND< 3 CFCUs Running in "Low Speed" MODE - 1, 2, 3, 4 BASIS*
Containment (CNTMT) pressure increase to> 15 psig (the CNTMT Spray initiation setpoint) indicates a major release of energy to the Containment. Failure of ALL Containment Spray with
<5 Containment Fan Coil Units (CFCUs) running in "low speed", or only one train of Containment Spray in service with <3 CFCUs running in "low speed", indicates a condition where systems designed for containment heat removal and depressurization do not have the capacity to maintain Containment pressure below the structural design limit. The threshold value for available Containment Depressurization and Cooling Systems is based upon system design basis for maintaining Containment integrity.
Barrier Analysis Containment Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION EAL - 3.3.2.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis The CFCUs and the Containment Spray system are redundant to each other in providing post accident cooling of the Containment atmosphere. With less than the minimum combination of sub-systems stated in the EAL threshold value, the ability to remove energy from the Containment atmosphere is severely impaired .. Containment pressure > 15 psig with a loss of Containment Cooling and Depressurization systems represents a potential loss of the Containment barrier.
DEVIATION None -l *~ .
REFERENCES NUMARC, NESP-007, PC2.
EOP-TRIP-1 EOP-FRCE-1 EOP-Setpoint Doc (T.02)
Technical Specification Section 3.6.2 EAL - 3.3.2.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.2 CONTAINMENT PRESSURE 3;3.2,;C IC Loss of Containment Barrier= 2 POINTS EAL A Rapid Unexplained Containment Pressure Drop* following an initial Rise to
> 4 psig BASIS Containment pressure increase to> 4 psig (the containment pressure Safety Injection initiation setpoint) indicates a major release of energy to the Containment. These releases can only be provided by a large release of either primary or secondary coolant into the Containment. For the cases that primary coolant provides the source of energy, a loss of the RCS barrier has also occurred. A rapid unexplained loss of Containment pressure following an initial pressure rise indicates a loss of Containment integrity.
Unexplained means that the pressure drop is not as a result of operator actions taken to reduce Containment pressure. The term rapid was added as an attempt to quantify the size of the Containment breach.
Emergency Coordinator judgment should be used to determine if this EAL applies for rapid, unexplained Containment pressure drops following initial rises to less than the 4 psig threshold.
Barrier Analysis Containment Barrier has been lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
- EAL - 3.3.2.c Rev. 00 Page 1 of 2
.SGS EAL/RALTechnical Basis DISCUSSION The threshold value of 4 psig was -selected to be consistent with the Safety Injection and Adverse Containment criteria. For those cases where secondary coolant provides the source of energy, a faulted Steam Generator is possible. This requires actions in EOP-LOSC-1 to isolate the Main Steam lines to maintain intact Steam Generators for an RCS Heat Sink, minimize Containment Pressure, and to minimize RCS cooldown.
DEVIATION.
None REFERENCES NUMARC NESP-007, PC2 EOP-TRIP-1 EOP-LOSC-1 Technical Specification Table 3.3-4 EAL - 3.3.2.c Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3~3.3 CONTAINMENT ISOLATION
MODE- 1, 2, 3, 4 BASIS The Containment (CNTMT) Sump threshold of78% (75% adverse) is based upon containment flooding concerns, and is consistent with the CFST level requiring implementation ofEOP-FRCE-
- 2. An indicated level greater than this value indicates that water has been introduced into the Containment from other sources. Potential flooding of critical system components and instrumentation required for responding to an accident or performing an orderly shutdown may be affected. Thus the Containment and associated systems may not be capable of performing their function as a fission product barrier.
Barrier Analysis Containment Barrier has been potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring are integrated into this EAL. The CFSTs are contained as a tab to the ECG.
The intent of using CFST status in this EAL is to simplify the identification of the EAL threshold criteria monitored in the Control Room. The EAL threshold of>78% (75% adverse) CNTMT sump level is consistent with the CFST criteria.
EAL - 3.3.3.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION . ':°'
None REFERENCES.
NUMARC NESP-007, PC7 EOP-TRIP-1 EOP-FRCE-1 EOP-FRCE-2 *.
EOP-Setpoint Doc (T.07, T.08)
EAL - 3.3.3.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.3 CONTAINMENT ISOLATION 3.3.3.b IC Loss of Coiltairiinent Barrier = 2 POINTS EAL Valid CNTMT <!>A, <l>B or CNTMT Vent Isol Signal Flow path from CNTMT to the environment MODE- 1, 2, 3, 4 BASIS A valid Containment (CNTMT) Isolation Signal represents a situation that requires closure of selected Containment Isolation valves to maintain containment integrity under abnormal conditions. The lines required to be isolated under these conditions connect potentially contaminated systems or Containment volume with systems outside the Containment.
Classification under this EAL is not required if manual closure attempts from Control Room are successful in the event that the automatic isolation signal fails. The term "valid" is defined as an actual condition which requires a CNTMT isolation due to instrumentation setpoints being exceeded and was included to exclude those conditions where Containment Isolation is not required, but has somehow resulted in a violation of CNTMT integrity.
The term "to the environment" is intended to include ANY flow path to the environment either directly or via systems which exhaust to the Plant Vent (e.g.; leakage to the Auxiliary Building ventilation system).
Barrier Analysis Containment Barrier has been lost.
EAL - 3.3.3.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis ESCALATION CRITERIA
..*This ev_ent will be classified and/or-escalated based on the potential loss or loss of additional barriers per EAL Section 3. o.
DISCUSSION
'rechnic~l Specifi~atioii3.6.3 ,;Containment Isolation Valves" was used to determine the signals required for Containment isolation. Any reference to Main Steam Isolation or Steam Generator Blowdown Isoiation is covered under the Containment Bypass "potential loss" EAL.
DEVIATION None REFERENCES NUMARC NESP-001';"-PC3 EOP-TRIP-1 OP-AR.ZZ-0003(Q)
SGS Technical Specifications EAL - 3.3.3.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 3.0 :Fission Product Barriers 3.3 Containment Barrier
- 3.3.4 RCS LINE BREAK/CONTAINMENT BYPASS 3.3.4.a IC Potential Lbss *of Containment Barrier ~ i POINT EAL Unisolable, Faulted Steam Generator OUTSIDE of Containment as indicated by SIG pressure dropping in an uncontrolled manner or completely depressurized Affected S/G tubes are intact MODE- I, 2, 3, 4 BASIS S/Gs which have unisolable faults outside of the Containment will require feed isolation and secondary side dryout in order to stop the resultant excessive RCS cooldown rate. This subsequent dryout will result in significant thermal stress and differential pressures across the tube sheet and greater risk of a SGTR on an already faulted S/G. As such, this event is considered to be a precursor to a more serious event and will lead to at least an Unusual Event classification.
This EAL excludes S/G depressurization ev.ents that are a direct result ofEOP directed operator action. The term "dropping in an uncontrolled manner or completely depressurized" is defined consistent with the EOP definition of a Faulted S/G. "Unisolable" is defined as a condition where manual isolation is not possible such as a pipe rupture with no accessible isolation valves, a stuck open safety or relief valve, etc. (excluding minor valve leakage).
Barrier Analysis Containment Barrier has been potentially lost.
ESCALATION CRITERIA EAL - 3.3.4.a Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis This event will be classified and/or escalated based on the potential loss or loss of additional barriers per EAL Section 3.0.
DISCUSSION
-This EAL was added. to the Fission Product Barrier Table 3. 0 as a Containment Bypass "Potential Loss" to ensure that all unisolable steam or feedwater break events, where the fault is outside of the Containment, are at least classified as an Unusual Event. The "potential loss" category (1 point) was selected to ensure that further challenges to other Fission Product Barriers result in Emergency Classifications consistent with current philosophy.
The Containment Barrier section was selected since Technical Specifications Section 3.6.3 "Containment Isolation Valves" require both Main Steam Isolation and Steam Generator Blowdown Isolation. The Containment Bypass sub-section was selected based upon the leakage being non-radioactive steam or feedwater with concerns for RCS integrity appropriately classified under the RCS Barrier section. An NRC inspection at Calvert Cliffs Nuclear Plant resulted in the addition of this EAL.
DEVIATION This EAL was added as a Potential Loss of Containment due to the Containment Bypass concern discussed in HUS "Uncontrolled RCS cooldown due to Secondary Depressurization". A review ofNRC Inspection Report Nos. 50-317/94-27; 50-318/94-27 for the Calvert Cliffs Nuclear Power Plant indicated that an unisolable, faulted SIG outside of containment represents at least a UE Classification. Technical Specification 3.6.3 for Containment Isolation Valves require OPERABLE Main Steam Isolation valves MS7s and MS18s. The Main Steam Isolation Valves (MS 167s) also receive a MSL Isolation Signal but are covered under their own Tech. Spec
- 3. 7. 1. 5. Therefore, failure of any Main Steam Isolation valve to close upon demand represents a potential loss of Containment integrity and was included in the Fission Product Barrier Table in order to properly classify events in conjunction with the RCS and Fuel Clad Barriers.
REFERENCES NUMARC NESP-007, PC7 NRC Inspection Report 50-317/94-27 EOP-TRIP-1 EOP-LOSC-1 OP-AB. STM-0001 (Q)
EAL - 3.3.4.a Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 3.'0 *Fission Product Barriers 3.3 Containment Barrier 3;3;4 RCS LINE BREAK/CONTAINMENT BYPASS 3j.4.b '
IC Loss of Containment Barrier = 2 POINTS EAL Primary to Secondary Leakage > Tech Spec Limits Prolonged, direct secondary leakage to the environment MODE- 1, 2, 3,'4 BASIS Primary to Secondary leakage greater than Technical Specifications along with indication of prolonged secondary side leakage outside the Containment indicates a Steam Generator (S/G) tube leak that is discharging directly to theenvironment. "Prolonged" is defined as an unisolable rupture (excluding minor valve leakage) of a steam or feed line outside of Containment, or a stuck open safety or relief valve on a secondary system connected to the steam side of the leaking SIG.
The term "direct secondary leakage to the environment" is intended to include all flow paths of contaminated secondary coolant to the environment either directly or via systems which exhaust to the Plant Vent (e.g.; leakage to the Auxiliary Building ventilation system) with the following exception: If the procedure in effect requires steaming the leaking S/G to the main condenser, the Condenser Air Ejector (RI 5) pathway is excluded from this EAL provided the release is both controlled and monitored.
For Steam Generator Tube Rupture (SGTR), this EAL is used in conjunction with the RCS Barrier SGTR EALs to ensure proper classification ifthe Ruptured S/G is also faulted outside of Containment.
Barrier Analysis Containment Barrier has been lost.
EAL - 3.3.4.b Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of additional b~rriers per EAL Section 3.0.
DISCUSSION Tlw primary intent of this EAL is to ensure, in conjunction with the RCS Barrier "Loss" SGTR EAL, that Ruptured S/Gs that are also faulted outside of Containment, are classified as at least a Site Area Emergency. The threshold for establishing the bypass of Containment was intended to be a_prolonged release of radioactivity from the Ruptured SIG directly to the environment.
The secondary purpose of this EAL is to classify S/G tube leak events which exceed Technical Specification limits, but do not exceed the RCS Barrier SGTR thresholds. If a prolonged release occurs from a S/G during a leak, only an Unusual Event would be declared based on the "Loss" of the containment barrier.
DEVIATION None '*
REFERENCES NUMARC NESP-007, PC4 EAL - 3.3.4.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical.Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3;3.4 RCS LINE BREAK/CONTAINMENT BYPASS 3.3.4.c IC Loss of Containment Barrier = 2 POINTS EAL LOCA conditions CNTMT Press. OR Sump Level NOT rising as expected MODE- 1, 2, 3, 4 BASIS The threshold conditions require that a Loss of Coolant Accident (LOCA) is known to be occurring. Such events are accompanied by release of energy and inventory from the RCS to the Containment (CNTMT), and should result in pressure and sump level rise in the Containment.
Failure of CNTMT Pressure or Sump Level indications to rise as expected following a known LOCA is an indication of a Containment Bypass situation.
Barrier Analysis Containment and RCS Barriers have been lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the potential loss or loss of the Fuel Clad Barrier per EAL Section 3 .1.
DISCUSSION A LOCA is expected to result in CNTMT pressure rise to > 4 psig. This leak rate should result in the accumulation of RCS inventory in the CNTMT Sump as well as a CNTMT SUMP PMP EAL - 3.3.4.c Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis START OHA as the level rises. A lack of expected CNTMT Sump level response or CNTMT pressure not rising indicates that the Containment Barrier has been bypassed.
DEVIATION None REFERENCES
-~ ...
NUMARC NESP-007, PC2 EOP-TRIP-1 EOP-LOCA-6, LOCA Outside Containment OP-AR.ZZ-0003(Q)
EAL - 3.3.4.c Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 3.0 Fission Product Barriers
- 3.3~ Containment Barrier 3.3:5 *CONTAINMENT RADIATION LEVELS IC Potential Loss of Containment Barrier= I-*POINT EAL l:.
MODE - 1, 2, 3, 4 BASIS A Containment High Range Monitor (R44) reading in excess of2000 RJhr indicates significant Fuel Clad damage, well in excess of that corresponding to a loss of the RCS and Fuel Clad barriers. The value corresponds to *a release of approximately 20% of the gap region.. Regardless of whether Containment is challenged, this amount of activity in Containment, if released, could have severe consequences and it is prudent to treat this as a potential loss of the Containment Barrier.
Barrier Analysis Containment Barrier has been potentially lost, the Fuel Clad and RCS Barriers have been lost.
ESCALATION CRITERIA NIA DISCUSSION This calculation is based upon a calculation of 20% Clad Damage as it relates to R44 measured Dose Rate values. This calculation was prepared by the Nuclear Fuels Group and is on file with Emergency Preparedness under file title DS 1.6-00XX "Verification of Emergency Action Levels for Event Classification" date 1/26/95.
DEVIATION None EAL- 3.3.5 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, PCS NUREG-1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents Calculation by Nuclear Fuels file title DSI.6-00XX "Verification of Emergency Action Levels for Event Classification , ** , .
EAL- 3.3.5 Rev. 00 Page 2 of 2
- SGS EAL/RALTechnical Basis 3.0 Fission Product Barriers
- . -3.3 Containment Barrier *. *..
3.3.6
- EMERGENCY COORDINATOR JUDGMENT
". 3.3.6.a/ 3.3.6.b* :.
IC Potential Loss ( = 1 POINT) 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 MODE- 1, 2, 3, 4 BASIS *_
This EAL allows the Emergency Coordinator (EC) to. address any factor not otherWise covered in
,the Fission Product Barrier Table to determine that the Containment barrier has been lost or potentially lost. A complete loss in the ability to monitor the Containment barrier should be considered a "Potential Loss" of that barrier Barrier Analysis Containment Barrier has been lost or potentially lost.
ESCALATION CRITERIA This event will be classified and/or escalated based on the loss or potential loss of additional barriers per EAL Section 3.0.
DISCUSSION None DEVIATION None EAL - 3.3.6.a/ 3.3.6.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, PCS
- EAL - 3.3.6.a/ 3.3.6.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 4.0 EC Discretion 4.1 Emergency Coordinator Discretion UNUSUAL EVENT- 4.1.1 lu\
IC _Other Conditions Exist Which Inthe Judgment of the Ernergency 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
- MODE-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
SGS EAL/RALTechnical 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 to Section 6 of the ECG if a RadiologiCal Release is ongoing.
DEVIATION None*
REFERENCES NUMARC NESP-007, HUl.3, HUS, Section 3.7.
EAL - 4.1.1 Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 4~0 EC Discretion
- 4. l 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 Coordinafor, 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 MODE-All BASIS Emergency Coordinator (EC) judgment to declare an Alert, based on the determination that Plant 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 more 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
SGS 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 (P AG) plume exposure level, i.e., 10 to 100 mRem,rEDE. Refer ~o ECG Section 6 if a Radiological Release is ongoing.
DEVIAJ'ION
.*, *~;
None
~- .
.REFERENCES NUMARC NESP-007, HA6, HAl.4, Se~tion 3.7.
EPA-400 EAL - 4. l.2 Rev.00 Page 2 of 2
SGS EALIRALTechriical 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 protection of the public MODE-All BASIS Emergency Coordinator (EC) judgment to declare 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 I 00 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 L
SGS 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 >I ooo 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. l.3 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 4.0 EC Discretion 4.1 *Emergency Coordinator Discretion
. i 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 MODE-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
SGS EAL/RALTechnical Basis DEVIA'FION.
None*
REFERENCES.
NUMARC NESP-007, HG2, Section 3. 7.
EPA-400 EAL - 4.1.4 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 5.0 Failure to Trip 5.1 ATWT ALERT - 5.1.2.a/5.1.2.b . --~ ... ,.... -...
~
IC Failure of the RPS to Successfully Complete a Reactor Trip (Automatic or Manual)
EAL EITHER one of the following conditions are met:
- Reactor Protection System Trip Setpoint Exceeded AND an Automatic Reactor Trip is NOT Confirmed
- ANY Manually Initi!lted Reactor Trip from the Control Room is NOT Confirmed MODE- I, 2, 3 BASIS This condition indicates failure of the Reactor Protection System to trip the Reactor, either
- automatically or on manual demand. This condition is more than a potential degradation of a safety system in that a front line protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel or Reactor Vessel may have been exceeded. An Alert is indicated because conditions exist that could lead to a potential loss of the fuel clad or RCS barriers.
The term "from the Control Room" is defined as any action taken by the NCOs in the Control Room Area which results in a rapid insertion of Control Rods into the core. The term for expressing an unsuccessful trip as "NOT confirmed" is defined as listed in the EOP network.
Confirmed Manual reactor trip is not considered successful if actions away from the Control Room Area (e.g. dispatch of an NEO to locally open the Reactor Trip Breakers) were required to trip the reactor.
ANY unsuccessful Manual attempt to trip the reactor will still be classified under this EAL regardless of the success of additional manual attempts. Any single manual attempt failure will constitute a major breakdown of a system designed to directly protect the health and safety of the General Public.
EAL - 5.1.2.a/5. l.2.b Rev. 00 Page 1 of 3
SGS EAL/RALTechnical Basis 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*ofthose bairiers.
ESCALATION CRITERIA For the case in which' the manual trip from the control room is not* successful with Reactor Power
~** 5%, this event would be escalated* to a Site Area Emergency.
DISCUSSION Entry-into EOP-FRSM-1 may be required ifthe manual Reactor Trip from the console Trip* *.
Handle"*or P-9-is not successful. Additional control console actions taken in EOP-TRIP-1, such as opening the Reactor Trip or opening 2E6D or 2G6D breakers to deenergize the Rod Drive MG Sets, would constitute a successful manual reactor trip from the Control Room. Manual trip is any action by the reactor operator at the controls which causes the.control rods to be rapidly inserted into the core and bring the reactor subcritical.
The threshold value of 5% reactor power for escalation criteria was selected to be consistent with EOP""FRSM-1 *entry criteria. Under these low power conditions, the reactor is providing less heat than the maximum decay heat load for which the safety systems are designed.
DEVIATION*
NUMARCEAL SA2 suggests that an Alert classification be based on an automatic RPS trip failure followed by a successful manual trip from the control room, with EAL SS2 escalating to a Site Area Emergency ifthe manual trip fails. In addition, EAL SS2 basis indicates that the SAE threshold should be such that following the automatic and manual trip failure, the reactor is producing more heat than the maximum for which the safety systems were designed. The EOPs indicate that this heat load is :::,5%.
The Salem Alert threshold was chosen so that unsuccessful manually initiated RPS trips from the control room, as well as unsuccessful automatically initiated trips via RPS would be classified at the Alert level. This will cover those situations which require a manual reactor trip under conditions where an automatic trip signal may not have been generated. In either case, failure of RPS to perform its intended function when demanded is indicated.
The Salem SAE threshold was chosen to include either automatic or manual failure (for the reasons stated above), with resulting power :::,5% as suggested in NUMARC EAL SS2 bases.
By defining an unsuccessful trip as Reactor Trip NOT confirmed (as defined in the EOP network), partial tnps that result in power levels< 5% would be classified as an Alert, whether automatically or manually initiated. *
- EAL - 5.1.2.a/5. l.2.b Rev. 00 Page 2 of 3
-SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, SA2.
EOP-TRIP-1, Reactor Trip or Safety Injection
_EQJ? .. CfST ~-1, Critical Safety Function Trees EAL - 5.1.2.a/5_ l _2.b Rev_ 00 Page 3 of 3
SGS EALIRALTechnical Basis 5.0 Failure to Tnp 5.1 ATWT SITE AREA EMERGENCY - 5.1.3 IC Failure of the RPS to Successfully Complete a Reactor Trip (Automatic or Manual) and Reactor Power is Above 5% * ...
- EAL EITHER one of the following conditions are met:
- Reactor Protection System Trip Setpoint Exceeded AND an Automatic Reactor Trip is NOT Confirmed
- ANY Manually Initiated Reactor Trip from the Control Room is NOT Confirmed ALL Reactor Trip attempts from the Control Room DID NOT reduce (and maintain)
Reactor Power to < 5%
MODE- I, 2 BASIS Failure to trip events should not be classified under this EAL before manual trips have been attempted. Automatic and manual trips are not considered successful if action away from the reactor control console were required to trip the reactor. Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that could lead to imminent loss or potential loss of both the fuel clad and RCS barriers.
The term "from the Control Room" is defined as any action taken by the NCOs in the Control Room Area which result a rapid insertion of Control Rods into the core. The term "reduce (and maintain)" was included to ensure that return to power events are still classified under this EAL.
Although this EAL may be viewed as redundant to the Fission Product Barrier Table EALs, its inclusion is necessary to better assure timely recognition and emergency response.
EAL - 5.1.3 Rev. 00 Page 1 of 3
SGS EAL/RALTechnical Basis 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 (perhaps imminent) loss of those barriers.
ESCALATION CRITERIA For the case in ,~hlch an adequate heat s~nk is not avaifable~ this event would be escalated to a General Emergency per EAL Section 5.1.4. * * '
DISCUSSION Entry into EOP-FRSM-1 will be required ifthe manual trip from the console "trip handle" or P-9
- is not successful. EOP-FRSM-1 requires an Equipment Operator to locally open the Reactor Trip Breakers and trip the Rod Drive MG Sets. Since this action is outside the control room, a successful remote Reactor Trip will require classification under this EAL. The threshold value of 5% reactor power was selected to-be consistent with CFST EOP-FRSM-1 entry criteria. For events which result in a return to > 5% reactor power from some lower value, classification under this EAL would be required: *-* ** , - *
- DEVIATION NUMARC EAL SA2 suggests that an Alert classification be based on an automatic RPS trip failure followed by a successful manual trip from the control room, with EAL SS2 escalating to a Site Area Emergency ifthe_manual trip fails. In addition, EAL SS2 basis indicates that the SAE threshold should be such that following the automatic apd manual trip failure, the reactor is producing more heat than the maximum for which the safety systems were designed. The EOPs indicate that this heat load is >5%.
The Salem Alert threshold was chosen so that unsuccessful manually initiated RPS trips from the control room, as well as unsuccessful automatically initiated trips via RPS would be classified at the Alert level. This will cover those situations which require a manual reactor trip under conditions where an automatic trip signal may not have been generated. In either case, failure of RPS to perform its intended function when demanded is indicated.
The Salem SAE threshold was chosen to include either automatic or manual failure (for the reasons stated above), with resulting power ~5% as suggested in NUMARC EAL SS2 bases.
By defining an unsuccessful trip as Reactor Trip NOT confirmed (as defined in the EOP network), partial trips that result in power levels < 5% would be classified as an Alert, whether automatically or manually initiated.
EAL - 5.1.3 Rev. 00 Page 2 of 3
SGS-EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, SS2 EOP-TRIP-1, Reactor Trip or Safety Injection EOP-CFST-1, Critical Safety Function Trees EAL- 5.1.3 Rev. 00 Page 3 of 3
SGS EAL/RALTechnical Basis 5.0 Failure to Trip
- 5.1 ATWT GENERAL EMERGENCY - 5.1.4 IC Failure of the RPS to Complete an Automatic Trip and Manual Trip was Not successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core EAL EITHER one of the following conditions are met:
- Reactor Protection System Trip Setpoint Exceeded AND an Automatic Reactor Trip is NOT Confirmed
- ANY Manually Initiated Reactor Trip from the Control Room is NOT Confirmed ALL Reactor Trip attempts from the Control Room DID NOT reduce (and maintain)
Reactor Power to < 5%
EITHER one of the following conditions exist:
- CORE COOLING RED PATH
- HEAT SINK RED PATH MODE- 1, 2 BASIS Automatic or manual trips are not considered successful if actions away from the reactor control console were required to trip the reactor. These conditions indicate a fundamental failure of the automatic and manual trip protection of the Reactor Protection System, and are indicative of heat generation significantly greater than the Heat Removal capabilities. The potential for rapid core degradation exists. The General Emergency declaration is intended to be anticipatory of fission product barrier failure and permits maximum offsite intervention time.
EAL- 5.1.4 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis Barrier Analysis If threshold for this;EAL is met, Table 3.0 Fission Pi;-oduct Barriers for Loss of the Fuel
- Clad (Core Cooling RED) and/or Potential Loss of the RCS (Heat Sink RED) Barriers may have been exceeded.
ESCALATION CRITERIA NIA DISCUSSION Entry into EOP-FRSM-1 will be required ifthe manual trip from the console "trip handle" or P-9 is not successful. EOP-FRSM-1 requires an Equipment Operator to locally open the Reactor Trip Breakers and trip the Rod Drive MG Sets. Since this action is outside the control room, a successful remote Reactor Trip will require classification under this EAL. The threshold value of 5% reactor power was selected to be consistent with CFST EOP-FRSM-1 entry criteria. For
- events which result in a return to > 5% reactor power from some lower value, classification under this EAL would be required.
If actions taken in EOP-FRSM-1 are ineffective, further CFST monitoring is utilized to determine when the additional thresholds are exceeded. Further degradation is indicated by the occurrence of valid CFST Core Cooling RED, or Heat Sink RED. These conditions are indicative of a loss or potential loss of the heat sink for core cooling.
DEVIATION None REFERENCES NUMARC NESP-007, SG2 EOP-TRIP-1, Reactor Trip or Safety Injection EOP-CFST-1, Critical Safety Function Trees EOP-FRSM-1, Response to Nuclear Power Generation EOP-FRHS-1, Loss of Secondary Heat Sink EAL - 5.1.4 Rev. 00 Page 2 of 2
SGS EALIRAL Technic~l Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release * \6\
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 loriger *' ; **
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-CDE Dose of~ 6.8E-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 MODE-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. l.d.
EAL - 6.1.1.a Rev.00 Page 1 of 3
i--
1 SGS EALIRALTechnical 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.\\'.Qen 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 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> release duration. Therefore .05 mRem/hr*4 hours=
0.2mRem.
Prorating the 1500 mRem/yr criterion for the Thyroid-CDE 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.
. 1500mReml yr Thyro1d-CDE MEA Dose Rate= ( )(2)(.5)= 0.17 mRem/hr 8766hr I yr The Thyroid-CDE Dose is based on a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> release durati.on. Therefore 0.17 mRem/hr*4 hours
= 0.68 mRem.
DEVIATION None EAL - 6.1. l.a Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, AUl.4 Off-Site Dose Cafoulation 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. I .a Rev. 00 Page 3 of 3
SGS EALIRALTechnical 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 Exceeds 2 Times the Radiological Technical Specifications for 60 minutes or longer EAL Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS
.05 mRem/hr above normal background Release is ongoing for :::, 60 minutes MODE-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 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.
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
SGS E.AL!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 mRern/yr cri.teJion 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.
- .. * *; *
- c
- 500mRem/yr * * ***
- Protected Area Boundary Dose Rate= ( )(2)(.5)= 0.57 mRem/hr
. _ .* . 8766hr I yr This is rounded to .05 mRem/hr DEVI.a.A TION **
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
SGS EALIRALTechnical 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 IOCFR20, Appendix B limits for 60 minutes or longer EAL Gaseous etlluent release sample analysis on EITHER one of the following indicates a concentration of:
- ~2.56E-03 µ.Ci/cc Total Noble Gas
- ~3._71E~08 µ.Ci/cc I"'.1~1 AND Dose Assessment results NOT available Release is ongoing for~ 60 minutes MODE-All BASIS A sample analysis of the release from all vent paths in excess of 2 times I OCFR20, Appendix B limits that continues for 60 minutes or longer represent an uncontrolled situation and hence a potential degradation in the level of safety. The EAL thresholds are based on 2 times I OCFR20, Appendix B limits Noble Gas and Iodine release rates limits.
The final integrated dose is very low and is not the primary concern; rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. It is not intended that the release be averaged over I hour, but exceed 2 times IOCFR20, Appendix B limit for I hour. 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 greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeded the conditions on the applicable permit.
EAL - 6.1.1.c Rev.00 Page 1 of 3
SGS EALIRALTechnical Basis Barrier Analysis NIA ;
ESCALATION CRITERIA Emergency Classification will escaiate to an Alert when the effluent release concentration increase to 200 times 10CFR20, Appendix B limit.
- DISCUSSION ..
Refer to.Basis Section for EAL 6.1.1.d for the 10CFR20, Appendix B Noble Gas release rate calculation.
10CFR20, Appendix B Thyroid Committed Dose release rate is calculated in the foll~~ing.
manner:
µCi/sec = 50mRem/year * (Allocation Factor)
(ODCM x(Q) * {ODCM THY DRCF)
WHERE: µCi/sec = 10CFR20, Appendix B Thyroid Committed Dose Release Rate .
50 mRem/year = 10CFR20, Appendix B thyroid Committed Dose limit ODCM x!Q = Salem specific dispersion factor at the Site Boundary in sec/m3 (2.20E-06 sec/m3)
ODCM DRCF THY = is the most limiting potential pathway (inhalation, child, Thyroid 1-131) dose rate conversion factor in 3
mRemJyear/µCi/m (1.62E+07 mRem/year/µCi/m 3)
- Allocation Factor= 5.00E-01
µCi/sec= 50 mRem/ year* (5.00E-01)
(2.20E-06 sec/m3) * (l .62E+07 mRem/yr/µCi/ m3)
µCi/sec= 7.0lE-01 7.0lE-OlµCi/sec
- 2 = 1.40 µCi/sec 1.40 µCi/sec= 2 times the 10CFR20, Appendix B Release Rate for Thyroid Committed Dose Calculation of the threshold sample concentrations are as follows:
. 9.68£ + 04µCi I sec Noble Gas Sample Concentration= 2.56E-03 µCi/cc 472x80000cfin 1.40E + OOµCi I sec . .
I-131 Sample Concentration= - 3.71E-08 µCi/cc 472x80000cfin EAL - 6. l. l.c Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis Where: 472 = conversion factor (28,317 cc/ft3 x 1 rnin./60 sec.)
80000 cfin =Plant Vent Flow (normal)
The noble gas release rate of9.68E+04 µCi/sec is obtained by multiplying the 10CFR20, Appendix B release rate of 4.84E+04 µCi/sec times 2.
The iodine release rate of 1.40E+OO µCi/sec is 9btain~d by multiplying the 10CFR20, Appendix B release rate of7.00E-01'1Ci/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 ()DCM: 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 fo'determine the classification, if any, due to the potential 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, AUl.2, AUl.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
SGS EALIRALTechnical 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 for 60 minutes or longer EAL Valid Plant Vent Effiuent Alarm Release Rate EXCEEDS 9.68E+04 µCi/sec Total Noble Gas AND Dose Assessment results NOT available Release is ongoing for ::::, 60 minutes MODE-All BASIS Valid High alarm and effiuent release rate Vlllues exceeding the EAL threshold, can result from a Gaseous Radiological Release in excess of 2 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 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.
Valid is defined as the High alarm actuating specifically due to a Gaseous Release exceeding I 0 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 EAL - 6.1.1.d Rev. 00 Page 1 of 3
SGS EALIRALTechnical Basis 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 evaluating 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.
The Total Noble Gas monitored Release Rate can be obtained from SPDS or in accordance with Sl.OP-AB:RAD-OOOl(Q) or S2.0P-AB.RAD-000l(Q) Abnormal Radiation.
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, Wis* 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 escalate to an Alert when the effluent release concentration increases to 200 times the 10CFR20, Appendix B limits.
DISCUSSION Release rate threshold for this EAL is obtained by multiplying the 10CFR20, Appendix B release rate for Noble Gas of 4.84E+04 µCi/sec times 2. This EAL does not include Iodine Release Rates, since the Plant Vent does not have an Iodine detector.
10CFR20, Appendix B Calculation for Noble Gas (lOOmRem/ year)*(Allocation Factor) u Ci/s econd = _;,,_-----=~---'-___.:,_-----~
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/uCi/m3.
EAL - 6.1.1.d Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis 3
ODCM XJQ = 2.20E-06 sec/m ODCM DRCF = 4. 70E+02 mRem/yr/uCi/m3_
Allocation Factor= 5.00E-01 4.84E~04- uC~Second = (lOO mRem 1 yr)* (5.00E -Ol) 3 3
.. _ '" (2.20£ - 06 sec/ m ) * ( 4.70£ + 02mRem I yr I µCi/ m )
4.84E+04 uCi/Second *2 =EAL value.
9.68E+04 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 OPCM. 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 meteorologicaLconditions and the isotopic mixture. EAL 6.1.1.d 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 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, AUl.1, AUl.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 - 6.1.1.d Rev. 00 Page 3 of 3
SGS EAL!RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.a IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 times Radiological Technical Specifications for 15 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 ~ 2.0E+Ol mRem
- Thyroid-CDE Dose ~ 6.SE+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 MODE-All BASIS Dose Assessment at or beyond the MEA exceeding the EAL threshold , can result from a Gaseous Radiological Release in.excess of200 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 3
SGS 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 escalated to a Site Area Emergency when the effiuent release concentration increase to a level that would cause a l 00 mRem dose at the Protective 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.
- 500mRem/ yr TEDE 4-Day MEA Dose Rate= ( )(200(.5) = 5.7 mRem/hr
- 8166hrlyr This is rounded to 5.0 mRem/hr.
The TEDE 4-day Dose is based on a default (assumed) 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> release duration. Therefore 5_0 mRem/hr x 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> = 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.
. I500mRem/ yr Thyr01d-CDE MEA Dose Rate= ( )(200)(.5) = 17 mRem/hr 8166hr I yr The Thyroid-COE Dose is based on a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> release duration. Therefore 17 mRem/hr x 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
=68mRem.
DEVIATION None EAL - 6.1 _2.a Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, AAI .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
SGS EALIRALTechnical 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 MODE-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 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.
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
. SGS EALIRALTechnical Basis ESCALATION CRITERIA Emerg*ency 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 rnRem/yr criterion for: time (8766 hr/yr); the 200 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of0.5 (50% per site), the associated site boundary dose rate would be 5.*7 inRem/hr.
500mReml yr
- Protected Area Boundary Dose Rate= ( )(200)(.5) = 5.7 mRem/hr 8766hr I yr This is rounded to 5 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
SGS EAL!RALTechnical Basis 6.0 *Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.c 1c* Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 times the 10CFR20, Appendix B limits for 30 minutes or longer.
- , .. ,-,* .
- r . ..
- '.I :. . . . .
Gaseous effiuent release sample analysis* on EITHER one of the following indicates a concentration of:
- ::::. 2.56E-01 µCi/cc Total Noble Gas
- ::::. 3.71E-06µCi/cc1-131 Dose Assessment results NOT available Release is ongoing for 2:: 30 minutes MODE-All BASIS Total gaseous effiuent sample analysis exceeding the EAL threshold for the Plant Vent, 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. 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. 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 10CFR20, Appendix B limit for 30 minutes or longer. Fu~her, 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.
EAL - 6.1.2.c Rev. 00 Page 1 of 3
SGS EALIRALTechnical Basis 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 Dose. of 500 mRem for 1-131 at the Protected Area Boundary.
DISCUSSION Refer to Basis Section for EAL 6.1.2.d for the 10CFR20, Appendix B Noble Gas release rate calculation or Basis Section for EAL 6.1.1.c for the 10CFR20, Appendix B Thyroid Committed Dose Release Rate Calculation.
Calculation of the threshold sample concentrations are as follows:
. 9.68£ + 06µCi I sec .
Noble Gas Sample Concentratwn = = 2.56E-01 µCi/cc 472x80000cfin
- * . 1.40E + 02µCi I sec .
I-131 Sample Concentratwn = = 3. 71E-06 µCi/cc 472x80000cfin Where: 472 =conversion factor (28,317 cc/ft3 x 1 min./60 sec.)
80000 cfin =Plant Vent Flow (normal)
The noble gas release rate of 9.68E+06 µCi/sec is obtained by multiplying the 10CFR20, Appendix B release rate of 4.84E+04 µCi/sec times 200. The Iodine release rate of l .40E+02
µCi/sec is obtained by multiplying the 10CFR20, Appendix B release rate of7.00E-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 an General Emergency classification or not meet the Alert classification, depending on the meteorological conditions and the 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 of this "default" EAL.
Two hundred 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.
EAL - 6.1.2.c Rev. 00 Page 2 of 3
SGS EAL!RALTechnical Basis 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.2,. AAl. l, 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 L
SGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences
. 6.1 Gaseous Effluent Release ALERT - 6.1.2.d ~- -- --- -
IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 times IOCFR20, Appendix B Limit for 30 minutes odonger EAL
. ...... .. - _-I .*
Valid Plant Vent Effluent Alarm Release rate EXCEEDS 9.68E+06 µ.Ci/sec Total Noble Gas Dose Assessment results NOT available Release is ongoing for::::, 30 minutes MODE-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 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. 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.
EAL - 6.1.2.d Rev.00 Page 1 of 3
SGS EAtlRALTechnical Basis The EAL value for Total PlantVerit release rate was determined using default X/Q values from the ODCM which provides a less accurate method of evaluating 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 meteor()logical conditions, plant vent flpws 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. L2.a for classificati_on.
The Total.noble gas monitor~d ReleaseRate can be obtained from SPDS or in accordance with Sl.OP-AB.RAD-OOI(Q). ~r S2.0P-AB.RAD-OOI(Q), Abnormal Radiation.
It is not intended thanhe release be-averaged-over 30 minutes, but exceed 200 times 10CFR20, 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 100 mRem dose at the Protected Area Boundary.
DISCUSSION The release rate thresholds for this EAL are obtained by multiplying the 10CFR20, Appendix B Limit release rate of 4.84E+04 µCi/sec for Noble Gases times 200. This EAL does not include Iodine Release Rates, since the Plant Vent does not have an Iodine detector.
10CFR20, Appendix B Calculation for Noble Gas uCi/Second = 100 mRem/year * (Allocation Factor)
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 SO mRem/year.
ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3.
EAL - 6.1.2.d Rev. 00 Page 2 of 3 I
I L
SGS EALIRALTechnical Basis ODCM DRCF = Site Specific (Salem or Hope Creek) dose rate conversion factor in mRem/year/uCi/m3.
ODCM X/Q = 2.20E-06 sec/m3
- ODCM DR,.CF ;== 4. 70E+02 mRem/yr/uCifm3 Allocation Factor= 5.00E-01 4.84E+04 uCi/Second .=. (100 mRem/year~ * (5.00E-01) * * . - .
- , _. . . . (2.20E.:o6 sec/m ) !I< (4.70E+02 mRem/yr/uCi/m3) 4.84E+04 uCi/sec
- 200 =EAL value 9.68E+06 µCi/sec= EAL value.
DEVIATION -
The value for EAL 6.1.2d 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 Dosi 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 10CFR20, Appendix B limits of 100 mRem/year noble gas 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, AAI.1, AAl.4 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effiuents OP-AB.RAD-0001 NUMARC Draft White Paper, 7/25/94; 9110194 Technical Specification 3.11.2.1 EAL - 6.1.2.d Rev. 00 Page 3 of 3
SGS 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 indicat~s EITHER one of the following at the MEA or beyond .as calculated
.on the SSCL:
- J'EDE 4~Day Dose~ 1.0E+02 mRem
- Thyroid-CDEDose ~ 5.0E+02 mRem based on Plant Vent effluent sample analysis and NOT on a default Nobfo-Gas to Iodine Ratio MODE-All BASIS TEDE 4-Day Dose ~ 1.0E+02 mRem corresponds directly to the NUMARC dose of I 00 mRem.
Thyroid-CDE Dose ~ 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 or 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
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.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION .
The EAL values provide 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 ca_lculation due to the assumption that releases of this magnitude will be from one site.
The dose projection code assumes a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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-CDE dose, 500 mRem/hr
- 4 hr = 2000 mRem.
- DEVIATION None REFERENCES NUMARC NESP-007, ASl.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
SGS EAL{RA.LTechnic~ 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 odmminent Release of Gaseous Radioactivity Exceeds I 00 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 MODE-All BASIS An actual dose rate of 100 mRem/hr which is expected to continue for :::: 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.
EAL - 6.1.3.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION .
An actual dose of 100 mRem Total Effective Dose Equivalent (TEDE) is based on the 10CFR20 annual average population exposure limit. *Measured dose rates will betaken at the Protected Area Boundary and a 2: 15 minute release duration threshold will be applied to be conservative.
Unless otherwise indicated, the conversion from whole body dose to TEDE is 1: 1.
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
SGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Ga~eous Effluent Release.
SITE AREA EMERGENCY - 6.1.3.c 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 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.I-131 MODE-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-13 I field survey sample concentration threshold is based on I-131 dose conversion factors (DCFs) 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 Page 1 of 3
SGS EALIRALTechnical 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:
mRern/hr = (µCi/cc)(Dose Conversion Factor)
Then; 500mRem I hr .
1-131 Sample Concentration = ( ) = 3.85E-07 µCi/cc
- . * * . 1.30£ + 09mRem I µCi I cc I hr Where l .30E+09 mRern/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4 and includes the EPA breathing rate.
The Corrected Counts per Minute reading is calculated using the 1-131 Sample concentration, an9 facfors for using an RM-14 or E-140N with an HP260 probe.
Solving the folloWing equation for CCPM:
µCi/cc = CCPM .
(Detector EfficicncyXCollection EfficicncyXConvcnion Factor - DPM to 11Ci)(Volume - ft3 XConvcnion Factor - cc to ft3 )
Then; CCPM = (3.85E-07 µCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/µCi) *
(10 ft 3) (2.832E+04 cc/ft3) = 4.36E+02 CCPM "Where:
2.00E-03 = Detector Efficiency - CCPMIDPM 0.9 (or 90%) ~ Collection Efficiency 2.22E+06 = Conversion factor -DPMlµCi 10jt3 = Volume 2.832£+04 = Conversion factor - cc to ft3 CCPM= Corrected Counts per Minute using an RM-14 or E-1-ION with an HP260 probe.
DEVIATION None REFERENCES EAL - 6.1.3 .c Rev.00 Page 2 of 3
SGS EALIRALTechnical Basis 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. I/July 1987, Guidance on Offsite Emergency Radiation Measurement Systems, Phase- I Airborne Release SORC Summary 07/10/89 RPCS T_hyroid Dose Commitment.Factor Paper (NRP-94-0557), 11-22-94 ..
- EAL - 6.1.3.c Rev. 00 Page 3 of 3
SGS EALIRALTechnical Basis 6.0 Radiological . :ff.eleases/Oc-currences
. 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 (TEDEYor 500 mRem Thyroid-CDE Dose-for the actual or projected duration of the release-EAL Valid Plant Vent Effluent Alarm Total Plant Vent release rate EXCEEDS 1.7E+09 µ.Ci/sec Total Noble Gas Dose Assessment results NOT available Release is ongoing for ~ 15 minutes MODE-All BASIS Valid High alarm and effluent release rate values exceeding the EAL threshold, indicates a substantial Gaseous Radiological Release which could exceed the 10CFR20 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 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 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 6.1.3.a for classification.
EAL - 6.1.3.d Rev. 00 Page 1 of 3
SGS EAlJRALTechnical Basis The Total Noble Gas monitored Release Rate can be obtained from SPDS or in accordance with Sl.OP-AB.RAD-OOl(Q) or S2.0P-AB.RAD-00l(Q), Abnormal Radiation.
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 Emergericy Classification will escalate to a General Emergency when effluent release concentration increases to a level that would cause a 1000 mRem dose at the Protected Area Boundary DISCUSSION 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.
Iodine Release Rates for this EAL are excluded since the Plant Vent Radiation Monitoring System does not include an Iodine detector.
A release rate of 1. 7E+09 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: Plant Vent Release Rate: 80,000cfin 3
ODCM X/Q = 2.20E-06 sec/m 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.
EAL - 6.1.3.d Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis 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-001. The combination of using the rSAR l~9topic mi~ture and the EPA 400 dose Rate Conversion Factors calculate an accurate accident sdurce term. . . . . .
REFERENCES NUMARC NESP-007, ASI. l, ASl.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 FSAR Section 15 EPA 400-R-001, Manual.of Protective Action Guides and Protective Actions for Nuclear Incidents EAL - 6.1.3.d Rev. 00 Page 3 of 3
SGS EALIRALTechnical 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 Assessmentindicates EITHER one of the following at the :MEA or beyond as calculated on the SSCL:
- .. TEDE 4-Day Dose 2:: 1.0E+03 mRem
- Thyroid-COE Dose ~* 5.0E+03 mRem based on P.lant Vent effluent sample analysis and NOT on a default Noble Gas to Iodine Ratio -*
MODE-All BASIS The TEDE 4-Day Dose of 1000 mRem corresponds directly to the NUMARC dose of 1000 mRem which exceeds EPA Protective Action Guideline Criteria for a General Emergency. The Thyroid-COE Dose of 5000 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Barrier Analysis NIA ESCALATION CRITERIA NIA 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.
EAL - 6.1.4.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION None
- REFERENCES NUMARC NESP-007, AGI.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
SGS 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 mRemlhr Release is expected to continue for~ 15 minutes MODE-All BASIS An actual dose rate of 1000 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 1000 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 1 Rem whole body. This is consistent with the emergency class description for a General Emergency. A release rate equivalent to 1000 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 1: 1.
EAL - 6.1.4.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis DEVIATION.
None ,
REFERENCES NUMARC NESP-007, AGl.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents EAL - 6.1.4.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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-mRemTotal Effective.Dose Equivalent (TEDE) or 5000 mRem Thyroid-COE 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: *
- i 4.36E+03 CCPM
- *~ 3~85E-06 µ.C.ilcc I-131 MODE-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-13 1 field survey sample concentration threshold is based on 1-131 dose factors from EPA-400. The thresholds are based on a dose rate of 5000 mRem/hr Thyroid CDE for 1-131.
Barrier Analysis NIA ESCALATION CRITERIA NIA DISCUSSION The release sample concentration calculations are as follows.
The sample concentration is calculated using the I-131 Dose Factor from EPA-400:
EAL- 6.1.4.c Rev. 00 Page 1 of 3
SGS EAL!RALTechnical Basis Solving the following equation for µCi/cc:
mRem/hr = (µCi/cc)(Dose Conversion Factor)
Then; SOOOmRem I hr .
J-131 Sample Concen_tration = ( . ) = 3.85E-06 µCi/cc
- ** * *;. : 1.30E + 09mRem I µCz I cc I hr Where l .30E+09 mRem/µCi/cc/hr is the Dose Conversion Factor (DCF) from EPA-400, Table 5-4 and includes the EPA breathing rate.
The Corrected Counts per Minute (CCPM) reading is calculated using the 1-131 Sample concentration, and factors for using an RM-14 or E-l 40N with an HP260 probe.
Solving the following equation for CCPM:
µCi/cc = CCPM .
(Detector Efficic:ncyXCollcction £fficic:ncyXConvcnion Factor-DPM lo µCiXVolumc - ft3 XConvClllion Factor - cc lo ft3)
Then; CCPM = (3.85E-06 µCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/µCi) *
(10 ft:3) (2.832E+04 cc/ft3)= 4.36E+03 CCPM Where:
2.00E-03 = Detector Efficiency - CCPMIDPM 0.9 (or 90%) = Collection Efficiency 2.22E+06 = Conversion factor - DPMlµCi 10jt3 = Volume 2.832E+04 = Conversion factor - cc to jt3 CCPM= Corrected Counts per Minute using an RM-14 or E-1-ION with an HP260 probe.
DEVIATION None EAL - 6.1.4.c Rev. 00 Page 2 of 3
SGS EALfRALTechnical Basis REFERENCES NUMARC NESP-007, AGI.4 EPA 400-R-92-00I, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents FEMA REP-2, Rev. I/July I 987, 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); I 1-22-94.
EAL - 6.1.4.c Rev. 00 Page 3 of 3
SGS EAL/RALTechnical 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 mRem Total Effective Dose Equivalent (TEDE) or 5000 mRem Thyroid-CDE Dose for the actual or projected duration of the release EAL Valid Plant Vent Effluent Alarm Total Plant Vent release rate EXCEEDS 1.7E+l0 µ.Ci/sec Total Noble Gas Dose Assessment results NOT available Release is ongoing for ~ 15 minutes MODE-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 1000 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 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 6.1.4.a for classification. The Total Noble Gas monitor Release Rate can be obtained from SPDS or in accordance with Sl.OP-AB.RAD-OOI(Q) or S2.0P-AB.RAD-000l(Q), Abnormal EAL - 6.1.4.d Rev. 00 Page 1 of 3
SGS* EAL/RALTechnical Basis Radiation. 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
. *~ : .: .
':ESCALATION CRITERIA None
- DISCUSSION*
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. Iodine Release Rates for this EAL are excluded since the Plant Vent Radiation Monitoring System does not include an Iodine detector.
Arelease rate of 1.7E+10 µCi/sec was backcalculated from a TEDE Dose of lOOOmRem/hour at the Site MEA. The assumptions that went into this calculation were as follows:
Release Point: Plant Vent Release Rate: 80,000cfm 3
ODCM X/Q = 2.20E-06 sec/m 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 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.
EAL - 6. 1.4.d Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, ASI. l, ASl.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 FSAR Ssection 15 _ .** _.
EPA 400-R-001, Manual of Protective Action Guides and Protective,Actions for Nuclear Incidents --
- c . * * " * -
EAL - 6.1.4.d Rev. 00 Page 3 of 3
SGS EALIRALTechnical 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 Two Times the Radiological Technical Specifications for 60 minutes or longer EAL Valid Alarm from ANY one of the following RMS Channels:
- Containment Fan Coil Process (R13)
- Liquid Radwaste Disposal Process (R18)
- Steam Generator Blowdown Process (R19)
- Chemical Waste Basin Process (2R3 7)
Sample analysis of liquid effluent indicates concentration in excess of 2 times Tech. Spec. Limits Release continues for~ 60 minutes after the alarm occurs MODE-All BASIS Releases in excess of 2 times Technical Specifications that continue for,::: 60 minutes represent an uncontrolled situation and hence a potential degradation in the level of safety. The final integrated dose is very low and is not the primary concern. Rather it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive release authorization wasn't prepared or that exceeds the conditions on the radioactive release authorization (e.g. minimum dilution, alarm setpoints, etc.).
It is not intended that the release be averaged over 60 minutes, but exceed 2 times Technical Specifications limit 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.
EAL- 6.2. l Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis Unplanned is defined as any release for which radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to an Alert when Liquid Effiuent Release exceeds 200 times Technical Specification-limits. - --- -
DISCUSSION The radiation monitors selected for this EAL moilitor radioactivity before it is discharged into the -
Delaware River and warns site personnel of an excessive effiuent concentration of radioactivity (greater than Technical Specification limits) being released to the environment.
DEVIATION None REFERENCES NUMARC NESP-007, AUl .2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effiuents Technical Specifications 3.11.1.1 (Ul and U2)
EAL- 6.2.1 Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 6.0 Radiologieal Releases/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 * * ** *
- Containment Fan Coil Proces~ (R13)
- .Liquid Radwaste Disposal Pr~cess (RI 8) *
- Steam Generator Blowdown Process (RI 9)
- Chemical Waste Basin Process (2R37)
Sample analysis of liquid effluent indicates concentration in excess of 200 times Tech. Spec. Limits Release continues for::::, 15 minutes after the alarm occurs MODE-All BASIS Releases in excess of200 times Technical Specifications that continue for 2: I5 minutes represent an uncontrolled situation and hence an actual degradation in the level of safety. This event escalates the Unusual Event by a factor of I 00. The required release duration was reduced to 15 minutes in recognition of the increased severity of a release of this magnitude. The calculation called for in this EAL should also be conducted whenever a liquid release occurs for which a radioactive release authorization wasn't prepared or that exceeds the conditions on the radioactive release authorization (e.g. minimum dilution, alarm setpoints, etc.). Unplanned is defined as any release for which a radioactive permit was not prepared, or a release that exceeds the conditions on the applicable permit.
EAL-6.2.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis It is not intended that the ~elease be averaged over 15 minutes, but exceed 200 times Technical Specifications limit for 15 minutes or longer. Further, 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 NIA DISCUSSION The radiation monitors selected for this EAL monitor radioactivity before it is discharged into the Delaware River and warns site personnel of an excessive effluent concentration of radioactivity .
(greater than Technic_al Specification limits) being released to the environment.
DEVIATION None REFERENCES NUMARC NESP-007, AAl.2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effluents EAL- 6.2.2 Rev. 00 Page 2 of 2
SGS _EAL/RALTechnical Basis
- . 6.0 Radiological Releases/Occurrences 6.3- In-Plant
. - Radiation Occurrences UNUSUAL EVENT - 6.3.1 IC *Unplanned Increase in Plant Radiatio~ .
EAL Unplanned rise in radiation levels inside the Protected Area~ 1000 time~ normal a_s i~dicated by EITHER one of the following:
- , Permanent or portable Area Radiation Monitors
- General Area Radiological Survey MODE-All.
BASIS An Unplanned rise in radiation levels within the Protected Area by a factor of 1000 times over normal represents 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 rising 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 This event will be escalated 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. RMS strip charts, RMS computer and/or SPDS can be used to confirm these values.
EAL- 6.3.1 Rev. 00 Page 1 of 2
I --
SGS EAL/RALTechnical Basis DEVIATION NUMARC IC AU2 includes unexpected increas~s in Airborne concentration in addition to plant radiation. The corresponding* Salem 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 6r. Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the ~orresponding IC was overlooked.
REFERENCES.
- 1*1, .....
NUMARC NESP-007, AU2.4 EAL- 6.3. l Rev. 00 Page 2 of 2
SGS 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 Rate >2000 mRem/hr in any area of the plant which requires ACCESS to maintain plant safety functions (excluding the Control Room or CAS) as indicated by EITHER one of the following:
- Permanent or portable Area Radiation Monitors
- e General Area Radiological Survey MODE-All BASIS The term "unplanned" is defined as those events which are not associated with pre-planned evolutions such that radiation levels are rising for reasons which cannot be immediately explained.
The EAL addresses radiation. levels which would impede operation of systems required to maintain safe operations or to establish or maintain Cold Shutdown. Radiation levels could be indicated by ARM or radiological survey.
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. The 2000 mRem/hr is not intended to be above the pre-existing background, but includes the pre-existing background. The Dose Rate of 2000 mRem/hr was chosen as a threshold based upon NAP-24 Administrative Dose Limits and Extension criteria which has Senior Radiation Protection Supervisor approval required prior to exceeding 2000 mRem/yr.
Barrier Analysis NIA EAL - 6.3.2.a Rev. 00 Page 1 of 2
SGS EAL/RALTechnical 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 to determine areas that contain systems that must 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 require HP coverage and continuous update of changing radiological conditions.
Areas of the plant which require access following an accident to maintain plant safety functions include, but are not limited to:
Area_s_ for ~emote Shutdown Core Residual Heat Re~oval system ~eas_.
CCW Pump Room Corridor next to the Spent Fuel Pit HX Room CCWHXRoom Electrical Control Center 4KV Switchgear Room Boric Acid Evaporator Room Unit 1 Diesel Generator Compartment Boric Acid Evaporator Room Unit 2 Diesel Generator Control Room Aux Feedwater Pump & Valve Area Diesel Oil Supply Tank CompartmentRadwaste Control Center Electrical Relay and Switchgear Room 100 ft Chiller Area DEVIATION None REFERENCES NUMARC NESP-007, AA3.2 NC.NA-AP.ZZ-0024(Q)- Radiation Protection Program S-C-VAR-MDC-1518 Rev 0, Draft EAL- 6.3.2.a Rev. 00 Page 2 of 2
SGS EAI'.JRAL Technical 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 systems required to maintain safe operations or to establish or maintain cold shutdown EAL Unplanned radiation levels> 15 mRem/hr in EITHER one of the following:
- The Control Room
- The Security Central Alarm Station (CAS)
MODE-All BASIS The term "unplanned" is defined as those events which are not associated with a pre-planned evolutions such that radiation levels are rising for reasons which cannot be immediately explained.
The EAL addresses radiation levels which would jeopardize continuous occupancy of the Control Room or Security CAS. Radiation levels could be indicated by ARM or radiological survey. 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 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.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to a Site Area Emergency when loss of control of radioactive materials causes significant off-site doses.
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 EAL - 6.3.2.b Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis General Design Criteria 19, which addresses continuous occupancy of the Control Room for 30 days after an accident. Additionally, since the Control Room is shielded, this radiation level
- represents a serious loss of control of radioactive material.
The Security Secondary Alarm Station (SAS) was excluded because it is fully redundant to the Security CAS. For a radiological event, SAS would be evacuated, with all Security functions performed by the CAS.
Events which may require Control Room evacuation to establish or maintain Cold Shutdown will be classified per Section 8 EALs. -* .
DEVIATION None REFERENCES NUMARC NESP-007, AA3.l IOCFRSO EAL - 6.3.2.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis
- _: ~ .. . '. '.
- 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event UNUSUAL EVENT - 6.4~1.a IC Unplanned inci:ease in Plant R~diation EAL
. An. uncontrolled .level drop in the Refueling Cavity as indicated by EITHER one of the following:
- Visual observation
- R VLIS - Refueling Mode MODE-6 BASIS This EAL condition indicates a possible failure of the Refueling Cavity Seal or RHR. System that results in inventory loss from the Refueling Cavity when flooded. Coverage of these events is appropriate due to the potential for higher doses to plant staff. These events have 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. Classification as an Unusual Event is warranted as a precursor to a more serious event. Uncontrolled means that the level drop cannot be terminated, or level cannot be maintained by operator action.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to an Alert as a result ofuncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor.
DISCUSSION Design of the Refueling Cavity is such that a liner failure in these volumes is unlikely; however, should such a failure occur, it would come under this EAL. Ifuncovery of fuel elements occur or if there is indication of high radiation levels on the refuel floor then the event will be classified as an Alert.
EAL - 6.4.1.a Rev. 00 Page I of2
SGS EALIRALTechnical Basis Dt.iririg refueling operations the Reactor Vessel and Refuel Cavity* are flooded~ During fuel handling operations, the Fuel Transfer Tube will connect the Reactor Cavity and the Spent Fuel Pool. An unexplained lowering of Refuel Cavity level or ,Spent Fuel Pool level can be an
- '.::: : inr:lication that these volumes are draining. A drop in Reactor Cavity apd Spent Fuel Pool level may result in a Spent Fuel Pool low level alarm. This alarm would pe validated by visual observation oflowering level in the Refuel Cavity/Spent Fuel Pool.
- ':" 1 ;., t , **
DEVIATION
~ -:.'.;, ~ Nl.JMAR.c states that this EAL will be applicable hi ali modes, cf operation. I~ m"odes* other. than Mode 6 the Reactor Vessel head will be fully tensioned and there will be no interconnection
- .1*
- between the Refueling Cavity and the Spent Fuel Pool. In other modes, a loss of Reactor Vessel inventory is addressed in Section 3. Uncontrolled loss of water level in the Spent Fuel Pool, however, is classified under EAL 6.4.1.b in all modes ofoperation.
REFERENCES NUMARC NESP-007, AU2. l
.. OP~AR.ZZ-OQ03(Q) OHA-C35 OP-AB.FUEL-0002(Q)
EAL - 6.4.1.a Rev. 00 Page 2 of2
SGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event UNUSUAL EVENT - 6.4.1.b IC *Unplanned increase in Plant Radiation EAL Valid SFP Low Level alarm - OBA C-35 Visual observation of an uncontrolled level drop in the Spent Fuel Pool MODE-All BASIS These EAL conditions indicate a possible failure of the Spent Fuel Pool Cooling System that results in inventory loss from the Spent Fuel Pool. This EAL also works in conjunction with the loss of Refueling Cavity EAL for Mode 6 operations, with the Spent Fuel Pool and Refueling Cavity connected via the Fuel Transfer Canal.
Coverage of this event is appropriate due to the potential for higher doses to plant staff 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. Classification as an Unusual Event is warranted as a precursor to a more serious event. Uncontrolled means that the level drop cannot be terminated, or level cannot be maintained by operator action.
Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert as a result of uncovery of a irradiated fuel as indicated by high radiation levels in the Fuel Handling Building.
EAL - 6.4.1.b Rev. 00 Page 1 of2
SGS EALIRALTechnical Basis DISCUSSION Design of the Spent Fuel Pool (SFP) is such that a liner failure in this volume is unlikely; however, should such a failure occur, it would be classified under this EAL. Lowering of water level in the SFP to below the level of the spent fuel bundles may result in a rise in the airborne contamination
- 1evel in the Fuel Handling Building. Ifuncovery of fuel elements occurs or if there is indication of high radiation levels in the Fuel Handling Building, then the event will be classified as an Alert.
This alarm would be validated by visual observation of lowering level in the Spent Fuel Pool. The added requirement for an uncontrolled drop in SFP level with a low level alarm is included to allow normal makeup to recover level for minor level deviations due to evaporation losses, et~.
DEVIATION None REFERENCES NUMARC NESP-007, AU2.2 OP-AR.ZZ-0003(Q) OHA-C35 OP-AB.FUEL-0002(Q)
EAL - 6.4.1.b Rev.00 Page 2 of2
SGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event ALERT - 6.4.2.a IC Major Damage to Irradiated Fuel or Loss of Water Level that has or will result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Major Damage to Irradiated Fuel reported in the Fuel Handling Bldg.
Valid High Alarm is received on EITHER one of the following RMS channels:
- RS
- R32A Valid High Alarm received from EITHER one of the following RMS channels:
- R41
- R45 MODE-All BASIS Major Damage to an irradiated fuel bundle that results in a High Fuel Handling Building Radiation Monitor alarm coincident with a Plant Vent Effluent Process Radiation Monitor alarm warrants declaration of an Alert, due to the potential for an 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 higher radiation levels due to shine, as a result of lowered shielding, but do not involve a release of fission products should not be classified under this EAL, but should be classified EAL 6.4.2.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, such that the magnitude of the damage specifically results in actuation of an Area Radiation Alarm. Valid is defined as the High alarm occurring as a result of the damage to the irradiated fuel bundle.
EAL - 6.4.2.a Rev. 00 Page I of2
SGS 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 Fuel Handling Building Area Monitors provide an early warning of developing problems which may be related to a damaged fuel bundle. The Plant Vent Exhaust Rad Monitors are Process Monitors and are designed to detect a release of Fission Products. Hence, they are included as part of the EAL threshold, to confirm the magnitude of damage to an irradiated fuel bundle.
DEVIATION None REFERENCES NUMARC NESP-007, AA2. l OP-AR.ZZ-0003(Q) OHA-C35 OP-AB.FUEL-0002(Q)
NUREG/CR-4982 NRC Information Notice no. 90-08 IOCFR50 EAL - 6.4.2.a Rev. 00 Page 2 of2
SGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event ALERT - 6.4.2.b IC Major Damage to Irradiated Fuel or Loss of Water Level that has or will result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel EAL Major Damage to Irradiated Fuel reported in the Containment Valid High Alarm received on ANY one of the following RMS channels:
- R2
- RIOA
- RIOB Valid High Alarm received from ANY one of the following RMS channels:
- RIIA
- Rl2A
- Rl2B MODE-All BASIS Major Damage to an irradiated fuel bundle that result in a High Containment Area Radiation Monitor alarm coincident with a Containment Process Radiation Monitors alarm warrants declaration of an Alert, due to the potential for an offsite release exceeding the Technical Specification limit. The intent of this EAL is to classify those events that result in the potential release of fission products from an irradiated Fuel Bundle, due to physical damage. Events that result in higher radiation levels due to shine, as a result of lowered shielding, but do not involve a release of fission products should not be classified under this EAL, but should be classified EAL 6.4.2.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, such that the magnitude of the EAL - 6.4.2.b Rev. 00 Page I of2
SGS EAL/RALTechnical Basis damage specifically results in actuation of an Area Radiation Alarm. Valid is defined as the High alarm occurring as a result of the damage to the irradiated fuel bundle.
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 Containment Area Monitors provide an early warning of developing problems which may be related to a damaged fuel bundle. The Containment Rad Monitors are Process Monitors and are designed to detect a release of Fission Products. Hence, they are included as part of the EAL threshold, to confirm the magnitude of damage to an irradiated fuel bundle.
DEVIATION None REFERENCES NUMARC NESP-007, AA2. l OP-AR.ZZ-0003{Q) OHA-C35 OP-AB.FUEL-0002{Q)
NUREG/CR-4982 NRC Information Notice no. 90-08 EPA 400-R-92-001, Manual of Protective Action Guide and Protective Actions for Nuclear Incidents EAL - 6.4.2.b Rev. 00 Page 2of2
SGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event ALERT - 6.4.2.c/6.4.2.d IC Major Damage to Irradiated Fuel or Loss of Water Level that has or will result in the Uncovering oflrradiated 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 2: 2000 mRem/hr
- R2
- RS
- R9
- R32A
- Visual observation oflrradiated Fuel uncovered MODE-All BASIS This EAL indicates a possible failure of the Refueling Cavity Seal, RHR System, or Spent Fuel Pool Cooling System that results in inventory loss from the Refueling Cavity when flooded or the Spent Fuel Pool. Design of the Refueling Cavity and Spent Fuel Pool is such that a liner failure in these volumes is unlikely; however, should such a failure occur, it would come under this EAL.
Lowering of water level in the Spent Fuel Pool to such a value as to cause Dose Rates to increase to this value will result in evacuation of the local areas. Uncovery of irradiated fuel elements can lead to their fuel clad failure due to loss of cooling.
The term "unplanned" is defined as those events which are not associated with a pre-planned evolutions such that radiation levels are increasing for reasons which cannot be immediately explained. The EAL addresses radiation levels which would impede operation of systems required to continue efforts to stop the loss of Refueling water level. Radiation levels could be indicated by ARM or radiological survey. The Dose Rate of2000 mRem/hr was chosen as a threshold based upon NAP-24 Administrative Dose Limits and Extension criteria which has Senior Radiation Protection Supervisor approval required prior to exceeding 2000 mRem/yr.
This value is low enough to ensure classification of an Alert before personnel access is severely
. EAL - 6.4.2.c/ 6.4.2.d Rev. 00 Page I of2
SGS EAL/RALTechnical Basis hampered and high enough to allow any unplanned rise in ~ormal radiation level, by a factor of 1000, to be classified as an Unusual Event per EAL 6.3.1.
Visual observation of irradiated fuel uncovered will result in onsite dose levels changing significantly.
The Area Radiation Monitors included in this EAL are:
- R2 Containment, General Area Low
- RS Fuel Handling Building Area Fuel Pool
- R9 Fuel Handling Building Fuel Storage Area
- R32A Spent Fuel Handling Crane, Area Monitor Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to a Site Area Emergency when loss of control of radioactive materials causes significant offsite doses.
DISCUSSION It is understood that a drop in Refueling Cavity water level will cause Dose Rates to rise due to the uncovery of irradiated Reactor components other than a spent fuel assembly. However, Dose Rates in excess of 2 Rem/hr indicate a loss of level such that recovery options may be limited and thus an Alert declaration is warranted.
DEVIATION None REFERENCES NUMARC :NESP-007, AA2.3 and AA2.4 OP-AR.ZZ-0003(Q) OHA-C35 OP-AB.FUEL-0002(Q)
NUREG/CR-4982 NRC Information Notice no. 90-08 EAL - 6.4.2.c/ 6.4.2.d Rev. 00 Page 2 of2
SGS EAL/RALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities UNUSUAL EVENT- 7.1.1 IC Loss of All Offsite Power to Vital Buses for Greater Than 15 Minutes EAL Loss of 13KV Offsite Power Availability to ALL 4KV Vital Buses as evidenced by a loss offunction of
- BOTH Station Power Transformers 13 (23) and 14 (24)
> 15 minutes have elapsed MODE-All BASIS Loss of Station Power Transformers 13(23) and 14(24) will result in a loss of offsite power to all 4KV Vital Busses for Unit 1 (Unit 2). The intent of this EAL is to identify a loss of off-site 500 KV or 13 KV power availability such that the 13(23) and 14(24) Station Power Transformers are unable to provide power to the 4KV Vital Buses.
Events which result in all available 4KV Vital Buses being supplied by their respective Diesel Generator with off-site power available should not be classified under this EAL (e.g.; all available 4KV vital buses in blackout loading during shutdown conditions due to inadvertent SEC Mode 2 "Blackout" loading with off-site power avaiiable).
Prolonged loss of AC power reduces redundancy and potentially degrades the level of safety by increasing plant vulnerability to a complete loss of AC power. 15 minutes was chosen to exclude transient or momentary power losses. Resetting of the 15 minute "clock" should not occur until a reliable source of power has been restored to the vital bus.
The term Power Availability to ALL 4KV Vital Busses is defined as the ability to restore off-site power to the Vital Bus (not just an open breaker which can reenergize the vital bus from an offsite source). The term loss of function is defined as the inability of these transformers to provide reliable offsite power due to transformer failure or other problems associated with equipment/power lines normally available.
EAL - 7.1. l Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis Barrier Analysis None ESCALATION CRITERIA This event will be escalated to the Alert classification level on loss of power to two 4KV Vital Buses.
DISCUSSION All Emergency Operating Procedures, except EOP-LOPA-1, are written assuming that at least two 4KV Vital Busses have power available. Two 4KV Vital Buses are required to ensure that at least one full train of ESF equipment is available. In Modes 1 and 2, a loss of all offsite power will result in or require a reactor trip and transition into the EOP Network. For Modes 3 and 4, OP-AB.LOOP procedures provide additional guidance.
DEVIATION None REFERENCES NUMARC NESP-007, SUI EOP-TRIP-1 EOP-LOPA-1 OP-AB.LOOP-OOOl(Q)
OP-AB.LOOP-0002(Q)
OP-AB.4KV-OOO 1(Q)
OP-AB.4KV-0002(Q)
OP-AB.4KV-0003(Q)
SGS 1(2) Technical Specifications Section 3/4.8 EAL - 7.1.1 Rev. 00 Page 2 of 2
~----------
SGS EAL/RALTechnical 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 station blackout EAL Loss of 4KV Vital Bus Power Sources (Offsite and Onsite) which results in the availability of only one 4KV Vital Bus Power Source (Offsite or Onsite)
> 15 minutes have elapsed MODE - 1, 2, 3, 4 BASIS The condition indicated by this EAL is the degradation of offsite and onsite power systems supply to the 4KV Vital Buses, with two separate concerns. First, this EAL declares an Alert for conditions such that any additional, single power source failure would result in a loss of power to ALL 4KV Vital Buses. Second, an Alert would also be declared for< 2 4KV Vital Buses energized to be consistent with EOP-LOP A-1 *entry conditions. At least 2 4KV Vital Buses are required to ensure one full train of ESF equipment is available for plant control. These conditions reduce redundancy and potentially degrade the level of safety by increasing plant vulnerability to a complete loss of Vital AC power. Availability means that the power source can be aligned to provide power to the bus within 15 minutes or is currently supplying power to at least one Vital Bus. Fifteen ( 15) minutes was chosen to exclude transient or momentary power losses. Resetting of the 15 minute "clock" should not occur until a reliable source of power has been restored to the vital bus.
Barrier Analysis None EAL - 7. I .2.a Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis ESCALATION CRITERIA This event will be escalated to the Site Area Emergency classification level on loss of power to all
- 4KV Vital Buses for > 15 minutes.
DISCUSSION The intent of this EAL is to classify events strictly as they relate to 4KV Vital Bus power availability. For the purposes of the EAL, availability of Diesel Generators that have not been challenged to start during degradation of AC power sources to the 4KV Vital Buses should be based on meeting Technical Specification action requirements for loss of offsite AC power sources. There are two separate conditions addressed by this EAL.
The first condition is directly related to the Initiating Condition, and is precautionary in classifying the event as an Alert if a single failure of one power source could result in a total loss of all 4KV Vital power. Should such a loss actually occur, it would result in classification at the Site Area Emergency Level after 15 minutes if no other power sources are available. Examples of this condition are:
- 1) Failure of the 13(23) Station Power Transformer with all Diesel Generators inoperable; or
- 2) loss of all offsite power with a failure of two Diesel Generators (results in only one 4KV Vital Bus energized by its associated Diesel Generator).
The second condition is unique to Salem Generating Station due to the three 4KV Vital Bus vs.
two trains ofESF equipment arrangement. Two energized 4KV Vital Buses are required to ensure the availability of one full train ofESF equipment. This threshold is consistent with EOP-LOPA-1 entry conditions used inthe EOP Network.
DEVIATION None REFERENCES NUMARC NESP-007, SAS EOP-'TRIP-1 EOP-LOPA-1 OP-AB.LOOP-0001 (Q)
OP-AB.LOOP-0002(Q)
OP-AB.4KV-0001(Q)
OP-AB.4KV-0002(Q)
OP-AB.4KV-0003(Q)
SGS 1(2) Technical Specifications Section 3/4:8 EAL - 7.1.2.a Rev. 00 Page 2 of 2
SGS 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 KV Vital Buses While the Plant is in Cold Shutdown , Refueling or Defueled Mode EAL Loss of power to ALL 4KV Vital Buses
> 15 minutes have elapsed MODE - 5, 6, Defueled BASIS Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Fan Coil Unit, Spent Fuel Pool Cooling and Service Water. When in cold shutdown, refueling, or defueled modes, this event can be classified as an Alert. This is because of the significantly reduced decay heat load with lower temperatures and pressures.
Fifteen ( 15) minutes was chosen to exclude transient or momentary power losses. Resetting of the 15 minute "clock" should not occur until a reliable source of power has been restored to the vital bus.
Barrier Analysis None ESCALATION CRITERIA Escalation to a Site Area Emergency would occur on Radiological Release (EAL Section 6.0), or on the long term inability to remove Decay Heat (EAL Section 8.0).
DISCUSSION In Modes 5, or 6, OP-AB.LOOP-OOOl(Q) provides guidance for maintaining plant control regardless of power remaining to the 4KV Vital Buses.
EAL - 7. 1.2.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis It is assumed that the plant will be maintained in a Cold Shutdown condition. If the plant is not able to be maintained in this mode, then escalation to Site Area Emergency would be appropriate based on Loss of Decay Heat Removal Capability EALs in Section 8.0 .
. DEVIATION None REFERENCES NUMARC NESP-007, SAl OP-AB.LOOP-0001 (Q)
OP-AB.4KV-0001(Q)
OP-AB.4KV-0002(Q)
OP-AB.4KV-0003(Q)
SGS 1(2) Technical Specifications Section 3/4.8 EAL - 7.1.2.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical 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 Vital AC Buses EAL Loss of power to All 4KV Vital Buses
> 15 minutes have elapsed MODE- 1, 2, 3, 4 BASIS Loss of power to Station Power Transformers 13 and 14 (23 and 24) will result in a loss of all offsite power to all 4KV Vital Buses for Unit 1 (Unit 2). With a failure of the Emergency Diesels to energize the 4KV Vital Buses, all plant safety system functions are compromised. Prolonged loss of AC power will cause core uncovery and loss of Containment integrity. The high potential decay heat loads in these modes warrants classification at the Site Area Emergency level. Fifteen minutes is chosen as a threshold to exclude transient or momentary power losses. Resetting of the 15 minute "clock" should not occur until a reliable source of power has been restored to the vital bus.
Barrier Analysis Prolonged loss of all AC power has the potential for causing a potential loss or loss of the Fission Product Barriers.
ESCALATION CRITERIA Escalation to General Emergency classification level will be via fission product barrier loss, or prolonged loss of offsite and onsite AC power.
EAL- 7.1.3 Rev. 00 Page 1 of 2
SGS EAtlRALTechnical Basis DISCUSSION All Emergency Operating Procedures except EOP-LOPA-1 are written assuming that at least two 4KV Vital Buses have power available. In Modes 1 and 2, a loss of all offsite power will result in
. or require a reactor trip. The threshold for this EAL is consistent with actions required by EOP-LOP A-1 to maintain the RCS Barrier, performing a rapid plant cooldown and depressurizing in order to minimize the potential of Reactor Coolant Pump seal failure, while continuing attempts to restore 4KV Vital Bus power. In Mode 3, operation within OP-AB.LOOP-0002(Q) is allowed without transition to EOP-TRIP-1 and EOP-LOPA-1. In Mode 4, OP-AB.LOOP-OOOI(Q) provide guidance for maintaining plant control regardless of the status of the 4KV Vital Buses.
DEVIATION None REFERENCES NUMARC NESP-007, SS 1 Station Blackout Coping Studies EOP-TRIP-1 EOP-LOPA-1 OP-AB.LOOP-0002(Q)
OP-AB.4KV-000l(Q)
OP-AB.4KV-0002(Q)
OP-AB.4KV-0003(Q)
SGS 1(2) Technical Specifications Section 3/4.8 EAL - 7.1.3 Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities GENERAL EMERGENCY - 7.1.4.a/7.1.4.b/7.1.4.c IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Vital
> 15 minutes have elapsed ANY one of the following:
- Restoration of Power to at least one 4KV Vital Bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is NOT likely
- CFST CORE COOLING RED PATH
- CFST HEAT SINK RED PATH MODE- 1, 2, 3, 4 BASIS Loss of all AC power compromises all plant safety systems requiring electric power. Prolonged loss of all AC power will lead to loss of Fuel Clad, RCS and Containment. Restoration of at least one 4KV Vital Bus within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is based on the station blackout coping analysis, and may still lead to core damage. Prudence in timely Protective Action Recommendation is necessary since core damage may occur even if AC power is restored.
CFST Core Cooling RED Path and Heat Sink RED Path provide indication of the loss or potential loss of fission product barriers. Because plant control strategies are limited with a prolonged loss of all AC power, these should be considered to indicate a loss of the fuel clad barrier, and a potential loss of the RCS or Primary Containment barriers. These threshold conditions are used to provide the Emergency Coordinator criteria for declaring a General Emergency based on degrading fission product.barriers.
EAL - 7.1.4.a/7. l.4.b/7. l .4.c Rev. 00 Page 1 of 3
SGS EAL!RALTechnical Basis Barrier Analysis Prolonged loss of all AC power has the potential for causing a potential loss or loss of the Fission Product Barriers.
ESCALATION CRITERIA NIA DISCUSSION This EAL is based on a station blackout occurring while the unit is in mode 1,2, 3 or 4 and power not being restored for > 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The status and availability of DC power may limit or prevent restoration activities. When prolonged powering of inverters and DC loads has occurred without AC power available for the battery chargers, DC voltage will degrade. This degradation of DC power may limit monitoring and assessment capabilities as instrumentation and control power may not be available. Since monitoring of overall plant conditions will be difficult with no AC power, CFST indiCations for
- determining barrier loss are used.
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:
- 1. 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 I 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 EAL - 7.1.4.a/7. l .4.b/7. l .4.c Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, SG 1 Station Blackout Coping Studies EOP-TRIP-1 EOP-LOPA-1 OP-AB.LOOP-0002(Q)
-OP-AB.4KV-000l(Q)
OP-AB.4KV-0002(Q)
OP-AB.4KV-0003(Q)
SGS 1(2) Technical Specifications Section 3/4.8 EAL - 7. 1.4. a/7. 1. 4. b/7. 1.4. c Rev. 00 Page 3 of 3
SGS EAL/RALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities UNUSUAL EVENT- 7.2.1.a IC Unplanned Loss of Required DC Power While the Unit is in Either Cold Shutdown or Refueling Mode for Greater Than 15 Minutes EAL Unplanned drop in Voltage to< 114 VDC on ALL 125VDC Vital buses
> 15 minutes have elapsed
- MODE-5, 6 BASIS A loss of all DC power compromises the ability to monitor and control plant functions. 125 volt DC system provides control power to decay heat removal systems, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the associated loads. If 125 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur.
15 minutes was chosen to exclude transient or momentary power losses. Although this EAL threshold is not met unless ALL 125 VDC is lost, EC judgment should be used to classify an event that results in loss of two of the three 125 VDC Vital buses ifthe loss causes an extensive loss of control of the plant and/or safety systems. Threshold values for bus voltage were derived from SC.MD-ST.125-0004(Q).
Barrier Analysis None ESCALATION CRITERIA This event would be escalated to an Alert based on Loss of Decay Heat Removal Capability.
EAL - 7.2. l.a Rev. 00 Page 1 of 2 L
SGS EAL/RALTechnical Basis DISCUSSION Two of the three 125 VDC buses are required operable in Modes 5 or 6 per Technical Specifications. This EAL addresses an unplanned loss of ALL 125 VDC buses such that Technical Specification requirements are not met. The minimum voltage value was selected based
- on the minimum allowable voltage (rounded to 114. 0 for consistency and readability on Control Room analog indications) required for DC bus operability as per SC.MD-ST.125-0004(Q).
Although continued operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed. Loss of DC power may result in the loss of control power and instrumentation associated with equipment necessary to maintain Cold Shutdown conditions.
DEVIATION None REFERENCES NUMARC NESP-007, SU7 OP-AR.ZZ-0002(Q)
SGS 1(2) Technical Specifications, 3/4.8 CBD DE-CB, 125-0018(Q)
SC.MD-ST.125-0004(Q)
EAL - 7.2.1.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities UNUSUAL EVENT- 7.2.1.b IC Unplanned Loss of Required DC Power While the Unit is in Either Cold Shutdown or Refueling Mode for Greater Than 15 Minutes EAL Unplanned drop in Voltage to< 25 VDC on ALL 28VDC Vital buses
> 15 minutes have elapsed Loss of control of Safety Related Equipment from the Control Room has been ~onfirmed MODE- 5, 6 BASIS A loss of all DC power compromises the ability to monitor and control plant functions. 28 volt DC system provides control power to provide for remote operation of switchgear, annunciators, vital instrument buses, communications to auxiliary control system relay cabinets for manual control ofESF equipment, non-safety related equipment, and 1RP4 Status Board indications. If 28 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions required to maintain safe plant conditions may not operate and core uncovery with subsequent reactor coolant system and primary containment failure might occur.
The requirement to have a confirmed loss of equipment control was added to ensure that classification will not be made if sufficient voltage is available to operate the required safety related equipment. The term loss of control is defined as the inability to manipulate the required piece of equipment. The term from the Control Room ensures that local manipulation is excluded from this EAL. The term confirmed is defined as evidence of a failure to operate such as the absence of a confirmatory push-button bezel light with associated changes in system parameters not observed (flow, pressure, etc.). Fifteen (15)minutes was chosen to exclude transient or momentary power losses. Threshold values for bus voltage were derived from SC.MD-ST.28D-0004(Q).
EAL - 7.2.1.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis Barrier Analysis None ESCALATION CRITERIA This event would be escalated to an Alert based on Loss of Decay Heat Removal Capability.
DISCUSSION One 28 VDC bus is required operable in Modes 5 or 6 per Technical Specifications. This EAL addresses an unplanned loss of ALL 28 VDC buses such that Technical Specification requirements are not met. The minimum voltage value was selected based on the minimum allowable voltage (rounded to 25.0 for consistency and readability on Control Room analog indications) required for DC bus operability as per SC.MD-ST.28D-0004(Q). Loss of DC power may result in the loss of control power and instrumentation associated with equipment necessary to maintain Cold Shutdown conditions.
DEVIATION Since Salem has a 28VDC system which is required to operate push-button controls in the Control Room, this EAL was added. There are only two 28VDC busses, and as such, a confirmation of the inability to operate safety related equipment was added to prevent inappropriate classification.
REFERENCES NUMARC NESP-007, SU7 SGS 1(2) Technical Specifications, 3/4.8 CBD DE-CB, 28D-0019(Q)
SC.MD-ST.28D-0004(Q)
EAL - 7.2.1.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities SITE AREA EMERGENCY - 7.2.3.a IC Loss of All Vital (IE) DC Power EAL Unplanned drop in Voltage to< 114 VDC on ALL 125VDC Vital buses
> 15 minutes have elapsed MODE- 1, 2, 3, 4 BASIS A loss of all DC power compromises the ability to monitor and control plant functions. 125 volt DC system 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 125 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions required to maintain safe plant conditions may not operate and core uncovery with subsequent Reactor Coolant System and Primary Containment failure might occur.
15 minutes was chosen to exclude transient or momentary power losses. Although this EAL threshold is not met unless ALL 125 VDC is lost, EC judgment should be used to classify an event that result in loss of two of the three 125 VDC Vital buses if the loss causes an extensive loss of control of the plant and/or safety systems. Threshold values for each individual bus voltage were derived from SC.MD-ST.125-0004(Q).
Barrier Analysis None EAL - 7.2.3.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis ESCALATION CRITERIA There is no direct escalation 'to a General Emergency. Escalation would be based on other EALs indicating Radiological Release (EAL Section 6.0) or loss of Fission Product Barriers (EAL Section 3.0).
DISCUSSION This EAL addresses plant conditions resulting in a loss of all 125VDC Vital power while the plant is in mode 1, 2, 3, or 4. The voltage selected was the minimum voltage on the bus based on the minimum allowable voltage required for DC Bus operability as per SC.MD-ST.125-0004(Q) ..
Although continued operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed.
DEVIATION None REFERENCES NUMARC NESP-007, SS3 OP-AR.ZZ-0002(Q)
SGS 1(2) Technical Specifications, 3/4.8 CBD DE-CB.125-0018(Q)
SC.MD-ST.125-0004(Q)
EAL - 7.2.3.a Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities SITE AREA EMERGENCY- 7.2.3.b IC Loss of All Vital (IE) DC Power EAL Unplanned drop in Voltage to <25 VDC on ALL 28VDC Vital buses
> 15 minutes have elapsed Loss of control of Safety Related Equipment from the Control Room has been confirmed MODE- 1, 2, 3, 4 BASIS A loss of all DC power compromises the ability to monitor and control plant functions. 28 volt DC system provides control power to provide for remote operation of switchgear, annunciators, vital instrument buses, communications to auxiliary control system relay cabinets for manual control ofESF equipment, non-safety related equipment, and 1RP4 Status Board indications. If 28 volt DC power is lost for an extended period of time (greater than 15 minutes) critical plant functions required to maintain safe plant conditions may not operate and core uncovery with subsequent Reactor Coolant System and Primary Containment failure might occur.
The requirement to have a confirmed loss of equipment control was added to ensure that classification will not be made if sufficient voltage is available to operate the required safety related equipment. The term loss of control is defined as the inability to manipulate the required piece of equipment. The term from the Control Room ensures that local manipulation is excluded from this EAL. The term confirmed is defined as evidence of a failure to operate such as the absence of a confirmatory push-button bezel light with associated changes in system parameters not observed (flow, pressure, etc.). Fifteen (15) minutes was chosen to exclude transient or momentary power losses. Threshold values for bus voltage were derived from SC.MD-ST.28D-0004(Q).
EAL - 7.2.3.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis Barrier Analysis None ESCALATION CRITERIA There is no direct escalation to a General Emergency. Escalation would be based on other EALs indicating Radiological Release (EAL Section 6.0) or loss of Fission Product Barriers (EAL Section 3.0).
DISCUSSION This EAL addresses plant conditions resulting in a loss of all 28VDC Vital power while the plant is in Mode 1, 2, 3, or 4. The voltage selected was the minimum voltage on the bus based on the minimum allowable voltage required for DC Bus operability as per SC.MD-ST.28D-0004(Q).
DEVIATION Since Salem has a 28VDC system which is required to operate push-button controls in the Control Room, this EAL was added. There are only two 28VDC busses, and as such, a ~
confirmation of the inability to operate safety related equipment was added to prevent inappropriate classification.
REFERENCES NUMARC NESP-007, SS3 OP-AR.ZZ-0002(Q)
SGS 1(2) Technical Specifications, 3/4.8 CBD DE-CB.125-0018(Q)
SC.MD-ST.125-0004(Q)
EAL - 7.2.3.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability ALERT - 8.1.2
\0\
IC Inability to Maintain the Plant in Cold Shutdown EAL An Unplanned, Complete loss of ALL systems providing Decay Heat Removal functions EITHER one of the following occur:
- RCS Temperature has risen to> 200°F (Excluding a < 15 minutes rise > 200°F with a heat removal function restored)
- An UNCONTROLLED temperature rise is RAPIDLY approaching 200°F (with NO heat removal functions restored)
MODE-5, 6 BASIS The intent of this EAL is to declare an Alert prior to boiling in the core when ALL RHR capability is lost in Cold Shutdown or Refueling. The specification of a 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 term "ALL systems providing Decay Heat Removal functions" is intended to represent a complete loss of functions providing core cooling during the Cold Shutdown and Refueling Modes including available injection pathways. The term "Unplanned" is included to preclude the declaration of an emergency for circumstances in which the RHR System is intentionally removed from service. This EAL allows actions taken in the appropriate OP-AB.RHR procedures to re-establish RHR Cooling or provide for alternate methods of decay heat removal, such as Hot Leg Injection, with the intent of maintaining RCS temperature below 200°F. For loss of "in-service" RHR. events with alternate cooling methods available, actions taken to provide for alternate DHR functions may require time to implement.
If the event results in RCS temperature momentarily (not to exceed 15 minutes) rising above 200°F with heat removal capability restored, Emergency Coordinator judgment will be required to EAL - 8.1.2 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis 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.
Barrier Analysis None ESCALATION CRITERIA This event would be escalated to a Site Area Emergency if RCS temperatures cannot be restored to below 350°F, or ifthe core becomes uncovered.
DISCUSSION Separate criteria was included in this EAL for the 200°F limit in order to recognize additional methods available to provide core cooling. A loss of Technical Specification components alone is not intended to be classified under this EAL. The same is true for momentary unplanned excursions above 200°F when an alternate cooling method is available and functioning to lower RCS temperature below 200°F, thus representing successful implementation of the loss ofRHR Abnormal Operating Procedure network. The EAL guidance related to uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from much lower than the Cold Shutdown temperature limit. With Core Exit Thermocouple indications available, this classification can be easily made in a timely manner. Wide range Hot or RHR System temperature indications are not considered accurate as they are dependent on RHR.
System flow. Reference to the Abnormal Procedures may be required for determining heatup rate when the CETs are disconnected for refueling operations or otherwise unavailable. Use of these curves provides sufficient detail to determine core heat up rate. This EAL satisfies the concerns of Generic Letter 88-17.
DEVIATION None REFERENCES NUMARC NESP-007, SA3 NUMARC Questions and Answers, June 1993, "System Malfunction Question #6b" OP-AB.RHR-OOOl(Q)
OP-AB.RHR-0002(Q)
Generic Letter 88-17 EAL - 8.1.2 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY- 8.1.3.a IC Complete Loss of Functions Needed to Achieve or Maintain the Plant in Hot Shutdown EAL An Unplanned, Complete loss of ALL systems providing Decay Heat Removal functions EITHER one of the following occur:
- RCS Temperature has risen to> 200°F (Excluding a < 15 minutes rise > 200°F with a heat removal function restored)
- An UNCONTROLLED temperature rise is RAPIDLY approaching 200°F (with NO heat removal functions restored)
Actions required by OP-AB.RHR have NOT maintained RCS temperature < 350°F MODE - 4 on RHR. Cooling, 5, 6 BASIS This EAL is a direct result of a loss of RHR event and takes advantage of the various RCS cooling options offered by the Abnormal Operating procedures for a loss ofRHR. capabilities.
Should this loss ofRHR cooling event result in an RCS heatup to >350 F, this EAL will allow classification based upon a significant loss of plant control and work in conjunction with the Fission Product Barrier Table or Radiological Releases/Occurrences EALs.
Barrier Analysis None EAL - 8.1.3.a Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis ESCALATION CRITERIA This event would be escalated to a General Emergency on loss of Fission Product Barriers or abnormal radiological releases.
DISCUSSION This EAL works in conjunction with EALs 8.1. 2 and 8.1. 3. d, depending upon the initial plant conditions. When in Modes 5 or 6 and RHR capability is lost (EAL 8.1.2), OP-AB.RHR-0001 and -0002 provide guidance on controlling the RCS temperature rise by various methods including injection or steaming of the Secondary plant. When a cooldown from Mode 3 into Mode 4 is required, EAL 8.1.3.d provides threshold values for a loss of Heat Sink event until RHR cooling can be established.
DEVIATION None REFERENCES NUMARC NESP-007, SS4 EOP-CFST-1 OP-AB.RHR-OOOI(Q)
OP-AB.RHR-0002(Q)
EAL - 8.1.3.a Rev 00 Page 2 of 2
SGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY-8.1.3.b IC Loss of Reactor Vessel Level that has or will Uncover Fuel in the Reactor Vessel EAL RVLIS Full Range< 57%
MODE-5, 6 BASIS This EAL is an extension of the Loss of Decay Heat Removal Capabilities EAL Alert classification as well as guidance for Modes 5 & 6 LOCA conditions. This EAL addresses Joss of inventory events such that the active fuel will be uncovered. The threshold value ofRVLIS Full Range< 57% is chosen from the EOP SET DOC for Top of Active Fuel level with no flow.
Barrier Analysis None ESCALATION CRITERIA This event would be escalated to a General Emergency on loss of Fission Product Barriers or abnormal radiological releases.
DISCUSSION This EAL addresses the effects of prolonged core boiling following a loss of decay heat removal or Mode 5/6 LOCA conditions. Full Range R VLIS indicates reactor vessel water level with no RCPs running. The intent of this EAL is to provide a RVLIS level which approximates core uncovery.
DEVIATION None EAL - 8.1.3.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, SSS EOP Setpoint Doc - K.02 EAL - 8.1.3.b Rev. 00 Page 2 of 2
SGS EAIJRALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY- 8.1.3.c IC Complete Loss of Functions Needed to Achieve or Maintain the Plant in Hot Shutdown EAL i HEATSINKREDPAIB MODE - 1, 2, 3, & 4 with RHR in Injection Mode BASIS This EAL addresses complete loss of a function required to reach Hot Shutdown conditions while operating in Mode 1, 2, 3, or Mode 4 with both trains ofRHR aligned for injection. The ability to place the plant in Mode 3 from any "at Power" condition represents the loss of Reactivity Control which is adequately addressed in Section 5.0, ATWS. CFST Heat Sink RED PATH will
- limit the ability of the Control Room crew to place the plant in a Hot Shutdown condition due the inability to remove heat from the RCS. This represents an actual loss of functions intended for protection of the public and is consistent with the Fission Product Barrier Table threshold values; thus declaration of a Site Area Emergency is warranted. This EAL works in conjunction with EAL 8.1.3.a for events which occur while the plant is in on RHR cooling.
Barrier Analysis Fuel Clad and RCS Barriers have been potentially lost.
ESCALATION CRITERIA Escalation to a General Emergency would be based on loss of Fission Product Barriers or Radiological Releases.
DISCUSSION Symptom based criteria from the Emergency Operating Procedures Critical Safety Function Tree (CFST) Monitoring program. The CFSTs are contained as a tab to the ECG. The intent of using CFST status is to simplify the identification of the threshold criteria.
EAL - 8. l.3.c Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SS4 EOP-CFST-1 EAL - 8.1.3.c Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY- 8.1.3.d IC Complet~ Loss of Functions Needed to Achieve or Maintain the Plant in Hot Shutdown EAL ALL Turbine Stop Valve Closed (MS 28)
LOSS of ALL Steam Dump Valves (TB 10, 20, 30, 40)
LOSS of ALL MS IO (Steam Generator Power-Operated Relief Valves) Valve Control (BOTH Auto AND Manual)
> 15 minutes have elapsed MODE - 1, 2, 3, and 4 with RHR in Injection Mode BASIS This EAL addresses complete loss of a function required to reach Hot Shutdown conditions while operating in Mode 1, 2, 3, or Mode 4 with both trains ofRHR aligned for injection. The inability to place the plant in Mode 3 from any "at Power" condition represents the loss of Reactivity Control which is adequately addressed in Section 5.0, ATWS. A total loss of Steam Generator heat removal capability will limit the ability of the Control Room crew to place the plant in a Hot Shutdown condition due to the inability to remove heat from the RCS. The 15 minute threshold value was added to allow for restoration of unavailable.systems. This represents an actual loss of functions intended for protection of the public; thus declaration of a Site Area Emergency is warranted. This EAL works in conjunction with EAL 8. 1. 3. a for events which occur while the plant is in on RHR cooling.
EAL - 8.1.3.d Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Escalation to a General Emergency would be based on loss of Fission Product Barriers or Radiological Releases.
DISCUSSION This EAL attempts to identify a condition where all secondary heat removal capabilities have been lost due to inability of the Steam Generators to transfer heat either to the atmosphere or the Main Condenser. This loss of heat removal capabilities will result in an inability to cooldown the RCS to a Hot Shutdown condition.
DEVIATION None REFERENCES NUMARC NESP-007, SS4 EAL - 8.1.3.d Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators UNUSUAL EVENT- 8.2.1 IC Unplanned Loss of Most or All Annunciation or Indication in the Control Room for Greater Than 15 minutes EAL Unplanned loss of> 75% of Control Room Overhead Annunciators EITHER one of the following:
- 15 minutes have elapsed since the loss of OHAs
- A significant transient is in progress MODE - 1, 2, 3, 4 BASIS A unplanned loss of most or all Control Room Overhead Annunciators without a plant transient in MODES 1, 2, 3, or 4 for greater than 15 minutes warrants a heightened awareness by Control Room Operators. Quantification of>75 is left to the discretion of the Senior Nuclear Shift Supervisor (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.
OP-AB.ANN-OOOl(Q) details increased monitoring and surveillance requirements as well as alternate indicators. 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 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.
EAL - 8.2.1 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis A significant transient is left to the determination of the SNSS/EC, but as a minimum, plant transients for this EAL should include:
- Reactor Trips (Manual and Automatic)
- Load Rejections > 25% Thermal Power
- ECCS Injections
- Thermal Power Oscillation > 10%
Barrier Analysis None ESCALATION CRITERIA This event will be escalated to an Alert if a transient is in progress or if alternate indications become unavailable and 15 minutes have elapsed since the loss of OHAs.
DISCUSSION This EAL is not required in Modes 5 or 6 due to the limited number of safety systems required for operation.
In judging the severity of the annunciator loss, consideration should be given to those annunciators needed by the operating staff for operation in abnormal and emergency operating procedures.
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 warrants an expeditious notification and not waiting the 15 minutes for the Unusual Event declaration.
REFERENCES NUMARC NESP-007, SU3 OP-AB.ANN-0001 (Q)
EAL- 8.2.1 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunicators
- ALERT - 8.2.2.a/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 Control Room Overhead Annunciators EITHER one of the following:
- Alternate Indications are NOT AVAILABLE per OP-AB.ANN-OOOI(Q)
- A significant transient is in progress 15 minutes have elapsed since the loss of OHAs MODE- 1, 2, 3, 4 BASIS A unplanned loss of most or all Control Room Overhead annunciators without a plant transient in MODES 1, 2, 3, or 4 for greater than 15 minutes warrants a heightened awareness by Control Room Operators. Quantification of "most" is left to the discretion of the Senior Nuclear Shift Supervisor (SNSS), and is considered approximately 75%. It is not intended that a detail~d count be performed, but that a rough approximation be used to determine the severity of the loss.
OP-AB.ANN-OOOI(Q) details increased monitoring and surveillance requirements as well as alternate indicators. 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 tiine of discovery, classification is not required under this EAL, but a review of the "After The Fact" RAL should be completed.
EAL - 8.2.2.a/8.2.2.b Rev. 00 Page 1 of 2 L_
SGS EAL/RALTechnical Basis Unplanned loss of annunciators excludes scheduled maintenance and testing activities.
A significant transient is left to the determination of the SNSS/EC; but, as a minimum, plant transients for this EAL should include:
- Reactor Trips (Manual and Automatic)
- Load Rejections > 25% Thermal Power
- ECCS Injections
- Thermal Power Oscillation > I 0%
Barrier Analysis None ESCALATION CRITERIA This event will be escalated to a Site Area Emergency with a failure of alternate indications and a plant transient in progress.
- 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.
This EAL is not required in Modes 5 or 6 due to the limited number of safety systems required for operation.
DEVIATION None REFERENCES NUMARC NESP-007, SA4 OP-AB.ANN-OOOI(Q)
EAL - 8.2.2.a/8.2.2.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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 Control Room Overhead Annunciators A significant transient is in progress Alternate Indications are NOT AVAILABLE per OP-AB.ANN-OOOl(Q)
Control Room indications are NOT AVAILABLE to monitor ANY one of the following:
- RCS Status
- Reactivity Control
- Secondary Systems (SGs, AFW}
- Containment Parameters MODE - 1, 2, 3, 4 BASIS A loss ( planned or unplanned) of most or all Control Room Overhead Annunciators with a plant transient in MODES 1, 2, 3, or 4 for any amount ohime warrants a heightened awareness by Control Room Operators. Quantification of>75% left to the discretion of the Senior Nuclear Shift Supervisor (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.
EAL- 8.2.3 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis A significant transient is left to the determination of the SNSS/EC, but as a minimum, plant transients for this EAL should include:
- Reactor Trips (Manual and Automatic)
- Load Rejections > 25% Thermal Power
- ECCS Injection
- Thermal Power Oscillations ~ 10%
The list of systems requiring Control Room monitoring ability (e.g.; RCS, Reactivity Control, ECCS, etc.) was included to ensure all safety functions (including the ability to shut down the reactor, maintain core cooling, maintain the RCS intact, provide for a heat sink, and maintain an intact Containment) can be determined by some form of Control Room instrumentation. OP-AB.ANN-000 l(Q), Loss of Overhead Annunciator System, details increased monitoring and surveillance requirements as well as alternate indicators.
Barrier Analysis None ESCALATION CRITERIA This event would be escalated to a General Emergency based on the loss of Fission Product Barriers or abnormal radiological releases.
DISCUSSION Without Control Room Overhead Annunciators, it may be difficult to monitor conditions associated with normal plant operations. During significant transient events such as those listed in the EAL, the difficulty becomes more acute. Compounding these, a concurrent loss of Control Room backup monitoring will further hinder Operations staff decision making needed to respond to the transient.
DEVIATION None REFERENCES NUMARC NESP-007, SS6 OP-AB.ANN-OOOI(Q)
EAL- 8.2.3 Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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:
- Station Page System (Gaitronics)
- *Station Radio System
- Direct Inward Dial System (DID)
MODE-All BASIS An Unplanned loss of communication ability significantly degrades the operating crew's ability to perform tasks necessary for plant operations and/or the ability to communicate with offsite authorities, warrants declaration of an Unusual Event. The loss of ALL ONSITE communications capability is more comprehensive than that addressed by 10CFR50.72.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 1 of2
SGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SU6 EAL - 8.3.1.a Rev. 00 Page 2 of2
SGS 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 MODE-All BASIS An Unplanned loss of communication ability significantly degrades the operating crew's ability to perform tasks necessary for plant operations and/or the ability to communicate with offsite authorities, warrants declaration of an Unusual Event. The loss of ALL OFF SITE communications capability is more comprehensive than that addressed by IOCFRS0.72.b.
Un planned 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.b Rev. 00 Page I of2
SGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SU6 EAL - 8.3.1.b Rev. 00 Page 2 of2
SGS EALIRALTechnical Basis 8.0 System Malfunctions
. 8.4 Control Room Evacuation ALERT - 8.4.2 IC Control Room Evacuation has been Initiated EAL Control Room Evacuation has been initiated MODE-All BASIS Control Room evacuation represents a serious situation since the degree of plant control at the remote shutdown locations is not as complete as it would be from the Control Room. The intent of this EAL is to declare an Alert when the determination to evacuate the Control Room has been made based on environmental/personnel safety concerns, and physical process of evacuating the Control Room has commenced.
Barrier Analysis None ESCALATION CRITERIA This event will escalate to a Site Area Emergency if Plant control cannot be established within 15 minutes from outside the Control Room.
DISCUSSION Control Room evacuation requires establishment of plant control from outside the Control Room (local control and Hot Shutdown Panels) and support from the Technical Support Center (TSC) and/or the Emergency Operations Facility (EOF) as necessary.
EAL - 8.4.2 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis The establishment of remote system control will bypass many protective trips and interlocks. In addition, much of the instrumentation and assessment tools available in the Control Room will not be available. Operator actions upon deciding that the Control Room should be evacuated include tripping the Reactor and Main Turbine, starting Auxiliary Feed Water Pumps, initiating a Main Steam Line Isolation and placing all Lockout Switches on RP4 in the Valve Operable Position.
DEVIATION None REFERENCES NUMARC NESP-007, HAS OP-AB.CR-0001 (Q)
OP-AB.CR-0002 (Q)
EAL- 8.4.2 Rev. 00 Page 2 of 2
SGS EAL!RALTechnical Basis 8.0 System Malfunctions 8.4 Control Room Evacuation SITE AREA EMERGENCY - 8.4.3 IC Control Room Evacuation has been Initiated and Plant Control Cannot Be Established EAL Control Room Evacuation has been initiated Control of the plant CANNOT be established from outside the Control Room within 15 minutes MODE-All BASIS Transfer of safety system control has not been performed in an expeditious manner and it is unknown if any damage has occurred to the fission product barriers. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and as a result there may be a threat to plant safety. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the hot shutdown areas, and reestablish plant control to preclude core uncovery and/or core damage. The term "control of the plant" will require SNSS judgment in deciding whether sufficient control has been established to maintain core cooling based upon initial reports of equipment status from Hot Shutdown Panel 213.
Barrier Analysis None ESCALATION CRITERIA This event will escalate based upon loss of Fission Product Barriers or abnormal radiological releases.
EAL - 8.4.3 Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis DISCUSSION This EAL is designed to address the conditions where due to environmental/personnel safety concerns Control Room evacuation is required. Additionally, Plant control cannot be established from outside the Control Room within 15 minutes.
DEVIATION None REFERENCES NUMARC NESP-007, HS2 OP-AB.CR-0001 (Q)
OP-AB.CR-0002 (Q)
EAL - 8.4.3 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 8.0 System Malfunctions 8.5 Technical Specifications UNUSUAL EVENT - 8.5.1 IC Inability to Reach Required Mode Within Technical Specification Limits EAL Plant is NOT brought to the required Mode within the Technical Specification required time limit MODE- 1, 2, 3, 4 BASIS Entry into this EAL should occur when it is discovered that a Technical Specification Limiting Condition for Operation (LCO) action statement requiring a plant Mode change has not been complied with. Limiting Conditions for Operation (LCOs) require the plant to be brought to a safe Mode when the Technical Specification required plant system or component configuration cannot be maintained/restored. This Unusual Event is entered when the plant fails to COMPLY WITH THE ACTION STATED in a LCO, not when the action is required.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated based upon system malfunctions or other conditions covered in various other EAL sections.
DISCUSSION Depending on the circumstances, this may or may not be a precursor to a more severe condition.
A shutdown required by the site Technical Specifications requires a 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 Technical Specifications. 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 time period and is not related to how long a plant condition may have existed.
EAL- 8.5.1 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SU2 SGS Technical Specifications IOCFRS0.72 EAL - 8.5. l Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.1 Security Threats \0\
UNUSUAL EVENT - 9.1.1 IC Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant EAL Confirmed security threat directed towards 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-10 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 MODE-All BASIS A security threat that is identified as being directed towards the station represents a potential degradation in the level of safety of the plant. 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 security threat is confirmed 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 will declare an Unusual Event upon consulting with 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.
EAL - 9.1.1 Rev. 00 Page 1 of 2 L
SGS EAL!RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to an Alert based upon an actual Protected Area intrusion, malicious acts, or destructive devices discovered within a Vital Area.
DISCUSSION Security events which do not represent a potential degradation in the level of safety of the plant are reported under RAL 11.7.1.a an a One Hour Non-Emergency Safeguards event.
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, HU4.1, HU4.2 Safeguards Contingency Plan EAL - 9.1.1 Rev. 00 Page 2 of 2
SGS 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 MODE-All BASIS This class of security event represents an escalated threat to the level of safety of the plant. This 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 plant Protected Area. The SNSS/EC will 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 This event will be escalated to a Site Area Emergency based upon a hostile intrusion in plant Vital Areas.
EAL - 9.1.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical 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
SGS EALIRALTechnical 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 the Vital Area, resulting in SCP-6 implementation
- M~icious acts or destructive device discovered in a Vital Area, resulting in SCP-I 0 implementation MODE-All BASIS This class of security event represents 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 Vital Area. The Vital Areas are within the Protected Area and are generally controlled by key card readers. These areas contain vital equipment which includes any equipment, system, device or material required for safe shutdown and for protection of the health and safety of the public and plant personnel.
The Security Contingency Procedure (SCP) number is referenced following the 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 This event will be escalated to a General Emergency based upon the loss of physical control of the Control Room or Remote Shutdown Capability.
EAL - 9.1.3 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION 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 Devices or Evidence of Malicious Acts" DEVIATION None REFERENCES NUMARC NESP-007, HSI.I, HSl.2 Safeguards Contingency Plan EAL - 9.1.3 Rev. 00 Page 2 of 2
SGS EALIRALTechnical 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:
- Control Room
- Remote Shutdown Panel 213 MODE-All BASIS Security events classified under this EAL represent conditions under which a hostile force has ken physical control of areas required to reach and maintain Cold Shutdown. Both the 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 equipment controlled from the Control Room.
Barrier Analysis None ESCALATION CRITERIA NIA DISCUSSION The Remote Shutdown Panel 213 was the only panel included in this EAL due to its central location and ability to allow for physical control of multiple Safety Related components without detailed knowledge of plant operations. Security threats which meet the threshold for declaration of a General Emergency are an actual loss of physical control of the Control Room or remote EAL - 9.1.4 Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis shutdown locations. This situation places the plant in a potentially unstable condition with high potential of multiple fission product barrier failures.
DEVIATION None REFERENCES NUMARC NESP-007, HGl Safeguards Contingency Plan EAL - 9.1.4 Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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 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)
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area
- Turbine Building 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 MODE-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, EAL- 9.2. l Rev. 00 Page 1 of 3
SGS EALIRALTechnical Basis due to the potential for a fire to spread from a non-safety related structure to an adjoining safety related structure.
A fire alarm received in the 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 Control Room or by visual confirmation if only a single detector is alarming.
This EAL EXCLUDES such items as fires in Plant Structures other then 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 of a 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 ifthe 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 portion of the plant, but has the potential to spread.
Excluded or non-vital structures include:
Unit 3 Main or Aux Guard House Circulating Water Structure Main, Aux, and Switchyard Transformers B-building EAL- 9.2.1 Rev. 00 Page 2 of 3
SGS EAL/RALTechnical Basis Onsite Trailers Salem Adrnin. Building Onsite Warehouses Nuclear Services Building
. DEVIATION None REFERENCES NUMARC NESP-007, HU2 MlO-FRS-I-0001, Control Room Fire Response NUMARC Q & A, JUNE 1993 EAL - 9.2.1 Rev. 00 Page 3 of 3
SGS 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:
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area The Fire is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
- TWO OR MORE Trains 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 MODE of operation MODE-All BASIS The primary concern in this EAL is the magnitude of the fire and the effects on safety systems required for the present MODE of Operation. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to EAL - 9.2.2 Rev. 00 Page 1 of 3
SGS EALIRALTechnical Basis 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/struct~re 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 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 if the structure houses or otherwise supports Safety Systems required for the present MODE of operation.
For example, a fire that has been confirmed to be localized to a single piece of equipment, like a 4KV 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 4KV Breakers for component(s) required for the present MODE of Operation, 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 This event will be 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 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 train 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 if the 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.
EAL-9.2.2 Rev. 00 Page 2 of 3
SGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HA2 MlO-FRS-I-001, Control Room Fire Response EAL- 9.2.2 Rev. 00 Page 3 of 3
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 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 MODE-All BASIS Occurrence of this event within the Protected Area, that causes 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 permanent plant structures, systems or components, has occurred.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated to Alert if the explosion damages more than one safety system or damages any plant vital structure as per EAL 9.3.2.
EAL - 9.3.1 Rev. 00 Page 1 of 2
SGS 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 MlO-FRS-1-0001, Control Room Fire Response EAL- 9.3.1 Rev. 00 Page 2 of 2
SGS 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:
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area AND The Explosion is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
- TWO OR MORE Trains 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 MODE of operation MODE-All BASIS The primary concern in this EAL is the magnitude of the explosion and the effects on Safety Systems required for the present MODE of Operation. 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 EAL - 9.3.2 Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis equip~ent/structure being INOPERABLE or otherwise incapable ofperfonning 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 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 MODE of Operation.
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 This event will be 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 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, ifthe explosion is big enough that it has damaged MORE THAN ONE safety system, or more than one train 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 MlO-FRS-I-001, Control Room Fire Response EAL - 9.3.2 Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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 MODE-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 evacuation 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 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 EAL - 9.4.1.a Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis None DEVIATION None REFERENCES NUMARC NESP-007, lill3.2 SC.OP-AB.CR-0003(Q)
EAL - 9.4.1.a Rev. 00 Page 2 of 2
SGS EALIRALTechnical 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 MODE-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 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 warrants declaration of an Unusual Event. A Cardox discharge inside any area that contains this safety feature (i.e. Diesel Room) does not warrant Unusual Event declaration, unless personnel injuries have occurred as a direct result of the discharge or personnel must enter the area using respiratory equipment.
EAL - 9.4.1.b Rev. 00 Page 1 of 2
- SGS EALIRALTechnical 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 ifthe Toxic Gas enters 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
- activities. These areas include 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, HU3.1 SC. OP-AB. CR-0003(Q)
EAL - 9.4.1.b Rev. 00 Page 2 of 2
SGS 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 MODE-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 can not 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.
Flammable Gas release is considered to be IMPEDING normal plant operations if concentrations are high enough to restrict routine operator movements. Access restrictions EAL - 9.4.1.c Rev. 00 Page 1 of 2
SGS 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 of 25% 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. l SC.OP-AB.CR-0003(Q)
EAL - 9.4.1.c Rev. 00 Page 2 of 2
SGS 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 Operations or to Establish or Maintain Cold Shutdown Conditions EAL Uncontrolled Toxic Gas release within ANY one of the following Plant Vital Structures
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Area
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area 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 MODE-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 EAL - 9.4.2.a Rev. 00 Page 1 of 3
SGS EAL/RALTechnical Basis 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 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.
Barrier Analysi~
NIA ESCALATION CRITERIA Emergency Classification will be 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 ifthe 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 activates. 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 EAL - 9.4.2.a Rev. 00 Page 2 of 3
SGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, HA3 .1 SC.OP-AB.ZZ-0003(Q)
EAL - 9.4.2.a Rev. 00 Page 3 of 3
SGS EAL/RALTechnical Basis 9.0 Hazards - InternaVExternal
. 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 Vital Structures
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Area
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area 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 MODE-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.
Releases classified under this EAL include those that originate both onsite and offsite.
EAL - 9.4.2.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis 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 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 Auxiliary 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 be escalated based on subsequent damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may discretion and escalate the classification to SAE based on the nature of the flammable gas 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, HA3.2 SC.OP-AB.ZZ-0003(Q)
EAL - 9.4.2.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.5 Seismic Events UNUSUAL EVENT - 9.5.1.a/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 the Seismic Monitor System cabinet in the # 1 CR Equipment Room MODE-All BASIS An earthquake of this magnitude is not expected to affect the capability of plant safety functions.
A seismic event recording a magnitude of>O.Olg is the threshold level at which the Seismic Monitoring System would monitor the event. The actual value can be determined by engineering evaluation of the acceleration of gravity as read on the seismic recorder, information provided by Hope Creek station, or confirmation by the National Earthquake Center.
The Overhead Annunciator, "SEIS RCDR SYS ACT" will alert operators to this event and the seismic monitoring instrumentation would begin to monitor the event. This value is well below the Operating Basis Earthquake of O. lg.
Barrier Analysis None ESCALATION CRITERIA Escalation of this event would occur if actuation of the Hope Creek Seismic Switch (>O. lg) has occurred. Call the Hope Creek SNSS to request this information.
EAL - 9.5.1.a/9.5. I .b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION An earthquake of this magnitude is not expected to affect the capability of plant safety functions.
For further information, the National Earthquake Center can be contacted at (303) 273-8500. An approximate relationship between acceleration of gravity 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 O.lg 5.5 0.2g 6.5 DEVIATION None REFERENCES NUMARC NESP-007, HUI.I UFSAR, Chapter 52, Seismic Monitoring System EAL - 9.5.1.a/9.5.1.b Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.5 Seismic Events ALERT - 9.5.2 IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Valid Actuation of the Hope Creek Seismic Switch(> O.lg) has occurred as verified by the Hope Creek SNSS MODE-All BASIS The Operating Basis Earthquake of O.lg has been exceeded for both Salem and Hope Creek.
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.
The actual value can be determine by engineering evaluation of the acceleration of gravity as read on the seismic recorder, information provided by Hope Creek station, or confirmation by the National Earthquake Center. The Overhead Annunciator, "SEIS RCDR SYS ACT" will alert operators to this event and the seismic monitoring instrumentation would begin to monitor the event.
Barrier Analysis NIA ESCALATION CRITERIA
. Escalation of.this event would occur ifthe seismic event caused additional dainage 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 event.
EAL- 9.5.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION The Overhead Annunciator, "SEIS RCDR SYS ACT" will ~lert operators to this event and the seismic monitoring instrumentation would begin to monitor the event. If analysis of the event indicates that the threshold value has been exceeded, immediate plant shutdown is required to evaluated possible equipment damage. This threshold value 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.
For further information, the National Earthquake Center can be contacted at (303) 273-8500. An approximate relationship between acceleration of gravity 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 O.lg 5.5 0.2g 6.5 DEVIATION None REFERENCES NUMARC NESP-007, HAI .1 UFSAR, Chapter 52, Seismic Monitoring System EAL - 9.5.2 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.6 High Winds UNUSUAL EVENT - 9.6.1.a/9.6.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report of a Tornado TOUCHING DOWN within the Protected Area Sustained wind speeds> 75 MPH for 15 minutes, from ANY elevation of the Met Tower MODE-All BASIS This EAL addresses either a tornado reported onsite or sustained, high winds being detected
.onsite. A tornado touching down within the Protected Area or sustained wind speeds in excess of 75 MPH are of sufficient velocity to have the potential to cause damage to Plant Vital Structures.
These conditions are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. "Sustained" wind speed means winds in excess of the threshold value for greater than 15 minutes.
Barrier Analysis None ESCALATION CRITERIA This event will be escalated to an Alert ifthe tornado or high winds cause damage to Plant Vital Structures. If it is determined that the abnormal weather condition results in a loss of shutdown cooling, then the event will be escalated based on the Loss of Decay Heat Removal Capability.
DISCUSSION 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 108 mph, and is set slightly above the threshold value used to EAL - 9.6.1.a/9.6. l.b Rev. 00 Page 1 of 2
SGS EAL/RALTechnical Basis characterize Category Level I Hurricane force winds (74 mph). Setting this threshold value at>
75 mph ensures site accessibility for emergency response.
NOTE: The Wind Speed indication from the Met Tower instrumentation is full scale at l 00 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 REFERENCES NUMARC NESP-007, HUl.2 and HUl.7 OP-AB.ZZ-OOOl(Q), Severe Weather SGS UFSAR, Sections 2.3, 3.3 EAL - 9.6.1.a/9.6. l.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 9.0 Hazards - IntemaVExtemal 9.6 High Winds ALERT - 9.6.2 IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL EITHER one of the following:
- Report of a Tornado TOUCHING DOWN within the Protected Area
- Sustained wind speeds> 75 MPH for 15 minutes, from 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 Trains of a Safety System
- MORE THAN ONE Safety System
- Rendering ANY of the following structures incapable of performing its Design Function:
- Auxiliary Building
- Service Water Intake Structure
- Control Point Area
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area Damaged Safety System(s) or Plant Vital Structure is required for the present MODE of operation MODE-All EAL- 9.6.2 Rev. 00 Page 1 of 3
SGS EALIRALTechnical Basis BASIS The primary concern in this EAL is the magnitude of the high winds and the effects on safety functions. 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 have caused component malfunction (pump trip, breaker trip, etc.) or a report of visible deformation 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 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 if the structure houses or otherwise supports safety systems required for safe shutdown of the Plant.
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. The value of 75 MPH is below the design basis wind speed of 108 MPH determined for Salem Generating Station. "Sustained" wind speed means winds in excess of the threshold value for greater than 15 minutes.
Barrier Analysis None ESCALATION CRITERIA This event will be escalated to higher classifications based upon damage consequences covered under various other EAL sections. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the winds.
DISCUSSION With damage to these areas confirmed, an actual degradation in the level of plant safety has occurred. EC judgment must be used to discriminate between minor "cosmetic" and "design function 11 structural damage.
NOTE: The Wind Speed indication from the Met Tower instrumentation is full scale at 100 mph.
EAL- 9.6.2 Rev. 00 Page 2 of 3
SGS EALIRALTechnical 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, HAl.2 and HAl.3 OP-AB.ZZ-OOOl(Q), Severe Weather SGS UFSAR, Sections 2.3, 3.3 EAL- 9.6.2 Rev. 00 Page 3 of 3
SGS E.At/RALTechnical 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 Severe Flooding of Safety System Areas HAS ENDANGERED safety related equipment per OP-AB.ZZ-0002 MODE-All BASIS This EAL addresses conditions where severe flooding is occurring in areas that affect safety related equipment. Endangered means that a determination has been made that the flooding is severe enough to jeopardize safe operation of Safety related equipment.
Barrier Analysis None ESCALATION CRITERIA This event will be escalated to an Alert based upon the loss of vital equipment due to flooding.
DISCUSSION Severe flooding can occur from several sources including the Circulating Water System, Service Water System, Demineralized Water, Component Cooling Water, Fire Protection and Refueling Water Storage Tanlc.
Flooding is detailed in these areas by visual report from staff or by confirmation of sump alarms.
OP-AB.ZZ-0002(Q) directs the operators to determine the exact location and severity of flooding. Attachments in this procedure delineates the affected plant areas, potential source(s) of water, affected vital equipment, flood rate and time to submerge vital equipment.
EAL- 9.7.1 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HUl.7 OP-AB.ZZ-0002(Q), Flooding EAL- 9.7.1 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.7 Flooding ALERT- 9.7.2 IC Internal Flooding Affecting the Ope~ability 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:
- Auxiliary Building
- Service Water Intake Structure
- Fuel Handling Building
- Service Building
- Containment The Flooding is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
- TWO OR MORE Trains 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 Safety System(s) or Plant Vital Structure is required for the present MODE of operation MODE-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 MODE of operation. 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 EAL- 9.7.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis 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 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 MODE of operation.
Barrier Analysis None ESCALATION CRITERIA This event will be escalated based upon the consequences of the loss of vital equipment as covered in various other EAL sections. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the flooding.
DISCUSSION Severe flooding can occur from several sources including the Circulating Water System, Service Water System, Demineralized Water, Component Cooling Water, Fire Protection and Refueling Water Storage Tank.
Flooding is detailed in these areas by visual report from staff or by confirmation of sump alarms.
OP-AB.ZZ-0002(Q) directs the operators to determine the exact location and severity of flooding. Attachments of this procedure delineates the affected plant areas, potential source( s) of water, affected vital equipment, flood rate and time to submerge vital equipment.
DEVIATION None REFERENCES NUMARC NESP-007, HAl.7 OP-AB.ZZ-0002(Q), Flooding EAL- 9.7.2 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.8 Turbine Failure I Vehicle Crash I Missile Impact UNUSUAL EVENT - 9.8.1.a IC Natural and Destructive Phenomena Affecting Certain Structures Within 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 MODE-All BASIS 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 or damage to plant systems due to missiles. The presence ofH2 gas in sufficient quantities may present a flammable/explosive hazard. Oil may also be present which may contribute to the flammability hazard.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated 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 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
SGS 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 EOP-TRIP-1 EAL - 9.8.1.a Rev. 00 Page 2 of 2
SGS EAL/RALTechnical 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 Vital Structures:
- Auxiliary Building
- Service Water Intake Structure
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area MODE-All BASIS A Vehicle Crash I Missile Impact with or within a listed Plant Vital 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 This event will be escalated to Alert if the crash or missile impact causes damage to Plant Vital Structures.
EAL - 9.8.1.b Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis DISCUSSION Any security aspects of this event should be considered under EAL sections covering Security Events.
- DEVIATION None REFERENCES NUMARC NESP-007, HUI .4 NUMARC Questions and Answers, June 1993, "Hazards Question #6" I
\/
EAL - 9.8.1.b Rev. 00 Page 2 of 2
SGS EAL!RALTechnical 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:
- Auxiliary Building
- Service Water Intake Structure
- Inner/Outer Penetration Areas
- Containment
- Fuel Handling Building
- Service Building
- RWST, PWST, and AFWST Area 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 Trains 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 MODE of operation MODE-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 MODE of operation. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system EAL- 9.8.2 Rev. 00 Page 1 of 2
SGS EALIRALTechnical Basis damage is not required prior to classification. The term "Damage" is defined as evidence that the vehicle crash I missile impact has caused component malfunction (pump trip, breaker trip, etc.)
that may have resulted in the equipment/structure being INQPERABLE or otherwise incapable of performing it's design functio'n.
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 crash 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 structure houses or otherwise supports safety systems required for the present MODE of operation.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated 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 to SAE based on the nature of the damage.
DISCUSSION No lengthy or 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 vehicle crashes I missile impact.
In short, if the vehicle crash I missile impact is big enough that it has damaged more than one safety system, or more than one train of a safety system, then the vehicle crash I missile impact 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. Any security aspects of this event should be considered under EAL sections covering Security Events.
DEVIATION None REFERENCES NUMARC NESP-007, HAI.5 and HAI.6 NUMARC Questions and Answers, June 1993, "Hazards Question #6" EAL - 9.8.2 Rev. 00 Page 2 of 2
SGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.9 River Level UNUSUAL EVENT - 9.9.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL River Level> 99.5' MODE-All BASIS This EAL indicates river level conditions that can threaten the level of safety of the plant due to flooding.
Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated 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' is indication of impending site flood conditions. Flood protection measures are required by Salem Technical Specifications and procedure at 99.5'(+10.5'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. Hope Creek performs these actions at 95.0' (+6.0'MSL).
The High river level threshold is below the river level that would require a plant shutdown.
Technical Specification actions required by a River Level of>I00.5' includes placing the plant in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
This is based on the river level at which facility flood protection features provide protection to safety related equipment. Hope Creek required actions are at 99.5' (+ 10.5'MSL).
EAL - 9.9.1.a Rev. 00 Page 1 of 2
SGS EALIR.ALTechnical Basis The grade level at the Salem station is lower than that for Hope Creek (Salem= 99.5', Hope Creek= 101.5').
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, HUl.7 OP-AB.CW-OOOI(Q)
OP-AB.ZZ-0001 (Q)
SGS UFSAR, Section 2.4.11.2, Figure 3.4-1 HCGS UFSAR, Section 2.4, Figure 2.4-3 EAL - 9.9.1.a Rev. 00 Page 2 of 2
SGS 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 i River Level < 80.0' MODE-All 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/Occurrences 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. l.b Rev. 00 Page 1 of 2
SGS EAL!RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HUl.7 OP-AB.CW-OOOl(Q)
Sl.OP-AB.ZZ-0001 (Q)
S2.0P-AB.ZZ-0001 (Q)
HC Operability Determination 961001148 SGS UFSAR, Section 2.4.11.2, Figure 3.4-1 HCGS UFSAR, Section 2.4, Figure 2.4-3 EAL - 9.9.1.b Rev. 00 Page 2 of 2
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels
. 11.1 Technical Specifications REPORT ABLE 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 MODE- I, 2 BASIS This RAL addresses the conditions requiring a one hour report in accordance with 10CFR50.72(b)(l)(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 plant shutdown as required by technical specifications. This includes any means of power reduction such as rod insertion or boron concentration changes.
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 T/S 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 10CFRS0.72.
REFERENCES 10CFRSO.72(b)(l)(i)(A)
NUREG 1022, Rev. 1, 2nd Draft Page 1of1 RAL - 11.1.1.a Rev. 00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications
. REPORT ABLE ACTION LEVEL - 11.1.1. b IC EXCEEDING ANY TECHNICAL SPECIFICATION SAFETY LIMIT
[ IOCFRS0.36(c)( 1)]
RAL Exceeding EITHER one of the following Technical Specification Safety Limits:
- TIS 2.1.1, Thermal Power, Pressurizer Pressure, Coolant Temperature combination
- TIS 2.1.2, RCS Pressure MODE - 1, 2, 3, 4, 5 (as applicable in TIS)
BASIS This RAL addresses the conditions requiring a one hour report IAW IOCFR50.36(c)(l) which states that exceeding a Technical Specification (TIS) Safety limit requires going to Hot Standby (Mode 3) by TIS (or, if already in Modes 3, 4, or 5, a restoration of RCS pressure to within its limits within 5 minutes).
For ANY Mode of Operation, exceeding EITHER Safety Limit in TIS Section 2.1 shall be reported under this RAL.
REFERENCES IOCFR50.36(c)(l)
TIS 6.7 Page 1of1 RAL - l l. I.1.b Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORT ABLE ACTION LEVEL - 11.1.1.c IC ANY DEVIATION FROM T/S OR LICENSE CONDITION PURSUANT TO 10CFR50.54(x) [10CFR50.72(b)(l)(i)(B)]
RAL Deviation from written procedures because no action consistent with Technical Specifications or license condition can provide adequate or equivalent protection in an emergency (see NC.NA-AP.ZZ-OOOS(Q) for guidance on deviation from procedures)
MODE-All BASIS This RAL addresses conditions that require a one hour report in accordance with 10CFRS0.72(b)(l)(i)(B). 10CFR50.54(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 CC!-n 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 of the affected Unit.
Refer to NC.NA-AP.ZZ-OOOS(Q), Station Operating Practices, for more information concerning the use of 10CFR50.54(x).
REFERENCES 10CFR50.54(x) 10CFRS0. 54(y) 10CFRSO.72(b )(1 )(i)(B)
NC.NA-AP.ZZ-OOOS(Q)
NUREG 1022, Rev. 1, 2nd Draft Page 1of1 RAL - 11. 1.1.c Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORTABLE ACTION LEVEL- 11.1.2.a IC STEAM GENERATOR TUBE INSPECTIONS WHICH FALL INTO CATEGORY C-3 THAT HAVE BEEN EVALUATED FOR REPORT ABILITY
[10CFR50. 72(b)(2)(i);T/S 4.4.5.2(6.2)]
RAL Results of S/G tube inspections which fall into Category C-3 ofT/S 4.4.5.2 (Unit 1) or T/S 4.4.6.2 (Unit 2)
An Engineering Evaluation has determined that it is reportable pursuant to 10CFR50.72(b)(2)(i)
MODE - 5, 6, Defueled BASIS T/S 4.4.5.5c (U-1) and 4.4.6.5c (U-2) Category C-3 require that the results of any Steam Generator Tube inspections that are performed while in Mode 5, 6 or defueled be evaluated for Steam Generator operability before exiting these Modes.
10CFR50.72(b)(2)(i) requires a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report on any event, found while the reactor is shutdown, that, had it been found. while the reactor was in operation, would have resulted in the plant's principal safety barriers being seriously degraded.
REFERENCES 10CFR50.72(b)(2)(i)
TS 4.4.5.5c(U/l)
TS 4.4.6.5c(U/2)
Pagel of l RAL - 11.1.2.a Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORTABLE ACTION LEVEL- 11.1.2.b IC ABNORMAL DEGRADATION OF THE CONTAINMENT STRUCTURE DETECTED DURING SHUTDOWN THAT HAS BEEN EVALUATED FOR REPORTABILITY [10CFRS0.72(b)(2)(i); TIS 4.6.1.6.2]
RAL Any abnormal degradation of the Containment structure detected by visual inspection of exposed accessible interior and exterior surfaces during shutdown An Engineering Evaluation has determined that it is reportable pursuant to 10CFRSO.72(b)(2)(i)
MODE - 3, 4, 5, 6, Defueled BASIS This RAL is based on the reporting requirements of 10CFR50.72(b)(2)(i), which requires a four hour report for any event found while the reactor is shutdown that, had it been found while the reactor was in operation, would have resulted in a principal safety barrier being seriously degraded or being in an unanalyzed condition.
REFERENCES IOCFRS0.72 (b)(2)(i)
TIS 4.6.1.6.2 Page 1of1 RAL- 11.1.2.b Rev. 00
SGS 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 [Salem Unit 2 Operating License, Sections 2.I]
RAL Violation of ANY one of the requirements contained in Section 2.C (Items 3 through 25) or Section 2E, 2F, or 2G of the Salem Unit 2 Operating License MODE-All BASIS This RAL addresses the conditions for a twenty-four hour report in accordance with Item 2.I of the Operating License of SGS Unit 2.
SGS Unit I Facility Operating License does not contain similar reporting criteria.
REFERENCES Salem Unit 2 Facility Operating License, Sections 2.C and 2.I Pagel of l RAL - 11.1.3.a Rev. 00
SGS 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
[Ul TIS 3.4.9.1, 3.4.9.2, 3.4.7, 3.7.9, JAN 1983, LTR TO NRC, 3.7.2.1]
[U2 TIS 3.4.10.1, 3.4.10.2, 3.4.8, 3.7.9, JAN 1983, LTR TO NRC, 3.7.2]
RAL As judged by the SNSS/EDO, ANY one of the following conditions have been satisfied:
- Any of the TIS LCOs for RCS or PZR heatup or cooldown rates are exceeded
- The concentration of either chloride or fluoride in the RCS is in excess of its Steady State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or in excess of its Transient Limit, thereby requiring
- an engineering evaluation to determine the effects of the out oflirnit condition on the structural integrity of the RCS
- One or more snubbers are found to be INOPERABLE and require an engineering evaluation performed in accordance with T.S.4.7.9 action statement
- Any PZR code safety valve discharges
- The temperature of EITHER the Primary or Secondary Coolant in any SIG~ 70° F WHEN the pressure of either the Primary or Secondary Coolant in the SIG is > 200 psig MODE-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.
These events require an Engineering Evaluation of the effects of the transient on plant materials and future operation. This RAL ensures that timely internal notification is initiated to implement the evaluations.
Page I of2 RAL-11.1.3.b Rev. 00
SGS EAL!RALTechnical Basis REFERENCES
- 1. [T/S 3.4. 9.1 OR 9.2] Ul
[T/S 3.4.10.1 OR 10.2] U2
- 2. T/S 3.4.7 Ul T/S 3.4.8 U2
- 3. T/S 3.7.9
- 4. JAN 1983, LTR TONRC
- 5. T/S 3.7.2. l Ul T/S 3.7.2 U2 Page 2 of2 RAL- 11.1.3.b Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition
.REPORT ABLE ACTION LEVEL - 11.2.1.a IC ANY EVENT OR CONDITION DURING OPERATION THAT RESULTS IN THE CONDITION OF THE PLANT BEING SERIOUSLY DEGRADED
[ l OCFR50. 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 operatifJ.g procedures MODE- 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.
- 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.
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 variations in Page I of2 RAL - 11.2.1.a Rev. 00
SGS EAL!RALTechnical Basis parameters or to problems concerning single pieces of equipment. The NRC understands that PSE&G will use engineering judgment and experience to determine if an unanalyzed condition exist.
When applying engineering judgment, if 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 require high energy line break restraints not being installed.
- One train of a safety system has been incapable of performing its design function for ap.
extended time.
REFERENCES I OCFR50. 72(b )( 1)(ii)
NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL- 11.2.1.a Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.2 *Design Basis I Unanalyzed Condition REPORTABLE ACTION LEVEL - 11.2.1.b IC PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS [IE Bulletin 79-17]
RAL Cracks in weld areas of Borated Safety Related piping (as reported by Engineering or ISI)
MODE-All BASIS This RAL deals with cracks in safety-related stainless steel piping systems and portions of systems which contain oxygenated, stagnant (or essentially stagnant) borated water.
REFERENCES IE Bulletin 79-17 Page 1of1 RAL - 11.2.1.b Rev. 00
SGS EALIRALTechnical 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 DURING OPERATION, WOULD HAVE SERIOUSLY DEGRADED THE PLANT OR RESULTED IN BEING IN AN UNANALYZED CONDITION [IOCFR50.72(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 MODE - 3, 4, 5, 6, Defueled BASIS See RAL 11.2.1.a 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 Page l of l RAL - 11.2.2.a Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORTABLE 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 fulfillment 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 MODE-All BASIS The intent of this RAL is to require reporting of events or conditions that could have prevented systems 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, in your judgment, the functional capability of the overall system is/was jeopardized.
Page 1 of2 RAL - 11.2.2.b Rev. 00
SGS EALIRALTechnical 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 10CFRS0.72 (b)(2) (iii)
NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.2.2.b Rev. 00
SGS EALIRAL Technical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORTABLE 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 MODE-All BASIS This RAL addresses the conditions requiring a prompt notification with written follow-up report of operating information in accordance with Regulatory Guides 1.133 and 1.16. Presence of a loose part maybe indicated by an overhead alarm and can be monitored both visually and audibly on the Metal Impact Monitoring System (MIMS).
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 failur*e or other plant events.
REFERENCES Reg. Guide 1.16 Reg. Guide 1.133, Rev. I Page 1of1 RAL - 11.2.2.c Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.3 Engineered Safety Features (ESF)
REPORTABLE ACTION LEVEL- 11.3.1 IC ANY EVENT THAT RESULTS OR SHOULD HAVE RESULTED IN ECCS DISCHARGE INTO THE RCS AS THE RESULT OF AVALID SIGNAL
[ 10CFRSO.72(b)(1 )(iv)]
RAL Valid SI Actuation signal received (or demanded)
ANY ECCS Pump start or Accumulator depressurization that results in or should have resulted in, discharge to the RCS MODE-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 SI actuation or would have resulted in SI actuation if some component had not failed or an operator action had not been taken are reportable.
For example, while performing a RCS cooldown following a controlled Reactor Shutdown, a Main Steam Line Af> SI is inadvertently generated. However, the Charging Pumps fail to start and RCS pressure remains above the SI Pump shutoff head pressure. Although no ECCS discharge to the vessel occurred, the event is reportable.
A valid signal refers to actual plant conditions or parameters satisfying the requirements for SI initiation. Valid actuations also include intentional manual actuations unless the actuation is part of a preplanned test. Excluded from this reporting requirement would be those instances in which instrument drift, spurious signals, human error or other invalid signals caused SI actuation (e.g.
jarring a cabinet, an error in the use of jumpers or lifted leads, error in actuation of controls switches, or equipment failures or radio frequency interference).
Page 1 of2 RAL-11.3.l Rev. 00
SGS EALIRALTechnical Basis IF the SI Actuation discharges or should have discharged into the RCS 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.
REFERENCES NC.NA-AP.ZZ-OOOO(Q), Action Request Process SGS UFSAR I OCFRSO. 72(b )(I )(iv)
IOCFRS0.73 NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL-11.3.l Rev. 00
SGS 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 [10CFR50.72(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.
MODE-All BASIS This RAL addresses the 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 operating mode 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:
- 1. 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 Control Room.
However, ifthe 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.
- 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.
Page 1 of2 RAL - 11.3.2 Rev. 00
SGS EALIRALTechnical Basis This would include required 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 specific guidance on the reportability and reporting requirements for such events and should be referenced prior to determining reportability.
REFERENCES NC.NA-AP.ZZ-OOOO(Q), Action Request Process SGS UFSAR l OCFRSO. 72(b )(2)(ii) 10CFRS0.73 NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.3.2 Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE 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:
- Release of radioactive material inside or outside of a Restricted Area so that, had an ~
individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the individual could have received ~ 5 times the occupational ALI (Annual Limit oflntake) which would usually equate to~ 25 Rem CEDE. This DOES NOT apply to areas where personnel are NOT normally stationed during routine operations MODE-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.
Terms:
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 l of2 RAL- 11.4.l Rev.00
SGS EAL!RALTechnical Basis REFERENCES 10CFR20.2202(a) 10CFR20, App. B Page 2 of2 RAL - 11.4.1 Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE 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:
- Release of radioactive material inside or outside of a Restricted Area so that had an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.
MODE-All BASIS This RAL addresses those conditions requiring a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report IAW 10CFR50.72(b)(2)(vi).
Terms: (The below listed terms are defined in RAL 11.4.1)
TEDE = Total Effective Dose Equivalent LDE = Lens Dose Equivalent SDE = Shallow Dose Equivalent CEDE= Committed Effective Dose Equivalent ALI = Annual Limit of Intake Page I of2 RAL - 11.4.2.a Rev. 00
SGS 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
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORT ABLE ACTION LEVEL - 11.4.2.b IC ONSITE FATALITY [IOCFR50.72(b)(2)(vi)]
RAL Any fatality has occurred within the Owner Controlled Area (OCA)
MODE-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 ON SITE 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 I OCFR50. 72(b)(2)(vi)
NUREG 1022, Rev. I; 2nd Draft Page 1of1 RAL - 11.4.2.b Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE ACTION LEVEL- 11.4.2.c IC RADIOACTIVELY CONTAMINATED 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.
MODE-All BASIS This RAL addresses the conditions requiring a four hour report in accordance with 10CFRSO.72(b)(2)(v). Transportation of a radioactively contaminated individual to an offsite medical facility has the potential for spreading the contamination 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 1022, Rev. 1, 2nd Draft Pagel of l RAL - 11.4.2.c Rev. 00
SGS EALIRALTechnical 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 [10CFR26.73]
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 MRO or designee.
MODE-All BASIS NC.NA-AP.ZZ-0042(Q) provides the guidance to determine reportability of Significant Fitness for Duty event which requires a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report IAW 10CFR26.73. 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) 10CFR26.73 Page I of I RAL- 114.3.a Rev. 00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE ACTION LEVEL- 11.4.3.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 10CFR26 as determined by the Medical Review Officer (l\1RO) 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.
MODE-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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report IAW 10CFR26.
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 l\1RO, 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
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.5 Environmental REPORTABLE ACTION LEVEL - I 1.5.2.a IC SPILL!DISCHARGE OF ANY NON-RADIOACTIVE HAZARDOUS SUBSTANCE
[10CFR50.72(b)(2)(vi); NJ.AC. 7: IE]
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 / discharge that has passed through the engineered fill and into the ground water as confirmed by Licensing
- Spill / discharge that CANNOT be cleaned up within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and no contact with groundwater is suspected NOTE:
This event MAY require IMMEDIATE (15 minute) notifications. DO NOT delay implementation of Attachment I 6.
MODE-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 I6, 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)
NJ.AC. 7: IE DPCC!DCR Plan, Part III Page I of I RAL - 11.5.2.a Rev. 00
SGS 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.7:1E]
RAL EITHER one of the following events occur:
- Observation of a spill/discharge of an industrial chemical or petroleum 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.
MODE-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: lE DPCC/DCR Plan, Part III Pagel of l RAL - 11.5.2.b Rev. 00
. SGS 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)
MODE-All BASIS This RAL addresses the conditions requiring reports IAW the Environmental Protection Plan.
Final determination or reportability will be made by Environmental Licensing as a result of implementing Attachment 15.
REFERENCES SGS Technical Specifications, ENVIRONMENT AL PROTECTION PLAN Pagel of l RAL - 11.5.2.c Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.6 After The Fact REPORT ABLE ACTION LEVEL - 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 MODE-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 Salem ECG Introduction Section NUREG I 022, Rev. I, 2nd Draft, Pg. 20 Page 1of1 RAL - 11.6.1 Rev.00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels
- 11. 7 Security I Emergency Response Capability REPORTABLE ACTION LEVEL- 11.7.1.a IC SAFEGUARDS EVENTS THAT ARE DETERMINED TO BE NON-ErvffiRGENCIES, BUT ARE REPORTABLE TO THE NRC WITIIlN ONE HOUR [10CFR73.7l(b)(l)]
RAL Any Non-Emergency safeguards event that is reportable in accordance with 10CFR73.71 as determined by Security (SCP-15)
MODE-All BASIS This RAL addresses the conditions requiring a one hour report in accordance with 10CFR73. 71 (b )( 1). 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 Pagel of l RAL- 11.7.1.a Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.7 Secu~ty I Emergency Response Capability REPORTABLE ACTION LEVEL- 11.7.1.b IC MAJOR LOSS OF EMERGENCY ASSESSMENT CAP ABILITY, OFFSITE RESPONSE CAP ABILITY, OR COJ\1MUNICATIONS CAP ABILITY
[ 10CFRSO.72(b)(1 )(v)]
RAL SNSS/EC determines that an event (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
- More than seven Offsite Sirens for > 1 hr
- Use of the EOF for> 8 hrs
- All Meteorological data (Salem AND Hope Creek) for > 8 in hrs
- Site access due to Acts of Nature (snow, flood, etc.)
MODE-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 Salem and Hope Creek and may be reported to the NRC by EITHER station as a Common Site Event.
- 1. Loss of the NETS or ENS for > 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.
Page l of2 RAL-11.7.1.b Rev. 00
SGS EALIRALTechnical Basis IF notified by the NRC Operations Officer of an inoperable ENS line, THEN NO further notification is necessary.
- 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)(I)(v)
NUREG-1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.7.1.b Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.7 Security I Emergency Response Capability REPORTABLE ACTION LEVEL - 11.7.1.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 (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:
- P250 or Aux Annunciator System for > 24 hrs
- SPDS for > 8 hrs ( > 2 CFSTs lnop, not due to plant conditions)
- Use of the TSC for > 8 hrs
- ALL Plant vent radiation effluent monitors for > 8 hrs
- 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 MODE- All BASIS NOTE: IF losses are part of a schedu.led test or preplanned maintenance activity AND WHEN compensatory actions have been taken, THEN NO report is required.
- 1. Loss of the P250 or Aux Annunciator System for a prolonged time is considered a loss of emergency assessment capability.
- 2. 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.
Page I of2 RAL - 11. 7.1.c Rev. 00
SGS EAL/RALTechnical Basis
- 3. Loss of SPDS for > 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ( > 2 CFSTs lnop, not due to plant conditions) is considered an event that significantly impairs safety assessment capabilities.
- 4. Loss of EROS, NRC phone line, Modem need to be reported to the NRC, so the NRC can have them repaired.
- 5. Loss of ALL Plant Vent Eftluent Radiation monitors (R41A, B, C and R45B & C) 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.
- 6. 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.
- 7. Concurrent multiple accident or emergency condition indicators which in the judgment*
of the SNSS significantly impairs assessment capabilities is specific to Salem in this RAL.
IF the loss of assessment capability is COMMON to both Salem and Hope Creek, THEN REFER to RAL 11.7.1.b.
REFERENCES 10CFR50.72(b )( 1)(v)
NUREG-1022, Rev. 1, 2nd Draft Page 2 of 2 RAL - 11.7.1.c Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.8 Public Interest REPORTABLE ACTION LEVEL- 11.8.2.a IC UNUSUAL CONDITIONS WARRANTING A NEWS RELEASE OR NOTIFICATION OF GOVERNMENT AGENCIES [10CFR50.72(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 Loca( State or Federal agency has been or will be made MODE-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.
Pagel of2 RAL - 11.8.2.a Rev. 00
SGS EAL/RALTechnical Basis REFERENCES 10CFRSO. 72(b )(2)(vi)
NUREG l 022, Rev. 1, 2nd Draft.
Page 2 of2 RAL - 11.8.2.a Rev. 00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.8 Public Interest REPORTABLE ACTION LEVEL- 11.8.2.b IC UNUSUAL CONDITIONS DIRECTLY AFFECTING LOWER ALLOWAYS CREEK TOWNSHIP (LACT) [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.
MODE-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, THEN 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. 0. U.
REFERENCES LAC-M.O.U.
Page I of I RAL - 11.8.2.b Rev. 00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/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 MODE-All 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, dilution or mis-operation.
This condition can be detected from the Control Room using available Nuclear Instrumentation by observation of a sustained positive startup rate on the Source or Intermediate Range Nls.
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 1of1 RAL-11.9.1.a Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/Rad Material Shipments - Releases REPORTABLE ACTION LEVEL- 11.9.1.b IC LOSS AND INVESTIGATION OF THE LOSS OF SPECIAL NUCLEAR MATERIALS/ SPENT FUEL [10CFR73.27(c), 10CFR73.71(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 oflost or unaccounted for SSNM shipment are received MODE-All BASIS This RAL addresses those conditions requiring a one hour report IAW 10CFR73.27(c) or 10CFR73.71(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) 10CFR73.71(a)(1) requires a one hour report ofa shipment loss, and on recovery of a lost shipment.
10 CFR 73.27(c) requires an immediate trace investigation of lost or unaccounted for shipments and reporting in accordance with 10CFR73. 71.
Page I of2 RAL - 11.9.1.b Rev.00
SGS EALIRAL Technical Basis REFERENCES 10CFR73 .27( c) 10CFR73. 71 (a)
Page 2 of2 RAL-11.9.1.b Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/Rad Material Shipments - Releases REPORTABLE ACTION LEVEL - 11.9.1.c IC THEFT OR LOSS OF LICENSED MATERIAL [IOCFR20.220l(a)(l)(i)]
RAL Lost, stolen or missing licensed material > I 000 times the quantity specified in I OCFR20 Appendix C, in such circumstances that an exposure could result to persons in Unrestricted Areas.
MODE-All BASIS This RAL addresses those conditions requiring an immediate report IAW 10CFR20.220l(a)(l)(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 IOCFR20.
Unrestricted Areas are any areas beyond the Minimum Exclusion Area (MEA). (outside the Owner Controlled Area (OCA) boundary)
REFERENCES IOCFR20.220l(a)(l)(i)
Page 1of1 RAL - 11.9. l.c Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/Rad Material Shipments - Releases REPORT ABLE ACTION LEVEL - 11.9.1.d IC RECEIPT OF SSNM MATERIAL [10CFR73.27(b)]
RAL Receipt of shipment of Strategic Special Nuclear Material (SSNM)
MODE-All BASIS This RAL addresses, in part, the conditions requiring an immediate report in accordance with IOCFR73.27(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 IOCFR73.27(b)
Page 1of1 RAL - 11.9. l.d Rev. 00
SGS EAL!RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/Rad Material Shipments - Releases REPORT ABLE ACTION LEVEL - 11. 9.1.e IC EXCESSIVE CONTAMINATION AND/ OR RADIATION LEVELS ON A PACKAGE
[10CFR20.1906(d)]
RAL Receipt survey indicates that package contamination I radiation levels equal or exceeds ANY one of the following:
- 2200 dpm/100 cm2
- 200 mR/hr on contact
- 10 mR/hr at 3 feet MODE-All BASIS This RAL addresses the conditions requiring an immediate report IAW 10CFR20.1906( d).
This requirement refers to values provided in 10CFR71.87(i)(l) for contamination, and to 10CFR71. 4 7 for radiation levels.
The RAL contamination level is based on the limit, adjusted for the standard swipe area used at Salem Generating Station. 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 10CFR20. 1906(d) 10CFR71.4 10CFR71.47 10CFR71.87(i)( 1)/(2)
Page 1of1 RAL-11.9.1.e Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special* Nuclear Material
/Rad Material Shipments - Releases REPORTABLE ACTION LEVEL - 11.9.2.a IC ACCIDENT OCCURRING DURING TRANSPORTATION OF LICENSED MATERIAL [10CFR71.5(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.
MODE-All BASIS 10CFR71.5(a)(l)(v) refers to 49CFR171.15116 for transportation oflicensed 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)(l)(v) 49CFR171.15/16 NC. NA-AP .ZZ-0024(Q)
Page 1of1 RAL - 11.9.2.a Rev. 00
SGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality/Special Nuclear Material
/Rad Material Shipments - Releases 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 (> 100 ft 2)
MODE-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 effluents. 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
SGS EALIRALTechnical Basis REFERENCES Commitment#: EP95-002 10CFRSO.72(b)(2)(vi)
.NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.9.2.b Rev. 00
SGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.10 Voluntary Notifications REPORTABLE ACTION LEVEL- 11.10.2 IC EVENTS/CONDITIONS WARRANT VOLUNTARY/COURTESYNRC NOTIFICATIONS [IOCFR50.72 - VOLUNTARY REPORT]
RAL In the judgment of the SNSS, notification to the NRC is warranted NO other EALs or RALs appear to be applicable MODE-All BASIS Salem 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 I 0 CFR 50. 72 reporting criterion.
Salem 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 1 of2 RAL - 11.10.2 Rev. 00
SGS EALIRALTechnical Basis an evaluation should generally proceed on a schedule commensurate with the safety significance of the question.
REFERENCES Commitment #: EP95-001 Salem 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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Pg. 2 of 4 SALEM EVENT CLASSIFICATION GUIDE TABLE OF CONIENTS/SIGNATIIRE PAGE January 21, 1997 TITLE REV# f AGES DATE 10.0 Reserved for future use NIA 11.0 Reportable Action Levels (RALs) 11.1 Technical Specifications 00 3 01/21/97 11.2 Design Basis/ Unanalyzed Condition 00 2 01121/97 11.3 Engineered Safety Features (ESF) 00 1 01/21/97 11.4 Personnel Safety/Overexposure 00 2 01/21/97 11.5 Environmental 00 1 01121/97 11.6 After-the-Fact 00 1 01./21/97 11.7 Security/Emergency Response 00 1 01121/97 CapabiJities 11.8 Public Interest 00 1 01/21/97 ll .9 Accidental Criticality/ 00 2 01/21/97 Special Nuclear Material/
Rad Material Shipments - Releases
- 11. JO Voluntary Notifications 00 1 01/21/97 WC Salem ECG Charts (Located In ER.Fs)
SGS Rev.00
ECG T.O.C.
Pg. 3of4 SALEM 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 01/21197 10 1 Hr Report - NRC Regional Office 00 3 01/21/97 11 1 Hr Report (Common Site) Security 00 3 01/21/97 Safeguards 12 1 Hr Report - NRC Operations 00 3 01/21/97 13 4 Hr Report - Contaminated Events Outside 00 7 01/21197 OfTheRCA 14 4 Hr Report - NRC Operations 00 3 01121197 15 Environmental Protection Plan 00 3 01/21/97 16 Spill I Discharge Reporting 00 7 01/21197 17 4 Hr Report - Fatality or Medical 00 4 01/21197 Emergency 18 4 Hr Report - Radiological 00 4 01/21197 Transportation Accident 19 24 Hr Report - Fitness For Duty (FFD) 00 3 01/21197 Program Events 20 24 Hr Report - NRC Regional Office 00 3 01/21197 21 Reportable Event - LAC/ Memorandum Of 00 2 01/21/97 Understanding (M.O.U.)
22 T/S Required Engineering Evaluation 00 2 01121/97 23 Reserved 24 UNUSUAL EVENT (Common Site) 00 10 01/21197 25 1 Hr Report (Common Site) - 00 3 01121/97 Major Loss of Emergency Assessment, Offsite Response, OR Communications Capability SGS Rev. 00
ECG T.O.C.
Pg. 4of4 SIGNATURE PAGE
/-~-91 Prepared By:
(If Editoria ~
Section/Attachments Revised: _ ____.A........_'"""L_L-
_ _ _ _ _ _ _ _ _ _ _ __ 1-t?/- 91 (List Non Editorial Only- Section/Attachments) Date
~////'/~
Reviewed By: _ _,,,~~'---L--a,___~--"--.,,,.,.~~~:...=..------------
,. Station Qualified Reviewer
/-/<;-/1 Date 1-1 c-22 Emergency Preparedness Manager Date Reviewed By: ---------"-1~._/--'"4_._____________
Director - QA/Nuclear Safety Review Date (If Applicable)
SORC Review and Station Approvals Mtg. No. Salem Chairman
,,,t-/o -'T'-9-- /-/6*7~
Date Date Effective Date of this Revision: 01/21/97 Date SGS Rev.00
ECG Section i Pg 1of10 SALEM EVENT CLASSIFICATION GUIDE INTRODUCTION & USAGE Section i
\O\ _ . .-
.... ~
I. PURPOSE OF THE 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:
- 1. 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) I Site Area Emergency (SAE)
- 1 General Emergency (GE) Most Severe B. Unusual Event:
- 1. 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.
SGS 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 normal 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:
- 1. 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 I 000 mRem TEDE in Unrestricted Areas.
SGS REV. 00 I
l
ECG Section i Pg 3of10 Ill. EVENT CLASSIFICATION GUIDE (ECG) STRUCTURE A. Overall Layout: The ECG is divided into 4 segments which are:
- 1. Front Matter: Information which includes the Table of Contents, Introduction & Usage, a Glossary of Acronyms, and Critical Function Status Trees (CFSTs).
- 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/non-emergency notification, reporting requirements, references and forms required to facilitate event communications.
- 4. ECG Chart: Wall chart (Locates! 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 CIC): 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. MODE: Refers to the Operational Mode at Salem during which a particular IC/EAL is applicable. The Mode that the plant was in when the event started, prior to any protection system or operator actions, should be utilized when classifying events.
(from SGS Technical Specifications, Sect. 1, Definitions)
THERMAL MODE Ketr POWER* TAVG
- 1. POWER OPERATION ::: 0.99 >5% ;::350 °F
- 2. STARTUP ::: 0.99 S5% ::: 350 °F
- 3. HOT STANDBY < 0.99 0 ::: 350 °F
- 4. HOT SHUTDOWN < 0.99 0 > 200 °F & < 350 °F
- 5. COLD SHUTDOWN < 0.99 0 s 200 °F
- 6. REFUELING** < 0.95 0 s 140 °F SGS REV. 00
ECG Section i Pg 4of10
- Excluding Decay Heat
- 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 UE 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 1, then RAL is 1 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 SGS 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 CRAL): A predetermined, site-specific, observable threshold used to define when the generic initiating condition ha:s been met, placing the plant in a given emergency class or non-emergency report. An EALIRAL 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 Attachm~nt 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.
- 2. MODE: Refers to the Operational Mode at Salem during which a particular IC/EAL is applicable. The Mode that the plant was in when the event started, prior to any protection system or operator actions, should be utilized when classifying events.
(from SGS Technical Specifications, Sect. 1, Definitions)
SGS REV. 00
ECG Section i Pg 6of10 THERMAL MODE Kerr POWER* TAvG
- 1. POWER OPERATION 2: 0.99 >5% 2: 350 °F
- 2. STARTUP 2: 0.99 ::; 5 % 2: 350 °F
- 3. HOT STANDBY <0.99 0 2: 350 °F
- 4. HOT SHUTDOWN < 0.99 0 > 200 °F & < 350 °F
- 5. COLD SHUTDOWN < 0.99 0 ::; 200 °F
- 6. REFUELING ** < 0.95 0 ::; 140 °F
- Excluding Decay Heat
- 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 GlTIDE (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 IAW 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 IAW the IC.
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.
SGS REV. 00
~--
ECG Section i Pg 7of10 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 assessment 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 PARs, etc.
They should be followed in a step-by-step fashion.
The ECG should be considered an "Implementing Procedure" and used in accordance with the requirements of a "Category II" procedure as defined in NC.NA-AP.ZZ-000 I (Q). The ECG'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 EALIRAL 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)
SGS 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. CONTfNUE 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 t_herefore 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-El\fERGENCY CONDITIONS - 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 classification and implementation of a non-emergency report (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 4 SGS REV. 00
ECG Section i Pg 9of10 hour, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), 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 pro\*ided in Attachment 5 of the ECG.
F. Voluntao1/Courtesy Reporting of Non-Emergency Events Guidance In accordance with NUREG 1022, Rev 1, voluntary reporting is encouraged.
PSE&G may make voluntary or courtesyNRC 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 10CFRSO.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. OBTAIN both the Salem General Manager's and Operations Manager's approval of any proposed retractions.
SGS REV. 00
ECG Section i Pg 10of10
- 2. COMPLETE "page l" of the NRC Data Sheet which was implemented to make the original notification. Event Description Section ofNRC Data Sheet should explain the rationale for the retraction.
- 3. CONT ACT the NRC Operations Center using the ENS and provide the "NRC 'Data Sheet" information.
- 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:
- l. COMPLETE 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 Events 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.
SGS REV. 00
ECG Section ii Pg. 1of6 SALEM, EVENT CLASSIFICATION GUIDE Glossary of Acronyms & Abbreviations Section ii AAAG Accident Assessment Advisory Group (Delaware)
AB Auxiliary Building AC Alternating Current AFST Auxiliary Feedwater Storage Tank AFW Auxiliary Feedwater ALARA As Low As Reasonably Achievable ARM Area Radiation Monitor AS Administrative Supervisor ASAP As Soon As Possible ASM Administrative Support Manager ATWT Anticipated Transient Without Trip BIT Boron Injection Tank BKGD Background BKR Breaker (electrical circuit)
BNE Bureau of Nuclear Engineering (NJDEPE)
CAS Central Alarm Station CCPM Corrected Counts per Minute CDE Committed Dose Equivalent CEDE Committed Effective Dose Equivalent CET Core Exit Thermocouple CFCU Containment Fan Coil Unit CFR Code of Federal Regulations CFST Critical Safety Function Tree CMl Primary Communicator (CR)
CM2 Secondary Communicator (CR)
CNTMT Containment (Barrier)
CP Control Point CPM Counts Per Minute CR Control Room CRD Control Rod Drive DC Direct Current DDE Deep Dose Equivalent DEi Dose Equivalent Iodine DEMA Delaware Emergency Management Agency SGS Rev. 00
ECG Section ii Pg. 2of6 DEP Department of Environmental Protection (NJ)
DID Direct Inward Dial (phone system)
DOE Department of Energy DOT Department of Transportation DPCC/DCR - Discharge Prevention, Containment, & Countermeasures/
Discharge Cleanup & Removal Plan DPM Decades per Minute DPM Disintegrations per Minute DRCF Dose Rate Conversion Factor EACS ESF Equipment Area Cooling System EAL Emergency Action Level BAS 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 Procedures EPA Emergency Preparedness Advisor EPA Environmental Protection Agency EPIP Emergency Plan Implementing Procedure EPM Emergency Preparedness Manager EPZ Emergency Planning Zone ERDS Emergency Response Data System ERF Emergency Response Facility ERM Emergency Response Manager ERO Emergency Response Organiz:ation ESF Engineered Safety Feature ESSX Electronic Switch System Exchange (Centrex)
FC Fuel Clad (Barrier)
FFD Fitness For Duty FHB Fuel Handling Building FPB Fission Product Barrier FRCC Functional Restoration Core Cooling FRCE Functional Restoration Containment Environment FRCI Functional Restoration Coolant Inventory SGS Rev. 00
ECG Section ii Pg. 3 of 6 FRERP Federal Radiological Emergency Response Plan FRHS Functional Restoration Heat Sink FRSM Functional Restoration Shutdown Margin FRTS Functional Restoration Thermal Shock FfS Federal Telecommunications System (NRC)
GE Geneni.I Emergency HEPA High Efficiency Particulate Absorbers HP Health Physics HVAC Heating, Ventilation & Air Conditioning HX Heat Exchanger IAW In Accordance With IC Initiating Condition ICMF Initial Contact Message Form IDLH Immediately Dangerous to Life and Health IR Intermediate Range I/S In Service ISOL Isolation 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 LPZ Low Population Zone MDA Minimum Detectable Amount MEA Minimum Exclusion Area MEES Major Equipment & Electrical Status (Form)
MET Meteorological MIMS Metal Impact Monitoring System MOU Memorandum of Understanding MRO Medical Review Officer MSIV Main Steam Isolation Valve MSL Main Steam Line SGS Rev. 00
ECG Section ii Pg. 4 of 6 NAWAS National Attack Warning Alert System NCO Nuclear Control Operator NDAB Nuclear Department Administration Building (fB2)
NEO Nuclear Equipment Operator NETS Nuclear Emergency Telecommunications System NFE Nuclear Fuels Engineer NFPB Normal Full Power Background NG Noble Gas NJSP* New Jersey State Police NOAA National Oceanographic and Atmospheric Administration NR Narrow Range NRC Nuclear Regulatory Commission NSP Nuclear Site Protection NSS Nuclear Shift Supervisor 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 OHA Overhead Annunciators OSB Operational Status Board (Form) osc Operations Support Center PAG Protective Action Guideline PAR Protective Action Recommendation PASS Post Accident Sample System PIM }>ublic Information Manager PMP Pump PORV Power Operated Relief Valve PSIG Pounds per Square Inch Gauge PWST Primary Water Storage Tank PZR Pressurizer RAC Radiological Assessment Coordinator RAD Radiation RAL Reportable Action Level RC Reactor Coolant RCA Radiologically Controlled Area RCAM Repair and Corrective Action Mission RCP Reactor Coolant Pump SGS Rev. 00
ECG Section ii Pg. 5of6 RCS Reactor Coolant System (Barrier)
RHR Residual Heat Removal RM Recovery Manager RMO Recovery Management Organiz.ation RMS Radiation Monitoring System RPS Radiation Protection Supervisor RPS Reactor Protection System RSM Radiological Support Manager RVLIS Reactor Vessel Level Instrumentation System RWST Refueling Water Storage Tank SAE Site Area Emergency SAM Severe Accident Management SAS Secondary Alarm Station (Security)
SAT Satisfactory SBO Station Blackout SCBA Self Contained Breathing Apparatus SCP Security Contingency Procedure SDE Shallow Dose Equivalent SDM Shutdown Margin SIG Steam Generator SGS Salem Generating Station SGTR Steam Generator Tube Rupture
- SI Safety Injection SJAE Steam Jet Air Ejector SNM Special Nuclear Material SNSS Senior Nuclear Shift Supervisor sos Systems Operations Supervisor (Security)
SPDS Safety Parameter Display System SRPT Shift Radiation Protection Technician SSCL Station Status Checklist SSE Safe Shutdown Earthquake SSM Site Support Manager SSNM Strategic Special Nuclear Material SUR Startup Rate T-COLD Temperature Cold (Leg)
T-HOT Temperature Hot (Leg)
TAF Top of Active Fuel TDR Technical Document Room TEDE Total Effective Dose Equivalent TPARD Total Protective Action Recommendation Dose T/S Technical Specifications SGS Rev. 00
ECG Section ii Pg. 6 of 6 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 VLV Valve WB Whole Body WR Wide Range SGS Rev. 00
The Salem Unit 1 Critical Function Safety Trees are still in draft and will be incorporated before Unit 1 goes to mode 4.
FIGURE 1 EOP-CFST-1 SHUTOOw~ MARGIN STATUS TREE i
I I
I POWER RANGE I LESS THAN 51.
I I
YES I ~~o I I
I I
I I
I JR SUR ZERO I
nD I V"
NF CAT l VF I YES NO I I
I I I
I I I
I SOURCE.
RANGE I ENERGIZED I l
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.:>UU~\...t. J rl .)Uri Ml.Jr\[_
R.~.NGE SUP NEG~.TJVE ZERO OR THAN NEGATIVE -0.2 OPM YES NO YES I ~JO GREEN YELLOW
...................... GREEN YELLOW I PURPLE I RED SAT rns1v1-2 SAT rnsu-2 I FRSt.;1-1 I FRSM-1 REV. z1
FIGURE 2 EOP-CF sT-*1 II CORE COOLING STATUS TREE I
I I
r
- .J rin Uri
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CE Ts I GREATER .THAN I 1200 DEGREES I YES NO I
I RCS I c:11Rrnni 11-.1r:
...JVL..'V"-'Vl-"'-'
I GREATER THAN I 0 DEGREES I YES NO I I
II I
JS ANY II RCP rl11J..IP.1Tt.lf' l\Ul'\11'\l!l't\,.i I
I YES NO I I
I 5 OR MORE I GREATER THAN CETs I 700 DEGREES I n.S NU I
I I
D\/I IC' t\,.. L 1_.1 I
[lYhJAMIC RANGE GREATER THAN:
44/ FOR 4 RCPs RVLIS RVLiS 30/ t"OR .3 RCPs FULi.. RANGE FULL RANGE
"')(\*1 L no 'I orn .::i...
.!...\.J'* VI\ .:.. I\\...~
GREATER THAN GREATER THAi~
13i: FOR 1 RCP .39i. 39/
vrr I(.) vrr It.) I "" l~U vr ,-
1 [..) ""
l\llJ RED GREUJ YELLOW
_....__ PURPLE I PURPLE RED
. YELLOW :I PURPLE
. ***~f:fr:1~:.:j***:1 FRCC-2
...._..I I
J:"RCC-1 SllT f:"RCC-3 FRCC-2 I FRCC-2 FRCC-1 SALEM UNJl 2 REV. 21
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FIGURE 4 EOP-CFS1 -1 THERMAL SHOCK STATUS TREE ALL T-COLD I
.::,ooLuowr*i RAT::s IC.:"( TUAt..!
I LL..._}..J ' I lf"".l'f 100" ~ IN LAS1 50 MINLm:s YES NO
."ii I p,..-::
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- :p, 7.~.JJ.1 TEi\*ic i:u1r,-:-s 3 ~: "} F T(* T~~ R!(~H"7
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I I
- .,L!_ -r * ,
- ..:r)LD:.
GK~ATE;; TH~i<
3 ~(' :> ["
'r'E5 I NC C~P.E~ i'l (;'lt:EN *::c.c_ Ow PURPLE GRt.U~ 'ft L:... Ow f="iJRF~ ~ RED S.P .)A"'" ;-RT~,-_:' :- ..
c-p-r~
',:::.- : SA: l="R13-2 FRT5-' FRT5i- I
FIGURE 5
\,,Vl't A!N&U"'l.l'T' I Mii l't'lt.1"4 I ENVIRONMENT STATUS TREE I ,..,....,llT
- 1 it.II J r l l l
- ._.UI~ 1 A PRESSURE J i\llV1L. l'l !
I L.ESS THMj 47 PSJC r
- ONT A! NMUJT PRE*ssuRE lI 1 r r r 111 A"'
LL..)...} I ~1-1.1\1 IS PS[G vr I'--'
c I r..u"l I ~U I
I CJ)r*n !>&~MEW SUMF
~E:.::; THAt; 78!.
- i:. ~c1 v~r:*st::>
YES I NC:
I R~I!~: :f~ii*-:
~r::s.:. HAr.
Hr GREE~; .,,*E.._ Jw PIJPF'_'.: PURPLE SAT ~ ~* - .., I ~Rt: E-2 I Fii::E:-1
FIGURE£ COOLANT INVENTORY STATUS TREE PZH _t. ~'!:_:.._
~ L SS TH;\~~
32%
I I
vc-=-
- 1......1 I .,,,
11*_1 I
- S A~/.
~._:::
~._.F
=l'...;N',. 'L YES RV~!S
- i; - 3 orNAMIC DANG~
GRt.Alt:.r- *'1AN
'GV
- *_1r
'."'l" _ ...... ':> r!V~ : .)
,/"'\,....*
1vu.
~ ur: ....
' 1"'"'1,-.r-._
- -c
- .1,_.* t""::i.
,......,..,,,,...r
=-c:P ~ ;::.~~::
71!. FOP 3 RCc;s .1nt""\1r Ur'!""'L_M :"'\.>-\1'H.Jt.
71*. ..JF*D~~ q~N*'JC jQ: :- =*"' ~ :{:,:?:; ~R: ,-:-~R -w.1i.~J so;. FOR :: RC?:: ~PEATEP f><.AN
]3;. FJR i ~c;: lOG'!.
- C;07o
'iES NIJ YES NC ff'.' lW iF; '.!W q~ -3 F P~. ! - '"> FRC -3 PC.!-3 YE __ JW GREEM CREE~!
i
'""°"*""*- -
- ""\ I
SGS ECG 1.0 Fuel Clad Challenge Rev. 00 Page I of!
1.1 RCS Activity Initiating Condition ~-~~~~~~~~~~~~-F_u_e_1_c_1a_d_D_e_g_r_ada-ti_o_n~~~~~~~~~~~~~~
MODE (I, 2, 3, 4, 5, 6) (I, 2, 3, 4, 5, 6) ( 1, 2, 3, 4, 5, 6)
EAL# 1.1.1.a 1.1.1.b 1.1.1.c IF IF IF E
M Reactor Coolant Activity Reactor Coolant Activity Valid Letdown Line E > I µ.Ci/gm Dose (Dose Equivalent Iodine) Monitor R Equfivalent Iodine 131 for exceeds limits in Alarm G > 48 Hours of Technical Specification (IR31A or 2R31)
E Figure 3 .4-1 N
c y
THEN A
c T
I 0 NOTE:
N Refer to Section 3. 0, Fission Product Barrier Table L pnor to Event Classification E
v E
L s
Action Refer to Auach~~~~-o Required UNUSUAL EVENT
SGS ECG 2.0 RCS Challenge .Rev. 00 Page I of I 2.1 RCS Leakage Initiating Condition MODE ( I, 2, 3, 4 ) ( 1, 2, 3, 4_) ( I, 2, 3, 4 )
EAL# 2.1.1.a 2.1.1.b 2.1.1.c IF IF IF E
M Reactor Coolant System Reactor Coolant System Reactor Coolant System E Pressure Boundary Leakage Unidentified Leakage Identified Leakage R > 10 gpm > 10 gpm >25 gpm G
E 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
Refer to Attach;e~nt Action I
Required UNUSUAL EVENT
--- -~*------
- 3.I Fuel Clad Barrier ; 3.2 RCS Barrier TABLE3.0 U.1 CBfl'ICAL 8AFZIY FUNCl10N 8TA'MJB FISSION PRODUCT &ALl~Ll.a 1.oss .. ,n...-..
EALlll.1.1.o PC1l'ENTIAL L088 * &Pr.
EAl..13.2.1.a Not~
1.oss * .fPT" Not Appllool!le BARRIERS CORE COOLING PUBPLE PNnl CORE COOLING BED PAffi THERllAL SHOCK RED PATH
.on.-----t
. \.~'>.
-OR- -t1R-BALIS.Ll.b EALAl.l.b 2A.LtS.S.J.b HEAT BINK BED PA.TB BEAT SINK BED PAm CORE COOLING BED PAmfor> 1nmmuw.
APPLICABLE &8..J CONTAINMENT PRE8BURE MODES ARE WSS*4PT* P01'ENTIAL l.088
- 1 !""'I' LOSS.:2Pf*
EA.Liii.Li EAL I 9.3.2.a EAL I 8.8.2.c 1,2,3,4 ONLY EALoll.ll.l!.a EALIB.2.2.b r----* 111----
8-dor Coolant ad.1Tft7' > 800 Ooe Cmtdfupl Charaln,s BubeooH.aa: Ill o T
- a re.ili Coatalnmeid U. > .t'l> A Rapid lJmmpWned uCL'p (k_., &qulHlebl 1-18)
Pu.p~ -.b:d.aln PZB olBC8i-bp Contalnmem Pre.uni Drop
~> 17'1oC-*~~RC8 E U. S.9.2.b CNTMTP'r--.> .lipclgwtlb t followfn.g ClD lnlttal Rlse to
>.f.pdl l!'.!lI§B. ooe cl 1 he foll'""9 DF
- No CN'IMJ' Bp-ay AND < n t!.m II tbe L:im or Potential i.e.. h oaa..tdinred IMMINENT CFCU11 Bunning~
. . . . OrOmll' wt.Udn Z Jumn), umo~ lllld cl.uslfy ..
- One CNTMT Spray Train lfthe threoholdlaomoodod. 118AND<S.CFCl!11
&\LI S.1.&a EAL 1 S.1.S.b Bun"'iiJiii In '"lDW 8rlf'ed*
a ~mare err. > '100 ep nor ID.ewe CEI'. > 1200 T Instructions: LOSS*.fPI'* &.S..B CONTAINMENT ISOLATION
- 1. In the table nmew the EmerlJmicy Action POl'EN'TIAL LOSS .. J Pl' EAL1&2.3.b LOSS,. 2P1"t1 Levein of all colunuul and Identify which EAL ' 3.2.lla OneC-~~ One~~ EA.Lt S.S.S.a need further review. Pomp CANNor maintain PZB Pump CANNor motntaln kAL*S.S.9.h CNTMT 8omp Lln'el > 78'1> ValldCNTMl'eA.eBor 9.1.4 BI VP'.8ID!:L LEVEL INDICATION 8YB'l'EM CBVU8) Jen!> J 'l'i> (u a nsnlt of* PZR~ (uancnll ('1M.adv<<.e) CNTMT Vent laol Btgnal
- 21. For ""ch or the U..- banlen, determine the 8GTR) of*BQTR)
AND PCJJ>>tTlAL 1088. 8PTm LO~S" EAL with the highest Point value, end circle OPT,. AND AND y.o; path hum CNTMT to the CC'iO'"trol Rocm h.89 detotml.ntd ~ Bt:Mm Genenato:r ePT\l'OOII:lent the coneepondlng EAL I and point vnlue. JL\L I S.1.La NotApJ>l.lcahl..
B\rUB Full aui. c 89'&. that an SGTR ham OClC'Urnld ~bi dropping in an No more than one EAL llbaa1d be Relecled uncontrolled. manner or for each barrier. BAL.S.L4.b completely depreesurhed llVUBJ>y..-RaDso AND S.S.'- RC8 LINE BREAKiCONTAINMF.NT BYPA88 ll. Add the point vain.,. alroled for the three IDdfoate. 4D]" one of the Pn>lougod. ~ ...,.,..wy POTENTIAL WSS
- J PT 1~oss~2rra barrlel'll nnd enter the mm below1 I~ 1Mkajpe to the environD1ent
- 9~118<80'lt ducll open .afety or relief Uoiaolable, Faulted a&ea.m Prlma:y to-dary 1-~
- IRCl'mlfB<iO"l> ..iv- NOI'E< SKE 8.8.-Lb GeDflnlltor OU1'81DE d >Tocb Bpl'elJmlt*
- IRCl'91/8<18'1> Ccbtahune~ bJ Ml!
&'Gp~ dropping in an Pro1oo~ dlr.d llll'C'l>DdAry
- 4. Claamlfy- on the polni value lllllID .., nDC<<>ntrolled DlUlJler or 1-bp lo lOO eo..-lronme>nl ronow.. UA CONTAINMENT RADIATION LEVEL8 completely depre-urlzed
, __ _ _ **It ----t
~
Uthe Cla...ityam Refer to: l'OI'EN11AL LOSS
- SP'I'*
Not Applleable LOSS *4.PI'*
EALls.iL*
Affected Ml tobeB anl lulad Valid CoDlalmnent Bad.laUoo 1,!l UNUSUAL EVENT Atlaclunent 1 leTel wbleh a~ ANY cme ol. lhe rono.tn, eoaiaiii'ment Bad.Monlion~
M ALERT Att.climent~
- mil> lBJhr 8.8.8 CONTAINMENT RADIATION LEVELS
- ~>lOBJh:r 11,8,7,8 SITE AREA AtiaclunODt 8
- R'4B > 10 R/hr POIVfl1AL l.OSS "' JPT EAL I 8.3..n NotApplfc.able 9,10 GENERAL Att..ahment4 BA4A. or &448 > ZOOO Rlhr
- 11. Implement the appropriate ECG S.S.8 EMERGENCY COORDINATOR.JUDGMENT Attachment per above cbarL 8.1.ll DIEBGENCY COORDINATOR .roooMENT POn:NrlAI. L088. 1Pf F01ZN11AL L088 .......
- 8. Contlnne to nmew the ZALi on thla Table EAL I S.S.6.a EAL I S.S.&.b for cbanil.. that couJd remit In l!Dlerilmtcy Mn' omidHlon,, bl lbe oplnlon .Mi! coudHlon,. ln &he opbilcm
_,,.lntlon or deeacalatlon. ol tho Fe. lhat bxllcataa
- cl lbe EC. that. lDdlcmea a Le.
Poteotl.al 1..- al the ol lbe Ccdalnmem Ban-le!'
Cont..s.nml!lal Barrlw
SGS ECG 4.0 EC Discretion Rev. 00 Page I oft 4.1 Emergency Coordinator Discretion Other Conditions Exist Other Conditions Exist *-,) . '()th-~~-Conditions Exist Whic~ Other Conditions Exist Which Which In the Judgment of the Which ln the Judgment of the In th..: J ndgment. 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 Declaration of an Unusual Event Declaration of an Alert / ' a Site Area Emergency a General Emergency
--- ---------~
MODE ( All ( All ') All ) All )
--~ '*-----~
EAL# 4.1.1 4.1.2 4.1.3 4.1.4
1 IF IF IF IF E -----1 M
Events are in progress or Events are in progress or Events are in progress or Events are in progress or E
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 judgment of the Emergency judgment of the judgment of the Emergency G
Coordinator, indicate a Coordinator, indicate Emergency Coordinator, Coordinator, indicate an E
N Potential Degradation EITHER one of the indicate an Actual or imminent c of Plant Safety following: Actual or likely major substantial core
- y
- Plant safety systems failure of plant functions degradation with the needed for protection of potential for loss of THEN (more than one) are or A the public containment may be degraded c
- ANY Plant Vital T *--~ -* ----.------**-----~
I Structure is degraded or 0 potentially degraded THEN THEN N
AND L
E Increased monitoring of v Safety Functions E is warranted L
THEN s
,, 1*
Action Refer to Attachment 1 Refer to Attachment 2 ,--- Rc~cr to Attachment 3 Refer to Attachment 4 Required UNUSUAL EVENT ALERT l_ SITE AREA EMERGENCY GENERAL EMERGENCY
SGS ECG 5.0 Failure to Trip Rev.00 Page I of I 5.1 ATWT Failure of the RPS to Complete an Automatic Failure of the RPS to Successfully Trip and Manual Trip was not successful and ~~e Failure of the RPS to Successfully Complete a Initiating Complete a Reactor Trip (Automatic and Manual) is Indication of an Extreme Challenge to the Ab1hty Reactor Trip (Automatic or Manual) and Reactor Power is Above 5%
Condition to Cool the Core
( 1, 2, 3 ) ( 1, 2, 3 ) ( 1, 2 ) ( 1, 2 )
MODE EAL# 5.1.2.a 5.1.2.b 5.1.3 5.1.4 IF IF E
M Reactor Protection ANY Manually E System Trip Initiated Reactor R Setpoint Exceeded Trip from the G Control Room I AND is NOT E
N An Automatic Confirmed c Reactor Trip y is NOT Confirmed A
c ALL Reactor Trip attempts from the T AND Control Room DID NOT reduce I
(and maintain) Reactor Power to < 5%
0 N EITHER one of the following:
AND
~*------~
- CORE COOLING RED path L
E I THE:
- HEAT SINK RED path v
E THEN L
s .. '
Refer to Attachment 2 Refer to Attachme~ Refer to Attachment 4 Action ALERT SITE AREA EMERG~~ GENEJRALEMERGENCY Required
SGS ECG 6.0 Radiological Releases/Occurrences Rev.00 Page I of4 6.1 Gaseous Effluent Release y Unplanned Release of Gaseous Radioactivity to Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds Initiating Environment that Exceeds 2 Times the Radiological 2 times the 10CFR20, Appendix B limits for 60 minutes or longer Condition Technical Specifications for 60 minutes or longer MODE ( All ) ( All ) ( All ) ( All )
EAL# 6.1.1.a 6.1.1.b 6.1.1.c 6.1.1.d Field Dose IF Measured IF Sample IF Alarm IF Assessment Dose Rate AnalL--is_ _ __ Indications E
M Dose Assessment Gaseous efflue nt release Valid Plant Vent Effluent Alarm Dose Rate measured at E indicates EITHER one sample ana lysis on the Protected Area R of the following at the EITHER one of the Boundary or beyond AND G MEA or beyond as following ind icates a EXCEEDS E calculated on the SSCL: concentrat ion of:
.05 mRem/hr above Release rate EXCEEDS 9.68E+04 pCi/sec N
- 2:::, 2.56E-03 µCi/cc
- TEDE 4-Day Dose normal background c Total No hie Gas Total Noble Gas y 2:::, 2.0E-01 mRem
- Thyroid-COE Dose .
- 2:::, 3.71E-08 µCi/cc A 2:::, 6.SE-01 mRem 1-131 c based on Plant Vent T effluent sample analysis I and NOT on a default I AND 0 Noble Gas to Iodine N Ratio Assessment results NOT available L I
'-----------~---------.--
.. I ..
E \AND v
E L
\ Release is ongoing f~~~ ~-~minutes I s lTHEN Action ~-R-e_fi_e_r-to-A~tt-a-~h~;I]nt I Required UNUSUAL EVENT
--~---
SGS ECG 6.0 Radiological Releases/Occurrences Rev.00 Page 2 of 4 6.1 Gaseous Effluent Release Any Unplanned Release of Gaseous Radioactivity Any UnplannedRelease of Gaseous Radioactivity to the Environment that Exceeds Initiating to Environment that Exceeds 200 Times Radiological Tech 200 times the 10CFR20, Appendix B limits for 30 minutes or longer Condition Specs for 15 minutes or Longer MODE ( All ) ( All ) ( All ) ( All )
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 Dose Rate Anal is Indications E Assessment M Dose Rate measured at Gaseous effluent release Valid Plant Vent Effluent Alarm E the Protected Area sample analysis on Dose Assessment indicates R Boundary or beyond EITHER one of the EITHER one of the G EXCEEDS folowing indicates a following E 5 mRem/hr concentration of: AND at the MEA or beyond N as calculated on the SSCL: * ~ 2.56E-Ol ,..Ci/cc Total c Release rate EXCEEDS
- TEDE 4-Day Dose Noble Gas 9.68E+06 p.Ci/sec y
~ 2.0E+Ol mRem * ~ J.71E-06 ,...ci/cc 1-131 Total Noble Gas A
- Thyroid-COE !Dose c ~ 6.8E+Ol mRem T based on Plant Vent effluent I sample analysis and NOT on 0 a default Noble Gas to N Iodine Ratio AND Dose Assessment results NOT available L
E I v I AND AND E
L I Release is ongoing for~ 15 minutes I Release is ongoing for ~ 30 minutes s l~~~~~~~~~~~~~-~--
i THEN Action Required Refer t~t~;~_m:nt 2 I
SGS ECG R~v. 00 6.0 Radiological Releases/Occurrences Page 3 of 4 6.1 Gaseous Effluent Release Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem Initiating Total Effective Dose Equivalent (TEDE) or 500 mRem Thyroid-COE Dose for the Actual or Projected Duration of the Release Condition MODE (~_All_) ( All ) ( All
~-~---~*
) ( All )
EAL# 6.~.3.a Field 6.1.3.b 6.1.3.c 6.1.3.d Field Survey Alarm Dose Measured Assessment IF Dose Rate IF Analysis IF Indications IF E
M Dose Assessment indicates Dose Rate measured at Analysis o f field survey Valid Plant Vent Effluent Alarm E EITHER one of the the Protected Area samples at the Protected R following at the MEA or Boundary or beyond Area Boun dary indicates beyond as calculated on the EXCEEDS EITHER one of the AND G
E SSCL: 100 mRem/hr foll owmg:
N
- TEDE 4-Day Dose * ~ 4.36E+02 CCPM Total Plant Vent release rate EXCEEDS c ~ 1.0E+02 mRem AND 1. 7E-f-09 µCi/sec Total Noble Gas y * =::, 3.85E-07 µCi/cc
- Thyroid-COE Dose Release is expected to I-131 continue for AND A ~ 5.0E+02 mRem
~
c based on Plant Vent effluent ~ 15 minutes Dose Assessment results NOT available T sample analysis and NOT *on I a default Noble Gas to 0 Iodine Ratio AND N
L I Release is ongoing for~ 15 minutes I
E v
E L
s Refer to Attachment 3 Action Required SITE AREA EMERGENCY
6.0 Radiological Releases/Occurrences SGS ECG Rev.00 6.1 Gaseous Effluent Release Page 4 of 4 Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem Total Effective Dose Initiating Condition Equivalent (TEDE) or 5000 mRem Thyroid-COE Dose for the Actual or Projected Duration of the Release MODE ( All ) ( All ) ( __~ ( All )
EAL# 6.1.4.a Field 6.1.4.b 6.1.4.c 6.1.4.d Dose IF Measured IF Field Survey IF Alarm IF Assessment Dose Rate Anal i s _- - - Indications E
M Dose Assessment indicates Dose Rate measured at Analysis of field survey Valid Plant Vent Effluent Alarm E EITHER one of the the Protected Area samples at the Protected R following at the MEA or Boundary or beyond Area Bou ndary indicates G beyond as calculated on the EXCEEDS EITHE Rone of the AND E SSCL: 1000 mRem/hr fol lowing:
N
- TEDE 4-Day Dose * > 4.36E+03CCPM Total Plant Vent release rate EXCEEDS c ~ 1.0E+OJ mRem AND
- ~ 3.85E-06 µCi/cc I. 7E+IO µ.Ci/sec Total Noble Gas y
- Thyroid-COE Dose Release is expected to 1-13 l continue for AND A ~ 5.0E+03 mRem c based on Plant Vent effluent ~ 15 minutes Dose Assessment results NOT available T sample analysis and NOT on I a default Noble Gas to 0 Iodine Ratio AND N
Release is ongoing for ~ 15 minutes L
I I E
v E THEN L
s Action Required Refer to Attachment GENERAL EMERGENCY hJ
SGS ECG 6.0 Radiological Releases/Occurrences Rev.00 Page I of!
6.2 Liquid Effluent Release Any Unplanned Release of Liquid Radioactivity Any Unplanned Release of Liquid Radioactivity Initiating to the Environment that Exceeds Two Times the Radiological to the Environment that Exceeds 200 Times the Radiological Condition Technical Specifications for 60 Minutes or Longer Technical Specifications for 15 Minutes or Longer MODE c All 6.2.1
)
c All )
EAL# 6.2.2 IF E
M Valid Alarm from ANY one of the following RMS Channels:
E
- Containment Fan Coil Process (R13)
R G
- Liquid Radwaste Disposal Process (R 18)
E
- Steam Generator Blowdown Process (R 19)
N c *Chemical Waste Basin Process (2R37) y A
c + THEft T AND AND I
Sample analysis of liquid effluent indicates concentration Sample analysis of liquid effluent indicates concentration 0
in excess of 2 times Tech. Spec. limits in excess of200 times Tech. Spec. limits N
AND AND L
E Release continues for ~ 60 minutes Release continues for ~ 15 minutes v after the alarm occurs after the alarm occurs E
L THEN THEN s
- Ir Action Refer to Attachment l Refer to Attachment 2 Required UNUSUAL EVENT ALERT
SGS ECG 6.0 Radiological Releases/Occurrences Rev.00 Page I of I 6.3 In-Plant Radiation Occurrences Initiating Release of Radioactive Material or Increases in Radiation Levels Condition Unplanned Increase in Plant Radiation Within the Facility that Impedes Operation of Systems Required to Maintain Safe 0 erations or to Establish or Maintain Cold Shutdown MODE ( All ) /Ail)
'---~*-*-~---/
c All )
EAL# 6.3.1. 6.3.2.a 6.3.2.b
~F IF IF.
E M Unplanned rise in radiation levels inside Unplanned Dose Rate Unplanned radiation levels E the Protected Area ~ 1000 times normal > 2000 mRem/hr > 15 mRem/hr in EITHER R as indicated by EITHER one of the in any area of the plant one of the following:
G following: which requires ACCESS to E maintain plant safety functions
- The Control Room N
- Per!llanent or portable (excluding the Control Room or
- General Area Radiological Survey one of the following:
A
- Permanent or portable Area c THEN Radiation Monitors T
- General Area Radiological l --------
I Survey 0
N L
E THEN v
E L
s Action Refer to Attachment I Refer to Attachment 2 Required UNUSUAL EVENT ALERT
- - _ _ _ _ _ _ _ _ _ _ _ _ ___J
6.0 Radiological Releases/Occurrences SGS ECG Rev.00 6.4 Irradiated Fuel Event Page I of 2 c~~~-U~np_l_ann~e_d_I_nc_r_ea_s_e_in_P_IantRa_d_ia_t_io_n_ _ _~
Initiating Condition MODE ( All )
( 6 )
EAL# 6.4.1.b 6.4.1.a E IF IF M Valid SFP Low E An uncontrolled Level alarm -
R level drop in the OHA C-35 G Refueling Cavity as E indicated by EITHER AND N one of the following:
c
- Visual observation y Visual observation of an uncontrolled
- RVLIS- Refueling A level drop in the Mode c Spent Fuel Pool T
I 0
N THEN L
E Note:
v Refer to Rad E * *
- Dose Rate EALs L pnor to classification s
Action Refer to Attachm;~t_l_J Required UNUSUAL EVENT
6.0 Radiological Releases/Occurrences SGS ECG Rev.00 6.4 Irradiated Fuel Event Page 2 of2 Initiating Major Damage to Irradiated Fuel or Loss of Water Level that has or will Condition result in the Uncovering oflrradiated Fuel Outside the Reactor Vessel MODE EAL#
c All ) c All ) c All ) c. All )
6.4.2.a 6.4.2.b 6.4.2.c 6.4.2.d E IF IF IF IF M
Major Damage to Irradiated Major Damage to Unplanned rise on Visual observation of E
Fuel reported in the Fuel Irradiated Fuel reported in ANY one of the Irradiated Fuel R
Handling Bldg the Containment following Area Rad uncovered G
E N
l AND 'AND monitors or by general area rad survey indicates c Valid High Alarm is Valid High Alarm received y on ANY one of the 2:.2000 mRem/hr:
received on EITHER one of following RMS channels:
- R2 the following RMS channels:
A *R2
- RS *RS c *R32A
- RIOA T
- RIOB
- R9 I
0 I AND 'AND
- R32A N Valid High Alarm received Valid High Alarm received from from ANY one of the L EITHER one of the following RMS channels:
E following RMS channels:
v
- RIIA E
- R4l
- R12A L
- R4S s
- R12B I I --*
Action l THE N Refer to Attachme Required ALERT
7.0 Electrical Power SGS ECG Rev. 00 7.1 Loss of AC Power Capabilities Page I ofl AC Power Capability to Vital Buses Reduced to a Loss of All riftiitc Power and All Onsite AC Power to 4 KV Loss of AIU Offsite Power to Vital Buses for Single Power Source for Greater than 15 Minutes such that Vital Buses While the Plant is in Cold Initiating Greater Than lS Minutes any Additional Single Failure Would Result in Shutdown, Refueling CIC Dcfuclcd Condition Station Blackout Mode MODE EAL#
c All 7.1.1
) (t, 2, 3, 4) 7.1.2.a
( 5, 6, Defueled) 7.1.2.b E IF IF IF M Loss of power to Loss of 13KV Offsite Power Availability to Loss of 4KV Vital Bus Power Sources E All AJLL 4KV Vital Buses (Offsite and Onsite) which results in the R 4KV Vital Buses as evidenced by a loss of function of availability of only one G
BOTH Station Power Transformers 4KV Yitai Bus Power Source AND E
13(23) and 14(24) (Offsite or Onsite)
N
> 15 Minutes have c
y AND I AND elapsed A
c
> 15 minutes have elapsed t 15 Minutesfave elapsed T
I THEN I
0 THEN N
L E
v E
L s .. **
Action Refer to Attachment 1 Refer to Attachment 2 UNUSUAL EVENT Required ALERT
7.0 Electrical Power SGS ECG Rev.00 Page 2 of2 7.1 Loss of AC Power Capabilities Initiating Condition Loss of All Offsite Power and All Onsite AC Power to Vital AC Busa C Prolonged Loss of All Offsite Power
_ _ _ _ _an_d_Pr_o_lo_ng_ed_Loss of All Onsite AC Power to Vital AC Buses (t, 2, 3, 4)
MODE (l, 2, 3, 4) (t,2,3,4) (1, 2, 3, 4)
EAL# 7.1.3 7.1.4.a 7.1.4.b . 7.1.4.c IF E
M Loss of power to E AU 4KV Vital Buses R
AND G
E >IS minutes N have elapsed c
y A
I THE: I AND ..
AND IAND c Restoration of Power T CFST CFST I
to at least one 4KV c ORE COOLING HEAT SINK Vital Bus within REDPATH RED PATH 0
N 2 Hrs.
is NOT likely L
E I -***
v THEN E
L s ', '
Refer to Attachment 3 . Re f er.io Attachment 4 Action SITE AREA GENERAL Required EMERGENCY EMERGENCY
SGS ECO 7.0 Electrical Power Rev. 00 .
Page I ofl 7.2 Loss of DC Power Capabilities Unplanned Loss of Required DC Initiating Power While the Unit is in Either Cold Condition Shutdown or Refueling Mode for > 15 Min. ( _ _ _ _Lo_s_s_or_A1_1_v1-*ta1_(1_E_)_D_c_P_ow_e_r_ _ _)
MODE ( 5,6 ) c 5, 6 ) <~ l, 2, 3, 4 ) c l, 2, 3, 4 )
EAL# 7.2.1.a 7.2.1.b 7.2.3.a 7.2.3.b IF IF IF IF E -
M Unplanned drop in Unplanned drop in Unplanned drop in Unplanned drop in E Voltage to Voltage to Voltage to Voltage to R < 114 VDC <25 VDC < 114 VDC <25 VDC G on on on on E ALL ALL ALL ALL N 125VDC Vntal buses 28VDC Vital buses l 25VDC Vital buses 28 VDC Vital buses c
y AND I AND AND l AND A > IS minutes have > IS minutes have > 15 minutes have > 15 minutes have c elapsed elapsed [ elapsed elapsed T . *-.
I I AND I AND 0
N Loss of control of Loss of control of Safety Related Safety Related L Equipment from the Equipment from the E Control Room has Control Room has v been confirmed been confirmed E
L s
I I lTHEN 1THEN Action Refer to Attachment l Refer to Attachment 3 Required UNUSUAL EVENT SITE AREA EMERGENCY
SGS ECG 8.0 System Malfunction~ Rev.00 Page I of2 8.1 Loss of Heat Removal Capabililty Initiating Condition Inability to Maintain the Plant in Cold Shutdown --J Complete Loss of Functions Needed to Achieve or Maintain the Plant in Hot Shutdown
'Loss of Reactor Vessel Level'~
that Has or Will Uncover Fuel in the Reactor Vessel MODE EAL#
~
8.1.2 4 onRHR Cooling, 5, 6 8.1.3.a 0 8.1.3.b IF IF E
M An Unplanned, Complete loss of I R VLIS Full Range E ALL systems providing Decay Heat Removal f unctior~_s____J <57%
R G
E N IAND
~------
c RCS Temperature has risen to An UNCONTROLLED y
> 200°F temperature rise is RAPIDLY (Excluding a < 15 minutes rise approaching 200°F A
>200°F with a heat removal (with NO heat removal c function restored) function restored)
T I ~-,------ -----------
0 I Actions required by N OP-AB.RHR AND have NOT maintained L RCS temperature E < 350°F v
E L I s -THEN Action Refer to Attachment 2 Refer to Attachment 3 Required ALERT SITE AREA EMERGENCY
8.0 System Malfunctions SGS ECG Rev. 00 8.1 Loss of Heat Removal Capabililty Page2of2 Initiating Complete Loss of Functions Needed to Achieve or Maintain the Plant in Hot Shutdown Condition 1, 2, 3, 1, 2, 3, MODE and 4 with RHR in ln"ection Mode and 4 with RHR in In* ection Mode EAL# 8.1.3.c 8.1.3.d IF IF E
M ALL Turbine Stop Valve (MS 28)
HEAT SINK RED PATH E Closed R
G I AND E LOSS of ALL N Steam Dump Valves c (TB 10, 20, 30, 40) y I AND A LOSS of ALL MS 10 (Steam c Generator Power-Operated Relief T Valves) Valve Control I (BOTH Auto AND Manual) 0 N I AND L >15 minutes have elapsed E
v I E THEN L
s Action Refer to Attachment 3 Required SITE AREA EMERGENCY
SGS ECG Rev.00 8.0 System Malfunctions 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 Annunciation or Indication in the Control Annunciators and a Significant Transient is in Progress or Compensatory Progress Condition Room for Greater Than I~ minutes Indicators are Unavailable MODE (t,2,3,4) (1,2, 3,4) (1,2,3,4) (t, 2, 3, 4) 8.2.1 8.2.2.a 8.2.2.b 8.2.3 EAL#
IF IF E
M Loss of> 75% of Control Room Unplanned loss of E Overhead Annunciators
> 75% of Control Room AND R Overhead Annunciators I G
E A significant transientu is in progress I N AND c lTHEN
- AND AND I y Alternate Indications Alternate Indications are NOT A significant are NOT AVAILABLE per AB.ANN-OOOl(Q) transientu AVAILABLE per AND A
c AND is in progress AB.ANN-000 I (Q) I Control Room Indications are NOT available T
15 minutes to monitor ANY one of the following :
I F
!have elapsed *RCS Status 0
since the loss of
- Reactivity Control N
OHAs JAND
- ECCS L 15 minutes have
- Secondary Systems E THEN elapsed since the loss (SGs/AFW) v ofOHAs
- Containment Parameters E
L s
l THEN l 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 judgement, but includes as a minimum any one of the following:
RX TRIP, LOAD REJECTION >25% POWER, ECCS INJECTION, THERMAL POWER OSCILLATION >10%.
SGS ECG Rev.00 8.0 System Malfunctions Page I oft 8.3 Loss of Communications Capabililty Initiating Unplanned Loss of All Onsite or Otfsite Communications Capabilities Condition MODE c All ) ( All )
8.3.1.b 8.3.1.a EAL#
IF IF E Unplanned loss of Unplanned loss of M 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 systems: systems:
E
- Station Page System
- Direct Inward Dial System N
(Gaitronics) (DID) c y
- Station Radio System
- Nuclear Emergency
- Direct Inward Dial System Telephone System (NETS)
A (DID)
- Essx Phone System (Centrex) c T
I 0
N L
E v
E THEN L
s Ir Action Refer to Attachment I Required UNUSUAL EVENT
SGS ECG 8.0 System Malfunctions Rev.00 Page I ofl 8.4 Control Room Evacuation Initiating Control Room Evacuation has been Control Room Evacuation has been Initiated Condition Initiated and Plant Control Cannot be Established MODE c All ) ( All )
EAL# lia 8.4.3 IF E
M Control Room Evacuation E has been initiated R
G E
N AND c Contr@I of the plant CANNOT be established y
from outside the Control Room within A 15 minutes c
T THEN I
0 N
L E
v E
L s
Action Refer to Attachment 2 Refer to Attachment 3 .
Required ALERT SITE AREA EMERGENCY
SGS ECG 8.0 System Malfunctions Rev.00 Page 1 ofl 8.5 Technical Specifications Initiating Inability to Reach Required Mode Within Condition Technical Specification Limits MODE c 1, 2, 3, 4 )
EAL# 8.5.1 IF E
M Plant is NOT brought to the required Mode within E the Technical Specification required time limit R
G E THEN N
c y
A c
T I
0 N
L E
v E
L s
Action Refer to Attachment I Required UNUSUAL EVENT
SGS ECG 9.0 Hazards - Internal/External Rev.00 Pagel ofl 9.1 Security Threats
/
Confinned Security Event ecuHty Event Resulting in Loss o Initiating Security Event in a Plant Which Indicates a Potential Degradation in ( Security Event in a Plant Vital Area Ability to Reach and Maintain Cold Condition the Level of Safety of the Plant Protected Area Shutdown
'-~*~~~~~~~~~--
MODE ( All ) ( All ) ( All ) ( All )
EAL# 9.1.1 9.1.2 9.1.3 9.1.4 IF IF IF IF E Confinned security threat directed toward Confirmed hostile Confirmed hostile intrusion or Security event resulting in the M the station as evidenced by ANY one of the intrusion or malicious acts malicious acts in Plant Vital Areas as actual loss of physical control E following: as evidenced by ANY one evidenced by : of EITIIER one of the following:
R
- Credible threat of malicious acts or of the following:
- Discovery of an intruder(s), armed
- Control Room G destructive device within the Protected
- Discovery of an and violent, within the Vital Area
- Remote Shutdown Panel 213 E Area resulting in SCP-5 implementation intruder(s), armed and resulting in SCP-6 implementation N
- Credible intrusion or assault threat to the violent, within the
- Malicious acts or destructive device Protected Area resulting THEN c Protected Area resulting in SCP-5 implementation in SCP-6 discovered in a Vital Area resulting y in SCP-10 implementation implementation
- Attempted intrusion or assault to the Protected Area resulting in SCP-7 or
- Hostage held on-site -in a THEN A non-vital area resulting SCP-I I implementation c
- Malicious acts attempted or discovered in SCP-8 T within the Protected Area resulting in implementation I
0 SCP-10 implementation
- Hostage/Extortion situation that threatens THEN c5 N nonnal plant operations resulting in SCP-8 implementation L
- Destructive Device dliscovered within the E Protected Area resulting in SCP-10 v implementation E
L THEN s
- Ir .. **
Action Refer to Attachment 24 Refer to Attachment 2 -Refer to Attachment 3 Refer to Attachment 4 Required UNUSUAL EVENT (Common Site) ALERT ITE AREA EMERGENCY GENERAL EMERGENCY
SGS ECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.2 Fire Initiating Condition MODE (
Fire Within the Protected Area Boundary Not Extinguished Within 15 Minutes of Detection All ) ( All )
G Fire Affecting the Operability of Plant afety Systems Required to Establish or Maintain Safe Shutdown
( All )
EAL# 9.2.1 9.2.1 9.2.2 IF IF IF E Valid Fire Alarm is received I Report of a fire from Fire within ANY one of the following Plant Vital Structures:
M in the Control Room I personnel at the scene
- Auxiliary Building E IAND
- Service Water Intake Structure R Fire within ANY one of the following Plant Structures
- Control Point Area G (EXCLUDING small fires that have NO potential to affect
- Inner/Outer Penetration Areas E Safety System!! or Protected Area Pennanent Plant Structures)
N
- Auxiliary Building
- Containment c
- Service Water Intake Structure
- Fuel Handling Building y
- Service Building
- Control Point Area
- Inner/Outer Penetration Areas
- RWST, PWST, and AFWST Area A
c
- Containment !AND T
- Fuel Handling Building The Fire is of a magnitude that it SPECIFICALLY I
- Service Building results in Damage to ANY one of the following:
0
- RWST, PWST, and AFWST Area
- TWO OR MORE Trains of a Safety System N
- MORE THAN ONE Safety System
- Turbine Building
- Any Plant Vital Structure which renders the structure L jAND incapable of performing its Design Function E The Fire is NOT extinguished within 15 minutes of --
v EITHER one of the following: jAND E
L s
- Receipt of a Valid Fire Alarm
- Report of a fire from the scene [__ Damaged Safety System(s) or Plant Vital Structure
' is required for the present MODE of operation Action lTHEN Refer to Attachment 1 l THEN Refer to Attachment 2 Required UNUSUAL EVENT ALERT
SGS ECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.3 Explosion
~-------.,.E-x-pl-o-,si_o_n-.,.Afti--=-ec-ti-=--.n-g-th-=-e-Ope=--r-:ab;-;i-:-;li_.,-ty-o-;;f-;:;P:;--la-n-:-t- - - - - . . .
Natural and Destructive Initiating Phenomena Affecting the Safety Systems Required to Establish or Condition Protected Area Maintain Safe Shutdown
( All ) ( All )
MODE 9.3.1 9.3.2 EAL#
IF IF E Confirmed Explosion Confirmed Explosion within ANY one of the following M within Plant Vital Structures:
E the Protected Area
- Auxiliary Building R AND *Service Water Intake Structure I
G
- Control Point Area E
Report of visible damage to Plant
- Inner/Outer Penetration Areas N
equipment or to Protected Area c Permanent Plant Structures
- Containment y
- Fuel Handling Building
- Service Building A THEN
- RWST, PWST, and AFWST Area c
T IAND I The Explosion is of a magnitude that it SPECIFICALLY 0 results in Damage to ANY one of the follo~ng:
N
- TWO OR MORE Trains of a Safety System
- MORE THAN ONE Safety System L
- Any Plant Vital Structure which renders the structure incapable of E performing its Design Function v --
E L
l-.- - IAND Damaged Safety System(s) or Plant Vital Structure is required for the present MODE of operation s ~THEN 1
Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
SGS ECG 9.0 Hazards - Internal/External R.:v.00 Page I of 2 9.4 Toxic/Flammable Gases Initiating Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant Condition MODE ( All ) <~ __D ( All )
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 with in the Protected Area release within the R to EVACUATE inAN Y area which Protected Area G non-essential personnel does not n ormally require an that RES ULTS in E due to an at mo spheric survey Flammable Gas concentrations N Offsite Toxic Gas release or Respira tory Protection for EXCEEDING c entry 25% of the LEL y
AND A
c I I SNSS deems evacuation AND T of non-essential personnel I is required 0 Routine Plant Operations are IMPEDED based N on EITHER one of the following:
THEN
- Ace ess restrictions caused by the uncontrolled release L
- Personnet injuries have occurred as a result of the release E
v THEN E '
L s '
Refer to Attachment 24 Refer to Attachment 1 Action UNUSUAL EVENT (Common Site) UNUSUAL EVENT Required
9.0 Hazards - Internal/External SGS ECG 9.4 Toxic/Flammable Gases Rev.00 Page 2 of 2 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 MODE ( All 9.4.2.a
) c All )
EAL# 9.4.2.b IF IF E Uncontrolled Toxic Gas release within ANY one Uncontrolled Flammable Gas release within ANY one M of the following Plant Vital Structures of the following Plant Vital Structures:
E
- Auxilary Building
- Auxilary Building R
- Service Water Intake Structure
- Service Water Intake Structure G
- Control Point Area
- Control Point Area E
- Inner/Outer Penetration Area
- Inner/Outer Penetration Area N
- Containment
- Containment c
- Fuel Handling Building
- Fuel Handling Building y
- Service Building
- Service Building
- RWST, PWST, and AFWST Area
- RWST, PWST, and AFWST Area A
c I AND AND T Toxic Gas co~centrations result in ANY one Flammable Gas concentrations EXCEED I of the following: 50% of 1the LEL 0
- An IDLH atmosphere N
- Plant personnel report severe adverse health reactions, including burning eyes, nose, throat, dizziness L
- The Threshold Limit Value (TLV) being EXCEEDED E
I v
E L Plant personnel are unable to perform actions necessary to complete a Safe s Shutdown of the plant without appropriate personnel protection equipment
_J THEN Action ~,;~~hment 2 I Required L_ _ _ ~~ERT -
SGS ECG 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 Affecting the Protected Area Affecting the Plant Vital Area MODE ( All ) ( All ) ( All )
EAL# 9.5.1.a 9.5.1.b 9.5.2 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 Seismic Monitor System cabinet in c the #I CR Equipment Room y
A c
T I
0 AND Valid Actuation of the Hope Creek Seismic N Switch(> O.lg) has occurred as verified by The Hope Creek SNSS L
E v
E L THEN s
Action Refer to Attachment 24 Refer to Attachment 2 Required UNUSUAL EVENT (Common Site) ALERT
SGS ECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.6 High Winds Initiating Natural and Destructive Pheno~ena Natural and Destructive Phenomena Affecting the Plant Vital Area Condition Affecting the Protected Area MODE ( All ) ( All ) ( All )
EAL# 9.6.1.a 9.6.1.b 9.6.2 IF IF E ~--*-----*---------------------------,
M Report of a Sustained wind speeds The Wind Speed is of a magnitude that it SPECIFICALLY E Tornado > 75 MPH for 15 results in Damage to ANY of the following:
R TOUCHING DOWN minutes, from G within the ANY elevation of
- MORE THAN ONE Safety System N
c y
I I
- Rendering ANY of the following structures incapable of performing its Design Function:
AND l
- Auxiliary Building A
- Service Water Intake Structure c THE:
- Control Point Area T
- Inner/Outer Penetration Areas I
- Containment 0
- Fuel Handling Building N
- Service Building
- R WST, PWST,and AFWST Area L
E I AND v Damaged Safety System(s) or Plant Vital Structure E
is required for the present MODE of operation L
s lTHEN Action Refer to Attachment 24 Refer to Attachment 2 Required UNUSUAL EVENT (Common Site) ALERT
SGS ECG 9.0 Hazards - Internal/External Rev. 00 Page I ofl
- 9. 7 Flooding Internal Flooding in Excess of Internal Flooding Affecting the Initiating Condition Sump Handling Capability Affecting Safety Related Areas of the Plant
( Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown MODE ( All ) ( All )
EAL## 9.7.1 9.7.2 IF E
IF M Visual Observation of Flooding within ANY one Severe Flooding of Safety System Areas E of the following Plant Vital Structures:
HAS ENDANGERED R
- Auxiliary Building safety related equipment per G
- Service Water Intake Structure OP-AB.ZZ-0002 E
- Fuel Handling Building N
c THEN
- Service Building y
- Containment I AND A
The Flooding is of a magnitude that it SPECIFICALLY c results in Damage to ANY one of the following:
T I *TWO OR MORE Trains ofa Safety Syetem 0
- MORE THAN ONE Safety System N
- Any of the above listed Plant Vital Structures which L renders the structure incapable of performing its E Design Function v ~
I AND E
L [-- Damaged Safety System(s) or Plant Vital Structure s 1s reqmred for the present MODE of operation
,, l THEN Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
SGS 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 Affecting Certain Structures Within the Protected Area Condition MODE
( All ) c All ) c All )
9.8.1.a 9.8.1.b 9.8.2 EAL# IF IF E Catastrophic damage to the Vehicle Crash I Missile Impact with M Main Turbine as evidenced by or within ANY one of the following E EITHER one of the following: Plant Vital Structures:
R
- Main Turbine casing penetration
- Auxiliary Building G
- Main Turbine/Generator Damage *Service Water Intake Structure E potentially releasing Lube Oil or
- Inner/Outer Penetration Areas N Hydrogen Gas to the Turbine
- Containment c Building y
- Fuel Handling Building
- Service Building A
- RWST, PWST, and AFWST Area The Vehicle Crash I Missile Impact is of a c magnitude that it SPECIFICALLY results T in Damage to ANY one of the following:
I AND
- TWO OR MORE Trains of a Safety System
l 0
- MORE THAN ONE Safety System N
- Any of the above Plant Vital Structures which
- L renders the structure incapable of performing its Design Function E
v AND E THEN L
Damaged Safety System(s) or Plant Vital Structure s is required for the present MODE of operation
,, THEN Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
SGS ECG 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 MODE ( All ) ( All )
EAL## 9.9.1.b 9.9.1.a IF IF E
M River Level > 99.5' 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 (Com~
Required
SGS ECG 11.0 Reportable Action Levels Rcv.00 Page I of 3 11.1 Technical Specifications '
INITIATION OF ANY UNIT SHUTDOWN EXCEEDING ANY TECHNICAL SPECIFICATION ANY DEVIATION FROM TIS OR Initiating REQUIRED BY THE TECHNICAL SPECIFICATIONS SAFETY LIMIT LICENSE CONDITION PURSUANT TO Condition (IOCFR50.36(cXI)) IOCFR50.54(x) (IOCFR50.72(bXIXiXB))
(IOCFRS0.7l(bXIXiXA)J MODE ( 1, 2 ) ( 1, 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 p Unit shutdown is Exceeding EITHER one Deviation from written INITIATED of the following procedures because no action 0
R to comply with Technical Specification Safety Limits: consistent with Technical Technical Specifications Specifications or license T
- T/S 2.1.1, Thermal Power, condition can provide adequate I Pressurizer Pressure, Coolant N or equivalent protection in an Temperature combination G emergency
- T/.S 2.1.2, RCS Pressure (see NC.NA-AP.ZZ-OOOS(Q) for A guidance on deviation from c procedures)
T I
0 N
L E
v E
L THEN s
- ir Action JRefer to Attachment 12 Required L I Hour Report
SGS ECG Rev.00 11.0 Reportable Action Levels Pagel of3 11.1 Technical Specifications STEAM GENERATOR TIJBE INSPECTIONS WHICH FALL INTO ABNORMAL DEGRADATION OF Tiffi CONTAINMENT Initiating CATEGORY C-3 TIIAT HAVE BEEN EVALUATED FOR STRUCTURE DETECTED DURING SHUTDOWN TIIAT HAS BEEN Condition REPORTABILITY [IOCFR50.72(bX2Xi);T/S 4.4.5.2(6.2)) EVALUATED FOR REPORTABILITY [IOCFR50.72(bX2Xi); T/S 4.6.1.6.2)
MODE ( 5, 6, Defueled ) ( 3, 4, 5, 6, Defueled)
RAL## 11.1.2.a 11.1.2.b R IF IF E
p Results of SG tube inspections which fall into category Any abnormal degradation of the Containment structure C-3 of T/S 4.4.5.2 (Unit 1) or T/S 4.4.6.2 (Unit 2) detected by visual inspection of exposed accessible interior and 0
R exterior surfaces during shutdown T
I N
G AND A
c An engineering evaluation has determined that it is reportable T
I pursuant to 10CFRSO. 72(b)(2)(i) 0 THEN N
L E
v E
L s
Action Refer to Attachment J4I Required 4 Hour Report _ __J
SGS ECG 11.0 Reportable Action Levels Rev.00 Pagc3 of 3
- 11. l Technical Specifications VIOLATION OF 1llE REQUIREMENTS NY EVENT REQUIRING AN ENGINEERING EVALUATION BY TECH SPECS OR COMMITME Initiating [Ul T/S 3.4.9.1, 3.4.9.2, 3.4.7, 3.7.9, JAN 1983, LTR TO NRC, 3.7.2.1)
CONTAINED IN 1llE OPERATING LICENSE Condition [Salem U2 Operating Lioense, Sections 2.1) [U2 T/S 3.4.10.1, 3.4.10.2, 3.4.8, 3.7.9, JAN 1983, LTR TO NRC, 3.7.2]
MODE (_AI_I~) (_AI_I~)
RAL# 11.1.3.a 11.1.3.b IF IF R
E Violation of ANY one of the p As judged by the SNSS/EDO, ANY one of the following conditions have been satisfied:
requirements contained in Section 0 2.C
- Any of the T/S LCOs for RCS or PZR heatup or cooldown rates are exceeded R (Items 3 through 25)
- The concentration of either chloride or fluoride in the RCS is in excess of its T or Section 2E, 2F or 2G I Steady State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or in excess of its Transient Limit, thereby of the Salem Unit 2 Operating requiring an engineering evaluation to determine the effects of the out of limit condition N License G on the structural integrity of the RCS THEN
- One or more snubbers are found to be INOPERABLE and require an engineering A evaluation performed in accordance with T.S.4.7.9 action statement c
T
- Any PZR code safety valve discharges I
- The temperature of EITHER the Primary or Secondary Coolant in any S/G is ~ 70° F 0 WHEN the pressure of either the Primary or Secondary Coolant in the S/G is > 200 psig N
THEN L
E v
E L
s Refer to Attachment 20 Refer to Attachment 22 Action Required 24 Hour Report [- _ _o_T_HE __ R_R_e_p_o_rt_ __ _ J
11.0 Reportable Action Levels SGS ECG Rev.00 11.2 Design Basis I Unanalyzed Condition Page I of2 ANY EVENT OR CONDITION DURING OPERATION PIPE CRACKS IN STAGNANT BORATED WATER SYSTEMS Initiating THAT RESULTS IN THE CONDITION OF THE PLANT BEING (IE Bulletin 79-17)
SERIOUSLY DEGRADED (I OCFR~O. 7l(bXI Xii))
Condition
( 1,2 ) ( All )
MODE RAL# 11.2.1.a 11.2.1.b IF IF R
E As judged by the SNSS/EDO, Cracks in weld areas of Borated Safety p an event or condition found Related piping 0 during plant operations that results * (as reported by Engineering or ISi)
R in ANY one of the following:
T I
- The condition of the plant, including its N principal safety barriers, being seriously G degraded
- The plant being in an unanalyzed condition A that significantly compromises plant safety c
T
- The plant being in a condition outside the I design basis of the plant 0
- The plant being in a condition not covered N by normal/abnormal or emergency operating procedures L
E v
E L THEN s ,,
Action Required
SGS ECG 11.0 Reportable Action Levels Rev.00 Page 2of2 11.2 Design Basis I Unanalyzed Condition ANY EVENT FOUND WHILE SHlJfDOWN THAT WOULD EVENT/CONDITION THAT ALONE COULD PRESENCE OF A LOOSE PART IN Initiating HAVE SERIOUSLY DEGRADED THE PLANT OR RESULTED IN HAVE PREVENTED CERTAIN SAFETY FUNCTIONS THE REACTOR COOLANT SYSTEM Condition EING IN AN UNANALYZED CONDITION (IOCFRS0.72(b)(2)(i) ( IOCFRS0.72(b)(2)(iii)J (Reg. Guide 1.133]
MODE (3,4,5,6,defueled) ( All ) ( All )
RAL# 11.2.2.a 11.2.2.b 11.2.2.c IF IF IF R
E Any evelllt, found while the Reactor is Any event or condition that Presence of a Loose Part in p shutdown, that, had it been found during alone could have prevented the the RCS is confirmed 0 operation, would have resulted in the plant, fulfillment of the safety function of R including its principal safety barriers being in structures or systems that are needed to T EITHER one of the following conditions: perform ANY one of the following:
I
- Seriously degraded
- Control the release of radioactive N
G
- In an unanalyzed condition that significantly material compromises plant safety
- Shutdown the reactor and maintain it A in a safe shutdown condition c
T
- Remove residual heat I
- Mitigate the consequences of an 0 accident N
L E
v E THEN L
s Action Refer to Attachmen~
Required 4 Hour Report __ _J
SGS ECG 11.0 Reportable Action Levels Rcv.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 RES ULTED IN ECCS DISCHARGE INTO THE RCS AS THE (INCLUDING THE REACTOR PROTECTION SYSTEM)
Condition RESULT OF A VALID SIGNAL [IOCFR50.72(b)(l)(iv)] EXCEPT PREPL~D [IOCFR50.72(b)(2)(ii)]
MODE (~_Al_t~) (..___Alt~)
RAL# 11.3.1 11.3.2 IF IF R
E Valid SI Actuation signal received (or demanded) p Any event or condition that results in manual or automatic AND actuation of any Engineered Safety Feature (ESF), except as 0
part of a preplanned sequence during reactor operation or R
T testing, including the Reactor Protection System (RPS)
ANY ECCS Pump start or I Accumulator depressurization that results in or N AND should. have resulted in, discharge to the RCS G
THEN ESF I RPS Actuation is determined A
to be reportable in accordance with c NC.NA-AP.ZZ-OOOO(Q), Action Request Process.
T I THEN 0
N L
E v
E L
s Action
- Ir Refer to Attachment 12 Refer to Attachment 14 Required 1 Hour Report 4 Hour Report
SGS ECG 11.0 Reportable Action Levels Rev.00 Page I of2 11.4 Personnel Safety I Overexposure ANY INCIDENT OR EVENT INVOLVINO BYPRODUCT, ANY INCIDENT OR EVENT INVOLVINO LOSS OF . ONSITE FATALITY Initiating SOURCE, OR SPECIAL NUCLEAR MATERIAL CAUSING CONTROL OF LICENSED MATERIAL CAUSING ANY OF
[10CFR50.72(b)(2)(vi)]
Condition ANY OF THE LISTED RESULTS(IOCFRl0.2202(a)) THE LISTED RESULTS (IOCFR20.2202(b))
MODE (____ All_) _( __AI_t_) (____AI_t_)
RAL# 11.4.1 11.4.2.a 11.4.2.b IF IF IF R Any fatality has E PERSONNEL OVEREXPOSURE or PERSONNEL OVEREXPOSURE or
- potential for overexposure as indicated by occurred within the p potential for overexposure as indicated by ANY one of the following: Owner Controlled 0 ANY one of the following:
Area(OCA)
R
- LOE exposure > 15 Rem N
- Release of radioactive material inside or
- Release of radioactive material inside or A outside of a Restricted Area so that, had outside of a Restricted Area so that, had c an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an individual been present for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, T the individual could have received 2:. 5 the individual could have received > 1 I times the occupational ALI (Annual times the occupational ALI (Annual 0 Limit of Intake) which would usually Limit of Intake) which would usually N equate to 25 Rem CEDE. This DOES equate to 5 Rem CEDE. This DOES NOT apply to areas where personnel are NOT apply to areas where personnel L NOT normally stationed during routine are NOT normally stationed during E operations routine operations v
E THEN THEN L
s
... '" *Ir Action Refer to Attachment 12 !Refer to Attachment 14 Refer to Attachment 17 Required 1 Hour Report l____ 4 Hour Report 4 Hour Report
11.0 Reportable Action Levels SGS ECG Rev.00 Pagel of2 11.4 Personnel Safety I Overexposure RADIOACTIVELY CONTAMINATED PERSON FITNESS FOR DUTY PROGRAM:
Initiating SIGNIFICANT FITNESS FOR DUTY FALSE POSITIVE DUE TO ADMINISTRATIVE ERROR TRANSPORTED FROM THE SITE TO AN OFFSITE MEDICAL Condition FACILITY FOR TREATMENT [IOCFRS0.72(bX2Xv)) EVENTS (IOCFR26.73) (BLIND TEST BY LAB) (IOCFR26, APP .A, 2.8(eXS)I MODE (~_AI_I~) c All ) c All )
RAL# 11.4.2.c 11.4.3.a 11.4.3.b IF IF IF R
E Transportation of a radioactively Any event that is determined to be The occurrence of a false positive error on p contaminated or potentially contaminated reportable by the Medical Review a blind lab performance test specimen 0 individual from the site to an offsite medical Officer (MRO) or designee IAW under 10CFR26 as determined by the R facility for 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
AND The reportable details of the event are A made available to the SNSS by the The reportable details of the event are made c MRO or designee. available to the SNSS by the MRO or T
designee.
I THEN 0 THEN N
L E
v E
L s
Action Required Refer to Attachment 17 Refer to Attachment 19 Refer to Attachment 19 4 Hour Report 24 Hour Report 24 Hour Report
11.0 Reportable Action Levels SGS ECG Rev.00 11.5 Environmental Pagel of I SPILUDISCHARGE OF ANY NON-RADIOACTIVE UNUSUAL OR IMPORTANT SPll..llDISCHARGE OF ANY NON-Initiating RADIOACTIVE HAZAROOUS SUBSTANCE HAZAROOUS SUBSTANCE INTO OR UPON TIIE ENVIRONMENTAL EVENTS Condition RIVER (10CFR50.72(b)(2) (vi); N.J.A.C.7:1E) [E.P.P. SECTION 4.1)
(IOCFR50.72(b)(2)(vi); N.J.A.C. 7:1E]
MODE ( All ) ( All
) c All )
RAL# 111.5.2.a 11.5.2.b 11.5.2.c IF IF IF R
EITHER one of the following events occur: As judged by the SNSS/EDO, ANY one of E Spill/discharge olf an industrial chemical or the following events has occurred:
p petroleum product outside of a Plant
- Observation of a spill/discharge of an industrial 0 Structure within the Owner Controlled
- Unusually large fish kill chemical or petroleum product from on-site into Area that results in EITHER one of the R following:
the Delaware River or into a storm drain
- Protected aquatic species impinge on T
- Observation of an oil slick on the Delaware River Circulating or Service Water intake I
- Spill / discharge that has passed through from any source screens (eg.; sea turtle, sturgeon) as N the engineered fill and into the ground reported by Site personnel water as confirmed by licensing G THEN
- Any occurrence of an unusual or
- Spill / discharge that CANNOT be important event that indicates or could A cleaned up within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and no contact 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 outbreaks 0
- Mortality or unusual occurrence of N any species protected by the Note:
This event May require IMMEDIATE Endangered Species Act of 1973 L .. ~ (15 minute) notifications. DO NOT
- Increase in nuisance organisms or E delay implementation of Attachment conditions .
v 16.
- Excessive bird impactation E
- NJPDES Permit violations
- Excessive Opacity (smoke)
L 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*
SGS ECG 11.0 Reportable Action Levels Rev.00 Page 1 ofl 11.6 After-the-Fact Initiating EMERGENCY CONDITIONS DISCOVERED Condition AFTER-THE-FACT MODE ( All )
RAL# 11.6.1 IF R
E Discovery of events or conditions that had p
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 I. 21 Required I Hour Report __J
SGS ECG 11.0 Reportable Action Levels Rev.00 Pagel of I
- 11. 7 Security I Emergency Response Capabilities
~-----*- - -----*
SAFEGUARDS EVENTS TIIAT ARE DETERMINED .) MAJOR LOSS OF EMERGENCY ASSESSMENT CAPABILITY, OFFSITE RESPONSE Initiating TO BE NON-EMERGENCIES, BlIT ARE REPORTABLE
. TO Tiffi NRC WffHIN ONE HOUR [IOCFR73.7l(bXI)).
( CAPABILITY, OR COMMUNICATIONS CAPABILITY (IOCFR50.72(b)(l)(v)J Condition MODE c All 11.7.1.a
_) All 11.7.1.b
) ( All 11.7.1.c
)
RAL# IF IF IF SNSS/EC determines that an event (excluding a scheduled test or preplanned R Any Non-Emergency safeguards event that is E reportable in accordance with 10CFR73. 71 maintenance activity) has occurred that would impair the ability to deal with an accident p as determined by Security (SCP-IS)
- or emergency as indicated by the Loss of ANY one of the following:
0 R
THEN _________I.______
T
- Nuclear Emergency I
- P250 or Aux Annunciator System Telecomunications System (NETS)
N for> 24 hrs for>l hr G
- SPDS for > 8 hrs ( > 2 CFST~
- ENS for> l hr in the Control lnop, not due to plant conditions)
>I hr (NIA if reported by the NRC) .
T
- More than seven Offsite Sirens
- Use of the TSC for> 8 hrs I
for> l hr
- ALL Plant vent radiation effluent 0
- Use of the EOF for> 8 hrs monitors for> 8 hrs N
- More than 75% OHA's for< 15 min
- All Meteorological data (Salem L
AND Hope Creek) for> 8 hrs Concurrent multiple accident or E
emergency condition indicators which in v
- Site access due to Acts of Nature the judgement of the SNSS significantly E (snow, flood, etc.)
impairs assessment capabilities L
s l
. THEN THEN Action Refer to Attachment 11 Refer to Attachment 25 Refer to Attachment 12 Required l Hour Report (Common Site) l Hour Report (Common Site) l Hour Report
SGS ECG 11.0 Reportable Action Levels Rev. 00 Page I of I 11.8 Public Interest Initiating UNUSUAL CONDITIONS DIRECTLY AFFECTING LOWER Condition ALLOWAYS CREEK TOWNSIDP (LACT) [LAC -MOU]
MODE ( All ) ( All )
RAL# 11.8.2.a 11.8.2.b IF IF R
E p SNSS/EDO judges that an event or situation has As judged by the SNSS/EDO, events which are the 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 I
- The health and safety of onsite personnel 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 I 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
E I
THEN v
E L
s Action Refer to Attachment 14 Refer to Attachment 21 Required 4 Hour Report LACT I MOU Report
SGS.ECG 11.0 Reportable .Action Levels Rcv.00 Page 1 of2 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases UNPLANNED I ACCIDENTAL LOSS AND INVESTIGATION OF THE LOSS OF SPECIAL THEFT OR LOSS OF Initiating CRITICALITY NUCLEAR MATERIALS/ SPENT FUEL LICENSED MATERIAL Condition (IOCFR70.52(a)] [IOCFR73.27(c), IOCFR73.7l(a)) (IOCFR20.2201(aX1Xi)]
MODE ( All ) ( All ) ( All )
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 ~ 1000 0 times the quantity specified THEN R
- Shipment of formula quantities of Strategic in 10CFR20 Appendix C in T Special Nuclear Material (SSNM) or Spent such circumstances that an I Fuel that is lost or unaccounted for after the exposure could result to N estimated time of arrival persons in Unrestricted G Areas.
- A lost or unaccounted for shipment of SSNM or A Spent Fuel has been recovered or accounted for THEN c
- Results of a trace investigation of lost or T unaccounted for SSNM shipment are received I
0 THEN N
L E
v E
L s
Action Refer to Attachment 12 Refer to Attachment 11 Refer to Attachment 11 Required 1 Hour Report 1 Hour Report (Common Site)
--~--**---~
1 Hour Report (Common Site)
SGS ECG 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 ONTAMINATION OlITSIDE OF Initiating RECEIPT OF SSNM MATERIAL AND/OR RADIATION LEVELS ON A OF LICENSED MATERIAL RADIOLOGICALLY CONTROLLED
[IOCFR73.27(b)] ( AREA [IOCFR50.72(bX2Xvi)]
Condition PACKAGE (IOCFR20.1906(d)) . [IOCFR71.S(aXIXv))
MODE C..____A1_1_) ( AJI ~ (~_AJ_l~) ( _____AJ_I_)
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 survey Indicates that Accidents during the Di~covery of a Contaminated p Strategic Special Nuclear package contamination/radiation Area OUTSIDE of the RCA 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
~
- 2200 dpm/100 cm 2 as the shipper that involve AND I AND (or potentially involve)
N
- 200 mR/hr on contact damage to the cargo 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 I AND have left the Protected L Size of Area E Contaminated v Area is E LARGE L (>100 FT2) s I
' ' I THEN Action Refer to Attachment I 0 Refer to Attachment I 0 Refer to Attachment 18 Refer to Attachment 13 Required I Hour Report I Hour Report 4 Hour Report 4 Hour Report
- -* - * - * - - - - - - - _ J
SGS ECG 11.0 Reportable Action Levels Rev.00 Page I of!
11.10 Voluntary Notifications initiating EVENTS/CONDITIONS WARRANT VOLUNTARY/COURTESY Condition NRC NOTIFICATION [IOCFR50.72 -VOLUNTARY REPORT]
MODE ( 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 NO other EALs or RALs appear to be applicable G
THEN A
c T
I 0
N L
E v
E L
s Action Required
~------------
Refer to Attachme~;*14*
4 Hour Report J
ECG ATT l Pg. l of 9 ATTACHMENT l UNUSUAL EVENT I. EMERGENCY COORDINATOR (EC) LOG SHEET A. DECLARE AN UNUSUAL EVENT AT SALEM UNIT EC EAL# Declared at hrs on time date B. NOTIFICATIONS
( ) 1. CALL communicators to the Control Room.
( ) 2. COMPLETE the INITIAL CONT ACT tviESSAGE FORM (ICtviF) (last page of this attachment).
( ) 3. PROVIDE the IC~iF to the Communicator (CMl) and DIRECT the CMl to implement Attachment 6.
( ) 4. DIRECT the Secondal)' 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."
"Salem Unit is in an UNUSUAL EVENT condition due to (Repeat)
C. SECURITY RELATED EVENT I. 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 202S, OSC Activation and Operations AND .
( ) IMPLEMENT Bomb Search Operations IAW Appendix 1.
( ) b. DIRECT the NCOs to check control boards for correct equipment lineups.
SGS Rev.00
ECG ATf l Pg. 2 of 9 Initials D. EMERGENCY COO RD INA TOR DUTIES
( ) 1. NOTIFY the Hope Creek SNSS, (NETS 5224; DID 3027, 3059) with Event Description.
( ) 2. IF required, IMPLEMENT Accountability by referring to the Accountability Instructions in Section II.
- 3. COMPLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmittal by EC the CMI 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
- 2. ENSURE appropriate reports are made IAW Section IV., Reporting, of this SNSS attachment.
SGS Rev. 00
ECG ATf l Pg. 3 of 9 II. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA A. Th1PLEMENTATION OF ASSEMBLY AND ACCOUNTABILITY Initials/Time I 1. IF NOT already done, EC THEN DIRECT the OSC Coordinator to activate the OSC IAW EPIP 202S, 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 Hope Creek SNSS to implement EPIP lOlH, Appendix 6, EC Accountability Instructions For An Unusual Event at Salem.
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."
"Salem Unit 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=O Min.)
"Attention, Attention. All accountability stations, Th1PLEMENT Accountability." (Repeat)
I 7. WHEN I 0 minutes have elapsed from Step 6, ANNOUNCE the following; (T+lOMin.)
"Attention, Attention. All accountability stations, COMPLETE YOUR INITIAL Accountability." (Repeat)
SGS Rev. 00
ECG ATil 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.
Initials 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 202S.
( ) 2. UPDATE Security as missing personnel are accounted for.
SGS Rev. 00
ECG ATil Pg. 5 of 9 ID. 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. 6) AND DETERMINE if the UE can be terminated by entering Recovery.
- 2. WHEN the above Step is completed, EC THEN CO.l\1PLETE the "UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM," (Pg. 7), as follows:
( ) a. IF terminating WITHOUT Recovery, CO.l\1PLETE Part A.
( ) b. IF terminating WITH Recovery, CO.l\1PLETE Part B.
- 3. IF termination 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 (termination 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 CMI.
( ) 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 Hope Creek SNSS.
EC
- 7. GO TO Section IV., Reporting.
SNSS SGS Rev.00
ECG ATil Pg. 6 of 9 ill. 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. l & D.2 are checked as YES, THEN SIGN below and GO TO Sect. III., Step 2 for Terminating WITH Recovery.
Emergency Coordinator Date Time SGS Rev. 00
ECG ATil Pg. 7 of 9 III. TERMINATION (cont'd)
UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM PART "A" - EMERGENCY TERMINATION WITHOUT RECOVERY:
THIS IS - - - - - - - - CO:MMUNICATOR IN THE CONTROL ROOM AT THE SALEM NUCLEAR GENERATING STATION, UNIT _ _
THIS MESSAGE IS TO NOTIFY YOU THAT AS OF ON _ _ _ __,
time due THE UNUSUAL EVENT HAS BEEN TERMINATED.
(EC Approval to transmit)
PART "B" - EMERGENCY TERMINATION WITH RECOVERY:
THIS IS _ _ _ _ _ _ __, COMMUNICATOR IN THE CONTROL ROOM AT THE SALEM NUCLEAR GENERATING STATION, UNIT _ _
THIS MESSAGE IS TO NOTIFY YOU THAT AS OF ____ _, ON ____ ___,
time date THE UNUSUAL EVENT HAS BEEN TERMINATED AND SALEM IS NOW IN A RECOVERY STATUS. IS THE RECOVERY MANAGER.
(DUTYEDO)
(EC Approval to transmit)
SGS Rev. 00
ECG ATil Pg. 8 of 9 IV. REPORTING INSTRUCTIONS
- 1. This is a permanent document.
- 2. ATIACH appropriate documents to this form and EXPEDITE the package through all steps.
Initials
- 1. PREP ARE an Action Request (AR).
SNSS
- 2. FORWARD this attachment and supporting documentation to the Openrtions 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 N u m b e r - - - - - - -
- 6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.
SGS Rev.00
ECG ATil Pg. 9 of 9 INITIAL CONTACT MESSAGE FORM I. THIS I S - - - - - - - - - - , COMMUNICATOR IN THE CONTROL ROOM (NAME)
AT THE SALEM NUCLEAR GENERATING STATION, UNIT NO. - - -
II.
0 THIS IS NOTIFICATION OF AN UNUSUAL EVENT WHICH WAS DECLARED AT ON (Time - 24 HR CLOCK)
- - - - -(DATE} ----
EAL# - - - - - - DESCRIPTION OF E V E N T : - - - - - - - - - -
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 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 of2.42E+-05 µCi/sec Noble Gas or 2. lE+-01 µCi/sec I-131.
SGS Rev. 00
ECG ATI2 Pg. 1 of 4 ATTACHMENT 2 ALERT I. EMERGENCY COO RD INA TOR (EC) LOG SHEET Initials ~**. .
A. DECLARE AN ALERT AT SALEM UNIT EC EAL# Declared at -------- hrs on ---------
time date B. NOTIFICATIONS
( ) 1. CALL communicators to the Control Room.
( ) 2. COI\1PLETE the INITIAL CONT ACT MESSAGE FORM (ICMF) (last page of this attachment).
( ) 3. PROVIDE the ICNff to the Communicator (CMl) and DIRECT the CMI to implement Attachment 6.
( ) 4. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for an ALERT.
- 5. NOTIFY the I.T.0.C. Operator on NETS x5027 (20i-430-7191 SNSS or 201-430-8153) with the following message:
"This is (your name) , Senior Nuclear Shift Supervisor at Salem. Please II\1PLEMENT EPIP 204S, Salem 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 Hope Creek SNSS. (NETS 5224; DID 3027, 3059)
- a. PROVIDE a briefing on the ALERT conditions.
- b. DIRECT implementation ofEPIP IOIH, Section 3.1.
- 7. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.
SGS Rev.00
ECG ATT2 Pg. 2 of 4 C. EMERGENCY COORDINATOR DUTIES Initials
- 1. IF NOT done previously, EC THEN DIRECT the OSC Coordinator to ACTIVATE the OSC IAW EPIP 202S, OSC Activation and Operations.
- 2. IMPLEMENT EPIP 102S, Alert, while continuing in this attachment.
EC
.3. C01\1PLETE 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)
D. TURi~OVER
( ) 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 ALERT, THEN DOCUMENT the name of your relief below:
Name 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
- 2. ENSURE appropriate reports are made IAW Section ill, Reporting, of this SNSS attachment.
SGS Rev. 00
ECG ATI2 Pg. 3 of 4 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
- 2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).
.., 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 OM be prepared. Provide this attachment and any other supporting documentation to the LERC.
- 5. PREP ARE required reports.
LERC Report or LER N u m b e r - - - - - - -
- 6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.
Rev. 00 SGS
ECG ATT2 Pg. 4 of 4 INITIAL CONTACT MESSAGE FORM I. THIS IS , COMMUNICATOR IN THE 0 CONTROL ROOM (NAME)
- - - - 0 TSC AT THE SALEM NUCLEAR GENERATING STATION, UNIT NO. _ _
II.
0 THIS 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 ICMF)
NOTE: Radiological Release is defined as: Plant Eftluent >Tech Spec Limit of2.42E+05 µCi/sec Noble Gas or 2. lE+ol µCi/sec 1-131.
SGS Rev. 00
ECG ATI3 Pg. 1 of 5 ATTACHMENT 3 SITE AREA EMERGENCY I. EMERGENCY COO RD INATOR (EC) LOG SHEET Initials A. DECLARE A SITE AREA EMERGENCY AT SALEM UNIT _ __
EC EAL #(s) _ _ _ _ _ _ _ _ _ _ _ __, - - - - - - -
Declared at - - - - - - hrs on - - - - - -
time date B. NOTIFICATIONS
( ) 1. CALL communicators to the Control Room.
( ) 2. C01\1PLETE the INlTIAL CONTACT MESSAGE FOR..rvf (ICMF) (last page of this attachment).
( ) 3. PROVIDE the ICMF to the Communicator (CMl) and DIRECT the C\.11 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 Salem. Please IMPLEMENT EPIP 204S, Salem 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 Hope Creek SNSS. (NETS 5224; DID 3027, 3059)
- a. PROVIDE a briefing on the SAE conditions.
- b. DIRECT implementation ofEPIP IOIH, Section 3.2.
- 7. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.
SGS Rev. 00
ECG ATI3 Pg. 2 of 5 C. EMERGENCY COORDINATOR DUTIES Initials
- 1. IF NOT done previously, EC THEN DIRECT the OSC Coordinator to ACTIVATE the OSC IAW EPIP 202S, OSC Activation and Operations.
- 2. IF the Emergency Coordinator is the EDO or SNSS, SNSS!EDO THEN REFER TO EPIP 103S, 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)
( ) b. NOTIFY EOF Staff of the change in classification.
- 4. COMPLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmital by EC the CMI 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. CO:MPLETE a~ IC:MF (ECG Attachment 3) for PAR UPGRADE.
( ) b. PROVIDE the ICMF to the CMI and DIRECT the CMI to IMPLEMENT a new ECG Attachment 6 for PAR UPGRADE notifications.
SGS Rev. 00
ECG ATI3 Pg. 3 of 5 Initials 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. TER"1L~A TION
- 2. ENSURE appropriate reports are made IAW Section II, Reporting, of this SNSS attachment.
SGS Rev. 00
ECG ATI3 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
- 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.
SGS Rev. 00
ECG ATI3 Pg. 5 of 5 INITIAL CONTACT MESSAGE FORM I. THIS IS
- - - - - - - - - , COMMUNICATOR IN THE 0 CONTROL ROOM (NAME) DTSC DEOF AT THE SALEM NUCLEAR GENERATING STATION, UNIT NO. _ __
~---------------------------------------------------------
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:
Ilb.
0 THIS IS NOTIFICATION OF A PROTECTIVE ACTION RECOMMENDATION UPGRADE WHICH WAS MADE AT HRS ON - - - - - -
(2-l HOlJR CLOCK) (DATE)
Reason for PAR Upgrade:
~
III. D NO RADIOLOGICAL RELEASE IS IN P.ROGRESS. 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. 0 NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TIME 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 Eftluent >Tech Spec Limit of 2.42E+05 µCi/sec Noble Gas or 2. lE+Ol µCi/sec I-131.
SGS Rev. 00
ECG ATT4 Pg. 1 of 7 ATTACHMENT 4 GENERAL EMERGENCY I. EMERGENCY COORDINATOR CECl LOG SHEET
~\
Initials A. DECLARE A GENERAL EMERGENCY AT SALEM UNIT - - -
EC EAL #(s) _ _ _ _ _ ____, _ _ _ _ _ _ _ __, - - - - - - - - -
Declared at hrs on
~------ - - -date time 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 COMPARE the two P ARs and choose the most appropriate for inclusion on the ICMF.
( ) 3. COMPLETE the INITIAL CONTACT MESSAGE FORM (ICMF) (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.
SGS Rev. 00
ECG ATI4 Pg. 2 of 7 Initials
- 6. IF NOT done previously, SNSS NOTIFY the I.T.0.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 Salem. Please IMPLE:rvIENT EPIP 204S, Salem 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 Hope Creek SNSS. (NETS 5224; DID 3027, 3059)
- a. PROVIDE a briefing on the GE conditions.
- b. DIRECT implementation ofEPIP IOIH, 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 202S, OSC Activation and Operations.
- 2. IF the Emergency Coordinator is the EDO or SNSS, SNSS/EDO THEN REFER TO EPIP 104S, 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 40 I, ERM Response, AND
( ) a. NOTIFY the EDO of General Emergency details;
- Time of declaration
~ EAL exceeded (Basis)
( ) b. NOTIFY EOF Staff of the change in classification.
- 4. COMPLETE and APPROVE the NRC Data Sheet for transmittal (Attachment 5) by EC the CM 1 within 60 minutes.
SGS Rev. 00
ECG ATI4 Pg. 3 of 7 Initials
- 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
( ) 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 GE, THEN DOCUMENT the name of your relief below:
Name time E. TERMINATION
- 2. ENSURE appropriate reports are made IAW Section II, Reporting, of this SNSS attachment.
SGS Rev. 00
ECG ATI4 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.
Initials
- 1. PREPARE an Action Request (AR).
SNSS
- 2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).
~. 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.
SGS Rev. 00
ECG ATI4 Pg. 5 of 7 PREDETERMINED PROTECTIVE ACTION RECOMMENDATION CHART PAR REQUIRED FOR GENERAL EMERGENCY GE BASED ON 10 EVACUATE ALL SECTORS 0- 5 MILES POINTS ON Yes
>----..+ EVACUATE DOWNWIND+/- 1 SECTOR 5-10 MILES BARRIER SHELTER ALL REMAINING SECTORS 5-1 O MILES TABLE No
'---------..+ EVACUATE ALL SECTORS 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 EVACUATE IN THE ABOVE SELECTED PAR.
SGS Rev.00
- cc ATT -+
RECOM~fE~DED PROTECTIVE ACTIONS WORKSHEET Pg 6 of 7 WI~D DIRECTION FROM PAR AFFECTED SECTORS DEGREES
- .3-+/-9 - 011 COMPASS N
NNE
-+ SSE s
DOWNWIND ::1 SECTOR s
SSW SSW SW 011 - O:J-t 03-t - 056 NE SSW - SW - WSW 056 - 0?9 ENE SW - WSW - w 079 - 101 E WSW - w - W);Vl 101 - 124 ESE w - WNW - :;;w 12-+/- - 146 SE WNW - NW - ~\W l-t6 - 169 SSE NW - \NW - \
169 - 191 s NNW - N - :\\£ 191 - 214 SSW N - NNE - \E 21-l- - 2:36 SW NNE - :-JE - E~E 2:36 - 259 WSW NE - ENE - E 2:59 - 281 w E~E - E E:3E 28 l - :30-t \V\W E - E.3E - .SE
-~i:)...J. - :3'.26 \\\' E5E - ~r
- - ~
- .~*.::.*[
2~~G - 3~9 \\\\' SE .~SE - ....:
\OTE co:-;::*IDER ..\DDI:\G A SECTOR TO THE PAR IF THE WIND DrRECTION I FROM) IS WITHIN
=:3° OF A SECTOR BOCNDARY LINE.
2a 1* L I "'-... \ I I I / \ _..t 079' WESTI I ~ I IEAST 259' r- ' ./ I I \ '\ "-.. I -, 101' 191' SOUTH 169' SGS Rev. 00
ECG ATI4 Pg. 7 of 7 INITIAL CONTACT MESSAGE FORM I. TIDS IS , CO.MMUNICATOR IN THE 0 CONTROL ROOM (NAME) 0 TSC DEOF AT THE SALEM NUCLEAR GENERATING STATION, UNIT NO. _ __
Ila. D TlilS IS NOTIFICATION OF A GENERAL EMERGENCY WfllCH WAS DECLARED AT ON - - - - - - - -
(TIME - 24 HOUR CLOCK) (DATE)
EAL #(s) _ _ _ _ _ _ _ _~ --------~-------
DESCRIPTION OF E V E N T : - - - - - - - - - - - - - - - - -
Ilb.
0 TIDS IS NOTIFICATION OF A PROTECTIVE ACTION RECOMMENDATION UPGRADE WIDCH 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 DIRECTION (From): WIND SPEED:
(From MET Computer) (DEGREES)
(MPH)
IV. Sectors Dist-Miles 0 WE RECOMMEND EVACUATION AS FOLLOWS 0 WE RECO:MMEND SHELTERING AS FOLLOWS EC Initials (Approval to Transmit ICMF)
NOTE: Radiological Release is defined as: Plant Effluent> Tech Spec Limit of 2.42E+05 µCi/sec Noble Gas or 2. IE+ol µCi/sec 1-131.
SGS Rev. 00
ECG ATT 5 Pg. 1 of 7 ATTACHMENT 5 NRC DATA SHEET COMPLETION REFERENCE \~.
I. INSTRUCTIONS 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 ~ttachment) each time you are directed to complete it (i.e., each change in classification or new event, begin again)
B. COMPLETE 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.
SGS 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 trip 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 affected 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 SG level is being maintained by the Aux. Feed water system) 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.
SGS 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit.
SGS Rev.CO
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.
SGS Rev. 00
l ECG ATT. 5 NltC DATA SHEET CPage l of _ l Pg. 5 of 7
- .~:1; :~~;~~ -iv.E ::..CIL!TY: Ur-.IT CAL-~2'S ~t.~E SALEM ~~t-.ERATING S~ATION E .~ :liTE EVENT TIME POWEP. I MODE ID.ORE E'vENT . POWER / MODE AFiEP EvEW EASTE.~N TIME ZONE EVENT CLASSIFICATION (Check One)
~ENERAL EMERGENCY ALERT 1HR 10CFR50.72:b)(1) *( ) 1-!R SEC:JRITY / SAFEGUARDS SITE AP.EA Ef.'ERGENCY UNUSUAL EVENT 4HR 10CFR50.72:b)(2) *( ) TRANSPORTATION EVE Ni OTHE~ (DESCRIBE):
- FC~ ~Oh-[M~P.G[NCIES P~OVICE 7HE SPECl'IC Sl:8°ART NUWBER Of THE IOCrRS0.72 R~PORTING REQUIREMENT 'ROM TH ECG INITIA1.NG CONOl.ION STATEMENT.
EVENT DESCRIPTION Include systems affected, actuo:ions & their initioting signals, causes, effect of event on plant, actions taken or planned, etc.
'l~te anything unus .. al or not Jnderstood. Indicate systems and safely-related ecuipment that are not ope~ational.
!Use a :ontinuation ocge if mo*e room is nee~ed)
Complete cnly 1! event includes an )
RCS/TUBE LEAlC DATA ( RCS or SG tube leak LOC/.ilON OF LEAK (e.g. SG, VALVE, PIP[, e t c . ) : - - - - - - - - - - - - - - - - - - - - - - -
T ME & )ATE LEAK STARTED: ON TIME DATE LEAK RA TE: gpm or gpd. T/S LEAK LIMITS: - - - - - -
LAST KNOWN COOLANT ACTIVITY: PRIMARY (DEi - µ.Ci/cc) - - - - - - SECONDARY (GBG - µ.Ci/cc)
HAS Tf'IS L~AK A SUDDEN OR LONG-TERM DEVELOPMENT?
NOTIFICATIONS CRGANIZATION NOTIFIED m No if ORGANIZATION NOTIFIED YES NO w ORGANIZATION NOTIFIED YES NO W~f NRC RESIDENT STATE OF NEW JERSEY STATE OF DELAWARE LOCAL (LAC TOWNSHIP) OTHER GOVERNMENT AGENCIES MEDIA / PRESS RELEASE MODE OF D?ERATION UNTIL CORRECTED: -I ESTIMATED RESTART DATE: ADDITIONAL INF'O ON PAGE 2?
Add;tional lnforrr.otion for Non-Emergency Nolifi:ations:
Re::iortcble Action ~eve! (RAL #) _1"""1.._._ _ SNSS/EC APPROVAL TO TRANSMIT SGS Rev. 00
ECG ATT. S NRC DAT A SHEET CPaqe 2 of _. l Pg. 6 of 7 NOTIFICATION DATE/TIME: - - - - - - - - -
RADIOLOGICAL RELEASE DATA: (This section is only required to be completed if a release exceeding Tech Specs is in progress QI has already occurred).
Check J ALL correct statement. and provide to the NRC.
_There is/was a gaseous release above Tech Spec limits in progress (Tech Spec Limit: Noble Gos = 2.42E+OS µCi/sec).
_ There is/was an Iodine release above Tech Spec limits in progress (Tech Spec Limit: locine-131 = 2.1 E+01 µCi/sec).
_There is/was o 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 con be isolated.
_ The release pathway is monitored by the Radiation Monitoring System.
_ Areas evacuated onsi!e due to release concerns o r e : - - - - - - - - - - - - - - - - - - - - -
Station personnel hove received exposure above 10CFR20 limits.
_ Station personnel hove been contaminated lo an extent requiring offsile assistance to decon.
SPECIFIC RADIOLOGICAL PARAMETERS: (Provide current values) Current Time: _ _ _ _ _ _ hrs.
Total Release Rote Noble Gos (from SSCL) is: - - - - - - - - - µ.Ci/sec.
Total Release Role lodine-131 (from SSCL) is: _ _ _ _ _ _ _ _ _ _ µCi/sec.
RELEASE PATHWAY MONITORS: (Provide reodin~s and alarm setpoints only for those beiow :isted monitors in Alarm or !hat are included in the release pathway).
Monitor # and Name Current Reading Alarm Setpoint 2R41 D Noble Gas Effluent _ _ _ _ _ µ.Ci/sec 2.00E+04 µCi/sec 1R45B Mid Plant Vent Gos _ _ _ _ _ µ.Ci/cc 3.00E-02 µCi/ cc 1R45C High Plant Vent Gos -----µ.Ci/cc 1.00E+02 µCi/cc 1(2)R46 Highest Steam Line _ _ _ _ _ mR/hr 1.00E +01 mR/hr (R46A thru D) 1(2)R15 Condenser Air Ejector _ _ _ _ _ cpm _ _ _ cpm 1(2)R19 Highest S/G Slowdown -----Cpm _ _ _ 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 SGS Rev. 00
ECG ATT. 5 NRC DAT A SHEET !Page _ of Pg. 1 of 1 NOTIFICATION DATE/TIME: - - - - - - - - -
0 EVENT DESCRIPTION (Continued):
0 OTHER PERTINENT INFORMATION (Conlinued):
0 NRC EVENT UPDATE (Document additional information _to NRC due lo their request or as a resull of plant/ event slalus changes):
(Use a continuation page if more room is needed)
SNSS/EC APPROVAL TO TRANSMIT SGS Rev. 00
ECG ATT6 Pg. 1 of 8 ATTACHMENT 6 PR'ARY COMMUNICATOR LOG Table of Contents Pages 1 - 3 Notifications & Incoming Calls 4 Tennination 5-8 Communications Log 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.
Initials
- 1. OBTAIN an approved Initial Contact Message Form (ICMF) from the Emergency CMlffSCl Coordinator (EC).
/EOFl CAUTION Fifteen minute clock for notification starts at time event was declared
- 2. CALL each Organization or Individual identified on the Communications Log CMlffSCl (Pgs. 5 - 8) and READ the ICMF.
/EOFl
_ _ _ 3. 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.
SGS Rev.00
ECG ATT6 Pg. 2 of 8 Initials
- 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.
_ _ _ 4. FAX the ICMF to Group A.
CMlffSCl
/EOFl B. TURNOVER
--- 1. WHEN CONTACTED by the TSC (or EOF) in preparing for notifications CMlffSCl responsibilities, THEN PROVIDE the following information;
- Organizations/Individuals notified.
- Phone numbers or locations oflndividuals for updates or changes in status.
- - - 2. WHEN the EC function transfers to the oncoming facility, CMlffSCl 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, CMlffSCl THEN DIRECT the caller to contact the TSC (or EOF if activated).
_ _ _ 2. WHEN contacted by any State Agency Officials (listed here),
CMlffSCl
/EOFl 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 SGS Rev. 00
ECG ATT6 Pg. 3 of 8 Initials PERFORM the following;
( ) a. OBTAIN and RECORD; Agency Caller's Name Phone#
( ) b. READ the latest EC approved SSCL.
C. INCOMING CALLS (cont'd)
( ) 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, CMlrfSCl 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."
D. CONTINUOUSDUTIES
_ _ _ 1. ASSIST the CM2 gathering and faxing operational data.
CMI SGS Rev. 00
ECG ATT6 Pg. 4 of 8 Initials
- - - 2.
TSCl/EOFl.
ASSIST the TSC2 (or EOF2) in maintaining facility status boards
- 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, CMlffSCl 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 SGS Rev. 00
ECG ATT6 Pg. 5 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 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: EMRAD 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 CUMBERLAND COUNTY Primary: NETS 5403 Secondary: 455-8770 U.S. COAST GUARD (Speak Only to Duty Desk)
Primary: 215-271-4940 Secondary: 215-271-4800 SGS 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 SGS 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: Refer to Roster Secondary: (Contact One)
Chris Bakken Office: 2613 Home: 769-5420 Pager: 478-5016 Car: 230-8814 Nick Conicella Office: 2124 Home: 223-0975 Pager: 478-5035 Car: 230-8164 Jay Laughlin Office: 2907 Home: 935-8545 Pager: 478-5004 Car: 230-7995 Dennis Mccloskey Office: 5021 Home: 302-328-8520 Pager: 573-1417 Car: 302-563-5008 PUBLIC INFORMATION SEE 70 MANAGER NUCLEAR NOTE 2 MIN. (Contact One)
Trish DuBrois Office: 1186 Home: 769-2454 Pager: 223-3012 Nancy Sooy Office: 1007 Home: 795-6831 Pager: 223-3393 NOTE 1 NOTIFY EDO for Unusual Events ONLY.
NOTE 2 After ENC activation, NOTIFY the ENC Manager (NETS -5300 or 609-273-1961)
SGS 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.
Charlie Marschall Office: 1078 or 935-3850 or 935-5151 Home: 610-444-0181 Pager: 772-4 7 42 Joe Schoppy Office: 1041 or 935-3850 or 935-5151 Home: 384-1365 Pager: 772-4742 Todd Fish Office: 1017 or 935-3850 or 935-5151 Home: 302-654-6612 Pager: 772-4742 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 AMElUCAN NUCLEAR INSURERS SEE 90 NOTE 4 MIN. (ANI) 203-561-3433 NOTE 3 Not required to notify External Affairs After the EOF is activated.
NOTE 4 Not required to notify ANI for Unusual Events SGS Rev. 00
ECG ATT7 Pg. 1 of 7 ATTACHMENT 7 PRIMARY COMMUNICATOR LOG (GE)
Table of Contents Pages 1 - 3 Notifications & Incoming Calls 4 Termination 5-7 Communications Log Emergency Classification: GENERALEMERGENCY or PARUPGRADE N a m e : - - - - - - - - - - - - - Position: CM1 /TSC1/ EOF1 (Print) (circle)
A. NOTIFICATIONS NOTE A new Attachment 7 is required to be implemented if the PAR is changed.
Initials
_ _ _ l. OBTAIN an approved Initial Contact Message Form (ICMF) from the Emergency cM1rrsc1 Coordinator (EC).
/EOFl CAUTION 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 CM1rrsc1 (Pgs. 5 - 7) and READ the ICMF. If needed obtain assistance from Secondary
/EOFl Communicator.
_ _ _ 3. FAX the ICMF to Group A.
cM1rrsc1
/EOFl SGS Rev. 00
ECG ATT7 Pg. 2 of 7
_ _ _ 4. - 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.
- 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
- 1. WHEN CONTACTED by the TSC (or EOF) in preparing for notifications CMlffSCl 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, CMl!TSCl 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, CMlffSCl THEN DIRECT the caller to contact the TSC (or EOF if activated).
Rev.00 SGS
ECG ATT7 Pg. 3 of 7 Initials C. INCOMING CALLS (cont'd)
- 2. WHEN contacted by any State Agency Officials (listed here),
CMlffSCl
/EOFl 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, CMlffSCl READ the appropriate message below;
/EOFl
( ) a. IF the ENC is mu activated (Unusual Event), say; "You are requested to contact the Nuclear Communications Office at any of the following numbers; 609-339-1001, -10069 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."
Rev.00 SGS
ECG ATT7 Pg. 4 of 7 Initials 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 .
--- 3. IF the telecopier is NOT working correctly, CMl THEN CALL the TSC - Emergency Preparedness Advisor (EPA) for assistance.
E. TERMINATION/REDUCTION
- l. WHEN the Emergency has been terminated or reduced in classification, CMlrfSCl 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, CM1rrsc1 THEN FORWARD this document and all completed Forms to the SNSS (TSS/SSM).
/EOFl SGS Rev. 00
ECG ATT7 Pg. 5 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME GENERAL EMERGENCY OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NEW JERSEY STATE POLICE/OEM 15 Primary: NETS 5400 MIN. Secondary: 882-4201 Call Back:
BACKUP: EMRAD DELAWARE STATE POLICE/DEMA Primary: NETS 5406/5407 Secondary: (SP)302-739-5851 or (DEMA)302-834-7250 Call Back:
BACKUP: NAWAS LAC TOWNSHIP li!/i!ij Primary: NETS 5404 "'
Secondary: 935-7300 Call Back:
SALEM COUNTY Primary: NETS 5402 Secondary: 769-2959 Call Back:
Backup: EM RAD ,;:"
CUMBERLAND COUNTY r:::11 Primary: NETS 5403 Secondary: 455-8770 Call Back: rt Backup: EMRAD NEW CASTLE COUNTY Primary: NETS 5408 Secondary: 302-738-3131 KENT COUNTY Primary:
~
NETS 5409 Call Back:
I Secondary: 302-678-9111 Call Back:
U.S. COAST GU.ARD 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.
SGS Rev.00
ECG ATT7 Pg. 6 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME 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)
~'i Trish DuBrois Office: 1186 Horne: 769-2454 :..-:-..-
Pager: 223-3012 ...
I Nancy Sooy Office: 1007 Horne: 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)
SGS Rev. 00
ECG ATT7 Pg. 7 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME GENERAL EMERGENCY OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NRC RESIDENTS 75 (Contact One)
MIN. Charlie Marschall Office: 1078 or 935-3850 or 935-5151 Home: 610-444-0181 Pager: 772-4742 Joe Schoppy Office: 1041 or 935-3850 or 935-5151 Home: 384-1365 Pager: 772-4742 Todd Fish Office: 1017 or 935-3850 or 935-5151 Home: 302-654-6612 Pager: 772-4742 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 Ed Johnson Office: 1486 Home: 678-2257 Pager: 478-5040 AMERICAN NUCLEAR INSURERS 90 MIN. (ANI) 203-561-3433 NOTE 2 Not required to notify External Affairs After the EOF is activated.
SGS Rev. 00
ECG ATT8 Pg. 1 of 8 ATIACHMENT 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 fTSC2/ 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 notifications.
CM2ffSC2
/EOF2
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.
Time: - - -
( ) b. ACTIVATE ERDS within 60 minutes from the Affected Unit's SPDS terminal;
- 1) PRESS <UNIT MASTER MENU> key.
- 2) PRESS <ERDS> key.
- 3) FOLLOW screen prompts.
SGS Rev. 00
ECG ATT8 Pg. 2 of 8 Initials A. NOTIFICATIONS (cont'd)
COMPLETE a Station Status Checklist (SSCL) Form;
- - - 4. (
CM2ffSC2 ) a. OBTAIN SNSS (TSS/SSM) assistance, as needed for Pg. l.
/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 A gencies 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 completed NRC Data Sheet and FAX form to Group B.
CM2ffSC2
/EOF2
- 6. REPEAT Step 4 approximately every halfhour OR IMMEDIATELY for significant CM2ffSC2 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.
- 8. IF Available for other duties CM-2 THEN upon the SNSS request MAN the Ops Data line.
B. DATA COLLECTIONrrRANSMISSION
- 1. WHEN in an ALERT or higher emergency OR AFTER significant changes in
--- plant status; CM2 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.
SGS Rev. 00
ECG ATT8 Pg. 3 of 8 Initials B. DATA COLLECTIONtrRANSMISSION (cont'd)
IF requested by the TSC,
-CM2- - 2. THEN COMPLETE the Operational Status Board (OSB) Fann every 15 minutes; (TSS may modify the frequency or data list as appropriate)
( ) a. OBTAIN Licensed Operator review.
( ) b. FAX to Group C.
_ _ _ 3. ENSURE the Facility OSB and MEES Status Boards are updated as follows; TSC2/EOF2
( ) a. OBTAIN OSB Data from SPDS "Unit Master Menu."
( ) b. IF SPDS is Out of Service, THEN REQUEST CM2 to perfonn step B.2, above. (data set and frequency of updates may be revised by the TSS based on event circumstances)
( ) c. WHEN significant changes in plant status occur,
. THEN REQUEST CM2 to perform step B.1, above.
--- 4. WHEN the emergency is terminated, CM2ffSC2 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, CM2ffSC2 THEN DIRECT the caller to contact the TSC (or EOF if activated).
_ _ _ 2. WHEN contacted by any State Agency Officials (listed here),
CM2!fSC2
/EOF2 DEMA - Delaware Emergency Management Agency AAAG - Delaware Accident Assessment Advisory Group BNE - NJ Bureau of Nuclear Engineeri~g DEP - NJ Department 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.
SGS Rev. 00
ECG
. ATT8 Pg. 4 of 8 Initials C. INCOMING CALLS (cont'd)
STATE OFFICIALS
( ) 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, CM2ffSC2 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 SPDS terminal of the affected Unit AND PROCEED as follows;
- a. PRESS <UNIT MASTER MENU> key.
- b. PRESS <ERDS> key.
- c. FOLLOW screen prompts.
- d. WHEN completed, NOTIFY the SNSS.
SGS Rev.00
ECG ATT. 8 Pg. 5 of 8 SALEM UNIT_
MAJOR EQUIPMENT AND ELECTRICAL STATUS
- ,Y = IN SERVICE DATE: ~~~~~~
- ! N
- OUT OF SERVICE
- ' cmcLE UNAVAILABLE EQUIP UPDATE TIME:
' i
,COOLING
- SYSTEMS iELE~CAL YIN ECCS SYSTEMS ELECTRICAL FEED YIN CONT. CONTROL SYSTEMS
- AUX FD 1 AID CHARGING* 1 B9D CONT. SPRAY 1 AID PUMPS PUMPS PUMPS 2 a BID a C9D C2D 3~
3 A7X
- CFCU f HI I ~ LOWi A3X
~~ H ~I SERVICE WATER 1 SAFETY PUMPS INJ ABO I 1 I A4X A2X I I
aI en a r ~~
1 I
! CSD I B3X : '
PUMPS 3 i B3D 3 I g~~ I cax RHR 1' A7D '
B7X '
I 4 BBD PUMPS 4 BBX BSX 2* B7D 5 3D C7X CSX S CBX 6 BD ELECTRICAL STATUS I Y/N YIN IS OFFSITE AC ; IODINE 1 . G7X COMP. Al OD : POWER AVAILABLE? I : REMOVAL COOLING i !
a, E7X PUMPS 2! BlOD '
lI EMER. DIESEL RUN ILoADI l sl Cl OD I EDG A ! [Ha l : AlSX I
B I RECOM aj BlSX REACTOR COOLANT PUMPS 1
aI 31 i H4D E4D F4D p
i I 1
3 GAS TURBINE ELEC DISTRIBUTION c
Y/N l MISC. EQUIPMENT FIRE PUMPS (DIESEL)
.Y/N:
l i a '
4! G4D AVAILABLE?
- COND.
VITAL BUS A STATION AIR COMP. IY/N 1 HID PUMPS B l lH6D i 2 EID I c 2 2GID I 3 FlD !
GROUP BUS E 3 lGID
<UU I (Ua) F EMERGENCY AIR COMP. iY/Ni I
G 1 1Cl4X I CIRC lA aAD/H'lD WATER lB 7BD/F'ID H a 2C14X !
PUMPS 2A 3AD/E'lD COMMEN'l'S:
2B BBD/G7D 3A 4AD/E3D 3B BBD/G3D LICENSED OPERATOR REVIEW:
INITIALS SGS Rev. 00
OPERATIONAL STATUS BOARD - SALEM UNIT#:[_~
l/)
C>
UPDATE: L - 1_ __ _ , _ _ =:J l/) TIME DATE IV. C.V,(,$,
I. EMERGENCY CORE COOLING SYSTEM CENT. CHRG. PUMP FLOW L __ _-]GPM LETDOWN FLOW EJGPM GPM CHARGING FLOW SI PUMP FLOW SI PUMP FLOW
- _1
GPM NO.
RHR PUMP FLOW # 1 NO. 2 SG LEVEL % (NR or WR)
RHR PUMP FLOW # 2 GPM EJ % (NR or WR)
NO. 3 SG LEVEL RWST LEVEL I IFT NO. 4 SG LEVEL % (NR or WR)
II. CONTAINMENT NO. 1 SG PRESS. PSIG CONT. PRESSURE ~SIG NO. 2 SG PRESS. PSIG CONT. TEMP {AVG) NO. 3 SG PRESS. PSIG CONT. H2 CONCEN. % NO. 4 SG PRESS. PSIG CONT. SUMP LEVEL % NO. 1 SG FEED FLOW % or LBS/HR CONT. RAD (HI RANGE) NO. 2 SG FEED FLOW % or LBS/HR R44A R/hr NO. 3 SG FEED FLOW % or LBS/HR R44B E J R/hr NO. 4 SG FEED FLOW % or LBS/HR AFST LEVEL %
THERMOCOUPLE (HOTTEST) 1---~
F VI. MJSC_, __ TANKS WASTE HOLD-UP TANK WASTE HOLD-UP TANK LEVEL
- _2
_1§% %
- THERMOCOUPLES >1200 F WASTE MONITOR HUT %
Tc LOOP _ 1 F 1-----i VII. SSC::L .INFORMATION YES or NO Tc LOOP _2 F 1-----<
OFFSITE POWER AVAILABLE?
Tc LOOP _3 F 1-----i lWO OR MORE DIESELS AVAILABLE?
Tc LOOP _4 F
- Tave (AUCTIONEERED) 1-----
F IS fHE CONTAINMENT ISOLATED?
PZR/RCS PRESSURE PSIG
--- IS IT CAPABLE OF BEING ISOLATED?
PZR LEVEL {HOT) %
Th LOOP _1 F VIII. SIGN!F!C.At:H__P.LANL.LYENTS
.-:-..* ==-====--~-~==------ ~
Th LOOP _2 F -u l> 1"'1
<ll --in Th LOOP _3 F . ;--10
- o Th LOOP _4 F ()) CXl ro
< RX PWR/NEUTRON FLUX %/A/CPS
- NHt ri rm Rc:r*s- ~RE RUNNr~~~ Tav:--oN--r;:;-E:-C:-;;-~y-R-;;~o~sol:[~s--1~~~~~0~ 0....,
0 SUBCOOLING MARGIN F LICCNSED OPERATOR REVIEW [=-~-:~---_] CXl 0
INITIALS
ECG ATT. 8 Pg. 7 of 8 STATION STATUS CHECKLIST SSCL I (Pg. 1 of 2)
Operational Information SALE~i GENERATING STATION Unit No. __ Message Date_ _ _ Time_ __
Transmitted By: Name_ _ _ _ _ _ _ _ __ Position:
( CR/TSC/EOF)
- 1. Date and Time Event Declared: Date - - - Time (24 hr clock)
- 2. Event Classification: D Unusual Event D Site Area Emergency D Alert D General Emergency
- 3. Cause of Event: Primary Initiating Condition used for declaration E.-\L #( s)
Description of the event - - - - - - - - - - - - - - - - - - - -
. Status of Reactor: u Tripped/Time 0 At Power ' Startup
- Hot Standby D Hot Shutdown D Cold Shutdown C Refuel
- >. PZR/RCS Pressure psig Core Exit TC 0 F
Hottest
- 6. Is offsi te power available? D YES LJ :\'O
- 7. Are two or more diesel generators operable? D YES D NO
- 8. Did any Emergency Core Cooling Systems actuate? DYES D NO
- 9. Containment:
A. Has the Containment been isolated? DYES D NO B. Is it capable of being isolated? DYES 0 NO
- 10. Other pertinent information - - - - - - - - - - - - - - - - - -
Approved: _ _ _ _ _ _ _ _ __
EC or TSS or SSM SGS Rev. 00
ECG
_HT. 8 STATION STATUS CHECKLIST Pg. 8 of 8
( PAGE 2 OF 2 )
RADIOLOGICAL INFORMATION SALEM GENERATING STATION UNIT NUMBER:_ _ CALCULATION TIME*. _ _ __ DATE:._ __
- 1. GASEOUS RELEASE> TECH SPEC (T /S) LIMITS:
(T /S LIMITS: 2.42E+05µCi/sec NG or 2.1 OE+01 µCi/sec IODINE)
YES: [ J RELEASE START TIME: DATE:._ _ _ __
NO: [ J 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) DELTA T: (deg C)
F. VENT PATH OF RELEASE: R41 [ ] R45B/C [ ] R44 [ ] R46 [ ]
G. NG RELEASE RATE: R41 _ _ __ R45B/C R44'-----
R46 (µCi/sec)
H. 1-131 RELEASE RATE: R41____ R45B/C._ _ _ __ R44._ _ __
R46 DEFAULT (µCi/sec) (circle if default)
I. TOTAL RELEASE RATE NOBLE GAS: (µCi/sec)
J. TOTAL RELEASE RATE IODINE-131: (µCi/sec)
'.'.. PROJECTED OFFSITE DOSE RATE CALCULATIONS:
=:1STANCE XU/Q TEDE DOSE THYROID- THYROID-Fr\OM VENT RATE ( 4 DAY) CDE RATE CDE DOSE (IN MILES) ( 1/M2) (MREM/HR) (MREM) (MREM/HR) (MREM)
MEA 0.79 2.00 LPZ 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 SGS Rev. 00
ECG ATT9 Pg. l of 3 ATTACHMENT 9 NON-EMERGENCY NOTIFICATIONS REFERENCE (SALEM)
I. INSTRUCTIONS NOTE 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 .
.-.\. REFER to Section II of this Attachment and >rOTIFY the required Individuals1 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 SGS Rev. 00
ECG ATT 9 Pg. 2 of 3 IL 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 Chris Bakken 2613 769-5420 478-5016 230-8814 James Webster 2985 935-7678 478-5236 230-5671 Mike Gwirtz 2622 358-7160 223-3830 230-5606 GENERAL MANAGER Dave Garchow 2900 610-274-3250 478-5096 230-5894:
Chris Bakken 2613 769-5420 478-5016 :J0-881-+
GOv~R...1\IMENTAGENCY PRIMARY# SECONDARY#
LAC DISPATCHER NETS x5404 935-7300 935-8127 (FAX)
~C OPERATIONS CENTER (ENS)301-816-5100 301-951-0550 301-816-515 l(F AX)
~C REGION ONE OFFICE 610-337-5000 TITLES/NAMES WORK# HOME# PAGER#
NRC RESIDENTS Charlie Marschall 1078 or 935-3850 444-0181 772-4742 Joe Schoppy 1041 or 935-3850 384-1365 772-4742 Todd Fish 1017 or 935-3850 302-654-6612 772-4742 NRC Office 2962 or 935a5151 SGS Rev. 00 I
I
ECG ATT9 Pg. 3 of 3 IL TELEPHONE NUMBER REFERENCE (cont'd)
TITLES/NAMES WORK# HO:ME# PAGER#
PUBLIC INFO MGR Trish DuBois 1186 769-2454 223-3012 l
Nancy Sooy* 1007 795-6831 223-3393 E'MERG 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 Max Lefevre 1243 263-7677 478-5094 Ross Bell 1239 455-7435 478-5213 Ed Johnson 1486 678-2257 478-5040 RADIOLOGICAL SUPPORT REPRESENTATIVE John Russell 2410 451-0845 478-5082
'.\!ark Simpson 2443 302-998-4 792 4 78-53 78 Bill \Veckstein 1558 455-3237 478-5186 RADIATION PROTECTION MANAGER Eric Katzman 2659 468-0709 478-5204 Bill Hunkele 2617 455-1583 478-5179 Dave Ruyter 2625 299-9487 478-5143
~CLEAR LICENSING DUTY PAGER HOLDER ----- ----------------- 779-4227 Gabe Salamon 5296 610-274-2297 573-1819 Dave Powell 2002 302-239-9912 573-2358 ENVIRONMENT AL LICENSIN~ (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* 1477 547-3795 573-8419 Paul Behrens
- 1577 691-4766 573-2496 Ed Keating* 1459 678-8160 573-4139
- For Spills, Hazmat, NQI Protected Aquatic Species SGS Rev.00
ECG ATilO Pg. l of 3 ATTACHMENT 10 ONE HOUR REPORT NRC REGIONAL OFFICE INSTRUCTIONS (SALEM 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 an NRC Data Sheet.
( ) OBTAIN a copy from ECG Attachment 5.
( ) OBTAIN assistance from Radiation Protection personnel, as needed.
( ) OBTAIN SNSS approval.
- 2. NOTIFY NRC Region I Office of the event within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
( ) 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 SGS Rev. 00
ECG A.TI 10 Pg. 2 of 3 Initials
- 5. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).
~~~~~~~~~~~~~~~~~
notified at hrs name time
- 6. NOTIFY the Public Information Manager (PIM) - Nuclear.
~~~~~~~~~~~~~~~~~
notified at hrs name time
- 7. NOTIFY Nuclear Licensing.
~~~~~~~~~~~~~~~~~
notified at hrs name time
- 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.
SGS Rev. 00
ECG ATilO Pg. 3 of 3 II. REPORTING
- 1. ENSURE that an Action Request (AR) is prepared.
SNSS AR# _ _ _ _ _ _ _ _ __
- 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.
SGS Rev. 00
ECG ATI' 11 Pg. 1 of 3 ATTACHMENT 11 ONE HOUR REPORT (COMMON SITE)
SECURITY/SAFEGUARDS \0 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 Initials
- 1. NOTIFY the Hope Creek SNSS (NETS x5224; DID 3027, 3059).
- 2. DETERMINE which Station SNSS will implement this attachment.
- 3. IF the Salem SNSS is responsible for this notification, THEN IMMEDIATELY CONTINUE with this attachment.
- 4. IF the Hope Creek SNSS will implement this attachment, THEN NO further actions are required by Salem except to lend assistance as necessary in restoring the lost equipment or capabilities.
INSTRUCTIONS (SALEM 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 : - - - - - - - - - - - - ----
IL NOTIFICATIONS
- 1. COMPLETE an NRC Data Sheet.
( ) OBTAIN a copy from ECG Attachment 5.
( ) OBTAIN assistance from Security personnel, as needed.
( ) ENSURE SNSS approval.
SGS Rev. 00
ECG ATill Pg. 2 of 3
- 2. NOTIFY the NRC Operations Center of the event within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
-- ( ) 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 (P') - 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.
notified at hrs name time SGS Rev. 00
ECG ATIU Pg. 3 of 3 ill. REPORTING Initials
- 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 Supenrisor (NSSS);
OM ( ) FORWARD this attachment and any other supporting documentation received from the SNSS.
( ) REQUEST a written report (required 30 days after the event).
- 5. PREP ARE the required Safeguards Event Report (30 day) !AW Security Contingency NSSS Plan Procedure, SCP-14.
- 6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).
- 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.
Rev.00 SGS
ECG ATT 12 Pg. 1 of 3 ATTACHMENT 12 ONE HOUR REPORT - NRC OPERA TIO NS INSTRUCTIONS (SALEM 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 by: - - - - - - - - - - - - Date:
I. NOTIFICATIONS
- 1. COiv1PLETE an P.'RC 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
( ) 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
notified at hrs
~
name time
- 5. NOTIFY the Public Information Manager (PIM) - Nuclear.
notified at hrs
~
name time SGS Rev.00
ECG AIT12 Pg. 2 of 3 Initials
- 6. NOTIFY Nuclear Licensing.
name time
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."
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.
SGS Rev. 00
ECG ATI 12
. Pg. 3 of 3 IL 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 ~umber - - - - - - - -
6 FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (>-~1-R).
LERC
- 7. ENSURE offsite (state and local) reporting requirements are met.
~R
- 8. FORWARD this Attachment/LER package to the Central Technical Document Room
~1NLR for microfilming.
Rev.00 SGS
ECG ATI13 Pg. 1 of 7 ATTACHMENT 13 FOUR HOUR REPORT CONTAMINATION EVENTS OUTSIDE OF THE RCA INSTRUCTIONS (SALEM 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 by: - - - - - - - - - - - - ----
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 Check.list (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 hrs name time SGS Rev. 00
ECG ATf 13 Pg. 2 of 7 Initials
- 5. NOTIFY the Hope Creek SNSS (NETS x5224; DID x3027, or x3059)
( ) 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 IAW the DPCC/DCR Plan be made as required.
name time 7 COJ\1PLETE 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
( ) RECORD additional information provided to the NRC on the NRC Data Sheet.
name time
- 11. NOTIFY the NRC Resident Inspector.
name time SGS Rev. 00
ECG ATT13 Pg. 3 of 7 Initials
- 12. IF. NOT done previously, THEN NOTIFY the Operations Manager (OM).
name
- 13. NOTIFY Nuclear Licensing.
name
- 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.
SGS Rev.00
ECG ATI13 Pg.4of7 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 (l\!INLR).
LERC
- 7. ENSURE offsite (state and local) reporting requirements are met.
rvtNLR
- 8. FORWARD this Attachment/LER package to the Central Technical Document Room rvtNLR for microfilming.
Rev. 00 SGS
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 monitorin~ at the Securitv Center.
- 5. NOTIFY the Salem RP Superintendent.
- 6. PROVIDE a briefing to the Hope Creek RP Superintendent and OBTAIN resource assistance (material and personnel), as needed.
SGS Rev. 00
ECG ATf 13 Pg. 6 of 7 ONSITE CONTAMINATION EVENT CHECKLIST (Page 2 or 3)
D. SUBSEQUENT ACTIONS Initials/
Date/Time
- 1. EST AB LISH 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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.
- 5. PERFORM isotopic analysis on several samples before decontamination activities be in.
- 6. REDUCE contamination< LLD, if reasonably achievable.
- 7. IF contamination CANNOT be reduced < LLD, THEN CONSIDER fixin the contamination to revent further s readin
- DOSE ASSESSMENT
- 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 havin~ 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 boundin~ case assessments)
- 12. PERFORM confirmatory WB Counts, as required.
- 13. COLLECT and PROCESS TLDs, as required.
SGS Rev. 00
ECG ATf 13 Pg. 7 of 7 ONSITE CONTAMINATION EVENT CHECKLIST (Page 3 or 3)
D. SUBSEQUENT ACTIONS (cont'd)
Initials/
- DOSE ASSESSMENT (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. OBI AIN 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 Salem SNSS after all items on the checklist have been addressed.
SGS Rev. 00
ECG ATI 14 Pg. l of 3 ATTACHMENT 14 FOUR HOUR REPORT - NRC OPERATIONS INSTRUCTIONS (SALEM 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 by: - - - - - - - - - - - -
I. ~OTIFICA TIO NS
- l. CO~fPLETE an \."RC Data Sheet.
( ) OBTAIN a copy from ECG Attachment 5.
( ) OBT A.IN assistance from Radiation Protection personnel, as needed.
( ) OBT A.IN SNSS approval.
- 2. ~OTIFY the NRC Operations Center of the event within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
( ) 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 S. NOTIFY the Public Information Mana1er (PIM) - Nuclear.
notified at hrs name time SGS Rev.00
ECG
..\TI l~
Pg. 2 of 3
- 6. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).
notified at hrs name time
- 7. NOTIFY Nuclear Licensing.
notified at hrs name time
- 8. NOTIFY External Affain.
notified at hrs name time 9 FAX the ~"RC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.
Rev.00 SGS
ECG ATI 14 Pg. 3 of 3 IL REPORTING
- l. ENSURE that an Action Request (AR) is prepared.
SNSS AR# _ _ _ _ _ _ _ _ __
- 2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).
- 3. REVlEW 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. PREPARE required reports.
LERC Report or LER ~umber - - - - - - - -
- 6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation\ \D.1.Ri LERC
- 7. ENSURE offsite (state and local) reporting requirements are met.
'.\1NLR
- 8. FORWARD this Attachment/LER package to the Central Technical Document Room
'.\~1.R for microfilming.
Rev.00 SGS
ECG ATf 15 Pg. 1 of 3 ATTACHMENT 15 ENVIRONMENTAL PROTECTION PLAN
\'0\
INSTRUCTIONS (SALEM 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
- 1. RECORD the Event D e s c r i p t i o n : - - - - - - - - - - - - - - - - -
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.
SGS Rev. 00
ECG ATI l5 Pg. 2 of 3 NOTE Required reports shall be made within the appropriate time limits from the Determination Time established in Step 2. above.
- 3. NOTIFY the NRC Resident Inspector within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
name time 4 IF the NRC Resident Inspector CANNOT be notified, THEN 'S"OTIFY the ~RC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
name t:r'.1e 5 IF NOT done previously, THEN NOTIFY the Operations Manager (OM).
name time SGS Rev. 00
ECG ATI 15 Pg~ 3 of 3 IL REPORTING
- l. ENSURE that an Action Request (AR) is prepared.
SNSS
- 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 ~umber - - - - - - - -
- 6. FORWARD this attachment to the Manager- Nuclear Licensing & Regulation (:\f\1.R)
LERC 7 ENSlJRE offsite (state and local) reporting requirements are met.
YINLR
- 8. FORWARD this Attachment/LER package to the Central Technical Document Room
'.\1NLR for microfilming.
Rev.00 SGS
ECG ATT16 Pg. 1 of 7 ATTACHMENT 16 SPILilDISCHARGE REPORTING \0\
CAUTION 15 minute notification to NJDEP may be required as identified in Step 4.
INSTRUCTIONS (SALEM 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 by: - - - - - - - - - - - -
I. ~OTIFICA TIO NS Initials
- l. 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. the Spill/Discharge was EITHER
- into a secondary containment
- onto an impervious surface AND the material CAN BE completely cleaned up; D b. 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. II., REPORTING (Pg. 6).
D c. OR IF OTHERWISE (more serious Spill/Discharge situation than above),
( ) THEN IMMEDIATELY GO TO Step 3 (next page).
SGS Rev. 00
ECG ATT 16 Pg. 2 of 7 NOTE DO NOT implement notification UNTIL directed to by EITHER Step 4 OR 5.
- 3. CO:MPLETE "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.!;!,
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 OBI AIN 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)
SGS 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 IMMEDIATELY (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:
D Call was made within 15 min. of discovery/ confirmation.
D Call was NOT made within 15 min., but was made within
- - - min. of discovery/confinnation.
( ) b. DIRECT Environmental Licensing to make any required notifications IAW the DPCC/DCR plan.
SGS Rev. 00
ECG ATT 16 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 PERFORM all of the following Notification Steps.
- 8. NOTIFY Hope Creek SNSS and provide description of the event.
name
- 9. NOTIFY LAC Dispatcher within 4 hrs.
name
- 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 houn.
( ) Use the NRC Data Sheet to record any additional information provided to the NRC.
- - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name time Rev. 00 SGS
ECG ATT 16 Pg. 5 of 7 Initials
- 12. Notify the NRC Resident Inspector.
name
- 13. NOTIFY Public Information Manager (PIM) - Nuclear.
notified at hrs name time
- 14. NOTIFY Nuclear Licensing.
notified at hrs name time
- 15. Notify External Affairs.
name
- 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).
Rev. 00 SGS
- 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. CONT ACT 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 the 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.
SGS Rev. 00
ECG ATT16 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 machine, 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 SpiIVDischarge."
This is (name)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, from Salem Generating Station.
My call back phone# is 609-339-5200 or 609-339- _ _ __
The Spill/Discharge location is: (provide specific location) at Salem 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 i s : - - - - - - - - - - - - - - - - - - - - - - - -
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 please?-----------~----------
SGS Rev. 00
ECG ATf 17 Pg. 1 of 4*
ATTACHMENT 17 FOUR HOUR REPORT \
FATALITY OR MEDICAL EMERGENCY \ \.J INSTRUCTIONS (SALEM 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 Initials
- 1. IF NOT done previously, THEN IMPLE~NT SC.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
~
name
- - - hrs
- -time
- 4. NOTIFY the NRC Operations Center of the event within 4 houn.
( ) RECORD additional information provided to the NRC on the NRC Data Sheet.
- - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name time SGS Rev. 00
ECG ATI17 Pg. 2 of 4 Initials
_ _ 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.
name time
- 8. NOTIFY Nuclear Licensing.
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 name
- -time - - hrs SGS Rev. 00
ECG ATf 17 Pg. 3 of 4 Initials
- 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
- 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 Manage~\- 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.
SGS Rev. 00
ECG ATT17 Pg. 4 of 4 REPORT OF SERIOUS INJURY/DEATH NUCLEAR BUSINESS UNIT EMPLOYEE EMPLOYEE INFORMATION NAME EMPLOYEE# _ _ _~- AGE HOME ADDRESS _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
HOME PHONE#
MARITAL STATUS - - - - - - - -
JOB TITLE _ _ _ _ _ _ _ _ _ _ _ _ _ _ LOCATION _ _ _ _ _ _ __
SOCIAL SECURITY# - - - - - - - - - - -
ACCIDENT/INJURY DESCRIPTION DATE OF ACCIDENT - - - - - - Til'v1E - - - - - AM/PM DID INJlJRIES RESULT IN DEATH 0 YES 0 NO EXTENT OF INJURIES DESCRIPTION OF ACCIDENT WHERE TAKEN AFTER ACCIDENT
~----------------~
SAFETY COORD. WORK# HO:ME# PAGER#
Cliff Knaub 2812 358-3074 478-5706 John Homer 2965 678-6308 342-5866 Andrew Caplinger 2828 478-5983 SGS Rev. 00
ECG ATT18 Pg. 1 of 4 ATTACHMENT 18 FOUR HOUR REPORT RADIOLOGICAL TRANSPORTATION ACCIDENT INSTRUCTIONS (SALEM 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. COMPLETE 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 SGS Rev. 00
ECG ATf 18 Pg. 2 of 4 Initials
- 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 (PIM) - Nuclear.
name time
- 7. NOTIFY the NRC Operations Center within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
( ) 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
- 10. 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.
SGS Rev. 00
ECG ATI18 Pg. 3 of 4 II. REPORTING Initials
- 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.
Rev. 00 SGS
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 D NO IF YES, Trucking Company N a m e : - - - - - - - - - - - - - - - - - - - - -
IF NO, What is the status of the driver?
~
LOCATION of Accident:
Roadwav/\lile .\1arkerilntersection Citv/Town State Number of Vehicles involved? 1-2-3-4 State or Local Police on the scene? D YES 0 NO Any personnel injuries? D YES 0 NO Any Fire involving truck contents? D YES 0 NO Trucking Company Dispatcher notified? D YES 0 NO Extent of damage to truck/trailer, container and contents:
ASK THE CALLER TO DO 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.
SGS Rev.00
ECG ATf 19 Pg. 1 of 3 ATTACHMENT 19 TWENTY-FOUR HOUR REPORT FITNESS FOR DUTY (FFD) PROGRAM EVENTS INSTRUCTIONS (SALEM 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 : - - - - - - - - - - - - ----
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 individua_l(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 boon of the time of discovery provided bytheMRO.
name notified at time
- - hrs
- 3. NOTIFY the NRC Resident Inspector.
name notified at
- -time hrs SGS Rev. 00
ECG ATf 19 Pg. 2 of 3 Initials
- 4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).
name 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 l\1RO Procedures AND ENSURE that this event is included in the 6 month FFD Report to the NRC.
Rev, 00 SGS
ECG ATI19 Pg. 3 of 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 MR.O IAW NC.NA-AP.ZZ-0042(Q).
- 3. Discovery Time: _ _ _ _ _ _ hrs o n - - - - - - - (date)
- 4. Work Dept. ofindividual(s): - - - - - - - - - - - - - - - - - - - - - -
- 5. Has plant safety been affected ? DYES D NO
- 6. Corrective actions taken or planned ? - - - - - - - - - - - - - - - - - - - -
- 7. Other pertinent infonnation: - - - - - - - - - - - - - - - - - - - - - -
SGS Rev. 00
ECG ATI20 Pg. 1 of 3 ATTACHMENT20 TWENTY-FOUR HOUR REPORT NRC REGIONAL OFFICE INSTRUCTIONS (SALEM 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 Initials
- 1. COrvIPLETE 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 Region 1 Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
( ) RECORD additional information provided to the NRC on the NRC Data Sheet.
name time
- 3. NOTIFY the NRC Resident Inspector.
name time
- 4. NOTIFY the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
( ) 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 SGS Rev. 00
ECG ATI20 Pg. 2 of 3 Initials
- 5. NOTIFY the Public Information Manager (PIM) - 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.
SGS Rev. 00
ECG ATI20 Pg. 3 of 3 Il. REPORTING
- 1. ENSURE that an Action Request (AR) is prepared.
SNSS FORWARD this attachment, along with the NRC Data Sheet and any supporting 2.
SNSS documentation, to the Operations Manager (OM).
REVIEW this ECG attachment, the AR and any other relevant information for correct 3.
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 .
Rev.DO SGS
ECG ATI21 Pg. 1 of 2 ATTACHMENT 21 REPORTABLE EVENT
- LAC/MEMORANDUM OF UNDERSTANDING (M.O.U.) \~
INSTRUCTIONS (SALEM 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. 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 SGS Rev. 00
ECG ATI21 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).
-'* 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.
1\.1NLR
- 8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.
SGS Rev. 00
ECG ATI22 Pg. 1 of 2 ATTACHMENT 22 TIS REQUIRED ENGINEERING EVALUATION ~\
INSTRUCTIONS (SALEM 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 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.
- 1. 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.
SGS 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 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.
SGS Rev.00
ECG ATI24 Pg. 1 of 10 ATTACHMENT24 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 SALEM SENIOR NUCLEAR SHIFT SUPERVISOR (SNSS) SHOULD:
A. NOTIFICATION OF HOPE CREEK SNSS
- 1. CONT ACT the Hope Creek SNSS and brief him/her on the specific SNSS circumstances as follows:
( ) a. SHARE information about the externally initiated event in progress.
( ) b. OBTA.IN 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 Salem SNSS is the EC, SNSS THEN IMMEDIATELY IMPLEMENT this attachment as EC.
- 3. IF an Offsite Toxic Gas Release is threatening Site Personnel (EAL 9.4.1.a),
EC THEN IMMEDIATELY IMPLEMENT appropriate Protective Actions for Site Personnel including initiation of Accountability and Evacuation per Section ill., Pg. 4, PRIOR TO notifications.
SGS Rev. 00
ECG ATf24 Pg. 2 of 10 II. EMERGENCY COO RD INATOR CECl LOG SHEET A. DECLARE A COMMON SITE UNUSUAL EVENT EC AT HOPE CREEK~ SALEM 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 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 (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 202S, OSC Activation and Operations AND
( ) IMPLEMENT Bomb Search Operations IAW Appendix 1.
( ) b. DIRECT the NCOs to check control boards for correct equipment lineups.
Rev. 00 SGS
ECG ATT24 Pg. 3 of 10 Initials D. EMERGENCY COORDINATOR DUTIES
( ) 1.. NOTIFY the Hope Creek SNSS. (NETS 5224; DID 3027, or 3059)
( ) 2. required, IMPLEMENT Accountability by referring to the Accountability Instructions in Section Ill.
- 3. WHEN provided by the CM2, EC THEN COMPLETE and APPROVE the NRC Data Sheet for transmittal 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 emergencJ:: 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
EC Recoyery (Pg. 6).
- 2. ENSURE appropriate reports are made IAW Section V., Reporting, of this SNSS attachment.
Rev. 00 SGS
ECG ATT24 Pg. 4 of 10 ill. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA A. IMPLEMENTATION OF ASSEMBLY AND ACCOUNTABILITY lnitialsffime I 1. IF NOT already done, EC THEN DIRECT the OSC Coordinator to activate the OSC IAW EPIP 202S, 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 Hope Creek SNSS to implement EPIP lOlH, Appendix 6, EC Accountability Instructions For An Unusual Event at Salem.
NOTE A.5 Steps A.5 thru A.9 may be delegated by the EC to any available CR Staff member.
I s: SOUND the Radiation Alert Alarm and make the following page announcement:
"Attention all personnel. Attention all personnel."
"Salem and Hope Creek 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, THEN 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)
SGS Rev. 00
ECG ATI24 Pg. 5 of 10 ill. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA (Cont'd) lnitialsffime I 8. WHEN 10 minutes have elapsed from Step 7, ANNOUNCE the following; (T+lO Min.)
11 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. 11 (Repeat)
I IO. 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 IAWEPIP 202S.
- 2. UPDATE Security as missing personnel are accounted for.
EC SGS Rev. 00
ECG ATT24 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 COMPLETE the "UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM," (Pg. 8), as follows:
( ) a. IF terminating WITHOUT Recovery, COMPLETE Part A.
( ) b. IF terminating WITH Recovery, COMPLETE Part B.
- 3. IF termination with Recovery is chosen, EC THEN DIRECT the EDO to assume the duties of the Recovery Manager including:
- a. EVALUATE 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 CMl.
( ) 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 Hope Creek SNSS.
EC
- 7. GO TO Section V., Reporting.
SNSS Rev. 00 SGS
ECG ATI24 Pg. 7 of 10 SNSS IV. TERMINATION (cont'd)
RECOVERY CHECKLIST FOR A COMMON 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 SALEM AND HOPE CREEK
( ) 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 VE with an EAL still exceeded is correct under the current circumstances?
( ) 2. SALEM EDO is prepared to assume the duties of Recovery Manager.
Time E. IF EITHER of the above are negative (unchecked),
THEN tennination 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. IV., Step 2 for Terminating WITH Recovery.
SGS Rev. 00
ECG ATI24 Pg. 8 of 10 Emergency Coordinator Date Time 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 SALEM NUCLEAR GENERATING STATION, UNIT THIS MESSAGE 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:
THIS IS - - - - - - - - . . . J COMMUNICATOR IN THE CONTROL ROOM AT THE SALEM NUCLEAR GENERATING STATION, UNIT THIS MESSAGE IS TO NOTIFY YOU THAT AS OF _____, ON - - - - -
time date THE CO:MMON SITE UNUSUAL EVENT AFFECTING BOTH HOPE CREEK AND SALEM HAS BEEN TERMINATED AND RECOVERY OPERATIONS IMPLEMENTED.
(SALEM DUTY EDO)
IS THE RECOVERY MANAGER LOCATED AT SALEM.
(EC Approval to transmit)
SGS Rev. 00
ECG ATI24 Pg. 9 of 10 V. REPORTING INSTRUCTIONS
- 1. This is a permanent document.
- 2. APPEND 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. PREPARE required reports.
LERC Report or LER N u m b e r - - - - - - -
- 6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.
Rev. 00 SGS
ECG ATI24 Pg. 10 of 10 INITIAL CONTACT MESSAGE FORM I. TIDS I S - - - - - - - - - - , COMMUNICATOR IN THE CONTROL ROOM (NAME)
AT THE SALEM NUCLEAR GENERATING STATION.
~------------------------------------------------
II.
0 TIDS IS NOTIFICATION OF A CO:MMON SITE UNUSUAL EVENT AFFECTING BOTH SALEM AND HOPE CREEK WIDCH WAS DECLARED AT ON _ _ _ _ _ _ _ __
(Time - 24 HR CLOCK) (DATE)
EAL# - - - - - - DESCRIPTION OF E V E N T : - - - - - - - - - - -
III.
NO RADIOLOGICAL RELEASE IS IN PROGRESS 33 FT. LEVEL WIND DIRECTION (From): WIND SPEED: - - - -
(From MET Computer) (DEGREES) (l\1PH)
IV. NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TIME EC Initials (Approval to Transmit IC:MF)
SGS Rev. 00
ECG AIT25 Pg. 1 of 3 ATTACHMENT 25 ONE HOUR REPORT (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 Initials
- 1. NOTIFY the Hope Creek SNSS (NETS x5224; DID 3027, 3059).
- 2. DETERMINE which Station SNSS will implement this attachment.
- 3. IF the Salem SNSS is responsible for this notification, THEN IMMEDIATELY CONTINUE with this attachment.
- 4. IF the Hope Creek SNSS will implement this attachment, THEN NO further actions are required by Salem except to lend assistance as necessary in restoring the lost equipment ot capabilities.
INSTRUCTIONS (SALEM 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: ---
Il. 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.
SGS Rev. 00
ECG ATf 25.
Pg. 2 of 3 Initials
- 2. NOTIFY the NRC Operations Center of the event within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
( ) RECORD additional information provided to the NRC on the NRC Data Sheet.
name time
- 3. NOTIFY the NRC Resident Inspector.
name time
- 4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).
name time
- 5. NOTIFY the Public Information Manager (PThf) - Nuclear.
name time
- 6. NOTIFY Nuclear Licensing.
name time
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 name time
- 8. NOTIFY External Affain.
notified at hrs name time SGS Rev. 00
ECG ATI25 Pg. 3 of 3 Initials
- 9. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.
ill. REPORTING
- 1. ENSURE that an Action Request (AR) is prepared.
SNSS AR# _ _ _ _ _ _ _ _ __
- 2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).
....). REVIEW this ECG attachment, the AR and any other relevant information for correct 0Y1 classification of event and corrective action taken.
-l FORWARD this attachment and any other supporting documentation to the LER 0\1 Coordinator (LERC).
- 5. PREP ARE required reports.
LERC Report or LER Number - - - - - - - -
_ _ 6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (yC\l1.R).
LERC
_ _ 7. ENSURE offsite (state and local) reporting requirements are met.
l\1NLR
_ _ 8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.
Rev. 00 SGS