ML20115B779
| ML20115B779 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 07/03/1996 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18102A240 | List: |
| References | |
| PROC-960703, NUDOCS 9607110037 | |
| Download: ML20115B779 (293) | |
Text
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- ECG SEdTIONS HOPE C' REEK ECG (FINAL DRAFT) ,
I 6 SECTIONS 1 - 9 g EMERGENCY ACTION LEVELS (EALs)
ECG SECTIONS NOTE: THESE 9 SECTIONS WILL REPLACE THE 18 SECTIONS CURRENTLY IN THE ECG.
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l ECG i T.O.C.
Pg.1 of 4 HOPE CREEK EVENT CLASSIFICATION GUIDE l TABLE OF CONTENTS / SIGNATURE PAGE August 15,1996 SECTION TITLE REV # PAGES DATE T.O.C. Table of Contents / Signature Page 0 4 08/15/96 l
i introduction / Usage 0 8 08/15/96 il Glossary of Acronyms 0 4 08/15/96 1.0 Fuel Clad Challenge 0 1 08/15/96 2.0 RCS Challenge 0 1 08/15/96 3.0 Fission Product Barriers 0 1 08/15/96 4.0 Miscellaneous (EC Discretion) 0 1 08/15/96 5.0 Failure to SCRAM 0 1 08/15/96 5
6.0 Radiological Releases / Occurrences 6.1 Gaseous Effluent Release 0 4 08/15/96 6.2 Liquid Effluent Release 0 1 08/15/96 6.3 in Plant Radiation Occurrences 0 3 08/15/96 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 0 2 08/15/96 7.2 Loss of DC Power Capabilities 0 1 08/15/96 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 0 1 08/15/96 8.2 Loss of Assessment Capability 0 2 08/15/96 8.3 Loss of Control Room Habitability 0 1 08/15/96 8.4 Technical Specifications 0 1 08/15/96 9.0 Hazards - Internal / External 9.1 Security Threats 0 1 08/15/96 1
9.2 Fire 0 1 08/15/96 9.3 Explosion 0 1 08/15/96 9.4 Toxic Gases 0 2 08/15/96 9.5 Seismic Event 0 1 08/15/96
( 9.6 High Winds 0 1 08/15/96 l 9.7 Flooding 0 1 08/15/96 9.8 Turbine Failure / Vehicle 0 1 08/15/96
-Missile impact HCGS Rev.O I
ECG T.O.C.
Pg. 2 of 4 I
v HOPE CREEK EVENT CLASSIFICATION GUIDE TABLE OF CONTENTS / SIGNATURE PAGE August 15,1996 SECTION TIT l,E REV # PAGES DATE 10.0 Reserved for future use 0 1 08/15/96 11.0 Reportable Action Levels (RAls) 11.1 Technical Specifications 0 2 08/15/96 11.2 Design Basis / 0 2 08/15/96 Unanalyzed Condition 11.3 Engineered Safety 0 1 08/15/96 Features (ESP) 11.4 Personnel Safety / Overexposure 0 2 08/15/96 11.5 Environmental 0 2 08/15/96 11.6 After-the-Fact 0 1 08/15/96
' 11.7 Security / Emergency Response 0 1 08/15/96 Capabilities 11.8 Public Interest 0 1 08/15/96 11.9 Accidental Criticality / 0 2 08/15/96 Special Nuclear Material (SNM)/
Rad Material Shipments - Releases i
i I
i i
lO HCGS Rev.O
. ECG Section i Pg 1 of 9 (v HOPE CREEK EVENT CLASSIFICATION GUIDE INTRODUCTION & USAGE Section l l
- l. PURPOSE OF THE EVENT CLASSIFICATION GUIDE (ECG)
! A. To provide a central reference document which enables the Sectior l 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.
l ll. EMERGENCY CLASSIFICATION DESCRIPTIONS f
(%) A. Emergency Classes: ;
- 1. The NRC/ 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 in ascending order are:
Unusual Event (UE) Least Severe Alert (A) l l Site Area Emergency (SAE) l l General Emergency (GE) Most Severe B. Unusual Event:
l 1. Plant events which are in progress or have occurred which
- indicate a potential degradation of the plant safety level.
!O HCGS REV . 00
. ECG Section i Pg7 of 9 l D
! .2. The lowest level of emergency at the plant, which can usually l be handled by the normal operating shift.
l 3. No releases of radioactive material requiring offsite response or l monitoring are expected unless further degradation of safety j systems occurs. Dose consequences would not exceed 5 t -
mrem TEDE.
l 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 1 and ENC are manned and may activate if needed for support. I
- 3. Any release of radioactive material is expected to be limited to j a small fraction of the EPA Protective Action Guideline
\ exposure levels. Dose consequences not to exceed 100 mrem TEDE.
l 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.
I 2. The entire emergency response organization is activated.
- 3. Any release of radioactive materialis not expected to exceed EPA Protective Action Guideline exposure levels beyond the
- p. ant boundary. Dose consequences not to exceed 1000 mrem TEDE.
iO
(
l HCGS REV . 00 i
. . - - _ ,.,7..
. ECG Section i 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.
- z. The entire emergency response organization is activated.
- 3. Release of radioactive material can be expected to exceed EPA Protective Action Guideline exposure levels of 1000 mrem TEDE offsite.
111. EVENT CLASSlFICATION GUIDE STRUCTURE A. The ECG is divided into 3 segments which are:
- 1. ECG Front Matter: Information which include the Table of Contents, Introduction & Usage, a Glossary of Acronyms, and Critical Functions Status Trees (CFSTs).
/^ 2. ECG Sections: Flow charts diagrams used to classify
\ events / conditions as emergencies or non-emergencies.
- 3. ECG Attachments: Implementing documents that provide direction for emergency and non-emergency classification, notification, reporting requirements, references and forms required to facilitate event communications.
B. ECG Sections Format With the exception of ECG Section 3, the ECG section flowcharts are comprised of the following segments:
- 1. Initiatina Condition (IC) A generic nuclear power plant condition or event where either the potential exist for a radiological emergency @ non-emergency reportable event QB such an emergency M non-emergency reportable event has occurred.
- 2. MODE: Refers to the Operating 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 protective system or O- operator actions, should be utilized when classifying events.
HCGS REV . 00
I l . ECG Section i l Pg-4 of 9 N
(d 3. EAL Number (EAL#): Each Emergency Action Level (EAL) has l been assigned a unique alpha numeric identifier called the
! EAL#. This EAL# is 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.
Sample EAL# = 9.4.1.a i
First Diait = 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.
Second Diait = identifies the subsection that the EAL is contained in. In the above example the Digit p 4 identifies that the EAL is found in Q subsection 4 of Section 9 thus 9.4, Toxic Gases.
Third Diait = The third digit identifies the emergency class associated with that particular EAL as follows:
If 3rd Digit is a 1 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 1hr report t
if 3rd Digit is a 2 THEN RAL is 4hr report if 3rd Digit is a 3 THEN RAL is 24hr report i QB GREATER i
HCGS REV . 00
. ECG Section i Pg-s of 9 i
\ Fourth Diait = If a fourth digit is used, it is always a lower case letter and delineate one of multiple events which lead to similar emergency or non-emergency class levels. In the above example the "a" delineate 1 of 3 EALs that result in an Unusual Event and fall under a common initiating condition.
A predetermined, site-specific, or observable threshold used to define a generic initiating condition that places the plant in a given emergency class or non-emergency report. An EAL/RAL can be an instrument reading, an equipment status indicator, a measurable parameter, a discrete observable event, analysis results, entry into specific EOPs, or another phenomenon which indicates the need for classification of an emergency or non-emergency.
- 5. Action Reauired: Identifies the specific emergency class or non-emergency report that is required and refers the user to a p specific ECG Attachment for implementation direction for the Q emergency or non-emergency event declared.
C. ECG Attachments:
The ECG attachments are comprised of various formats since the attachments are used for implementing directions, phone listings, and informational data.
IV. EVENT CLASSIFICATION GUIDE (ECG) USE A. The EALs described in the ECG are not allinclusive and will not identify each and every condition, parameter or event which could lead to an event classification. If the Emergency Coordinator, using his best judgment, determines an Initiating Condition has been satisfied but the specific EAL is in ouestion, he/she should promptly classify the event in accordance with the initiating Condition. If it is clear that the EAL has not been satisfied, then the Emergency Coordinator should not classify the event based on the initiating Condition (IC). If the EAL has been satisfied and the Initiating Condition is in auestion classify the event. In any event, if the plant
- conditions are equivalent to one of the four emergency classes as described in Section 11 above, that classification should be declared.
HCGS REV . 00 l
1
)
, ECG Section i Pg-s of 9 Assessment of an Emergency Condition should be completed in a timely manner which is considered to be within about 15 minutes of i j- recognition of an event. If an EAL specifies a duration (e.g. loss of l annunciations for > 15 min), then the assessment time runs l concurrently with the EAL duration time and is the same length. If an event is recognized or reported and the required duration 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. The ECG is not a stand alone document. At times, the ECG will l refer the user to other attachments or procedures for l accomplishment of specific evolutions such as: accountability, recovery, development of PARS, etc. They should be followed in a ,
step-by-step fashion.
I The ECG should be considered an " Implementing Procedure" and used in accordance with the requirements of a " Category 11" procedure as defined in NC.NA-AP.ZZ-0001(O). The ECG's j -
classification Sections allow for judgement and decision making as
- l. to whether or not an Emergency Action Level (EAL) is exceeded.
i i C. To use this ECG volume, follow this sequence:
NOTE:
Confirmation of actual plant conditions should be made by comparing redundant instrumentation, indications, and/or >
alarms.
- 1. Assess the event and/or plant conditions and determine'which ECG section(s) is most appropriate.
- 2. Refer to Section EAL/RAL Flowchart diagram (s), review and identify j the initiating condition (s) that are related to the event / condition that l has occurred or is ongoing.
E i (ECG Section 3 has its own unique usage instruction as part of the !
Fission Product Barrier Table 3.0) l HCGS REV . 00 I -
I
. ECG Section i Pg -7 of 9 U
NOTE:
The Emergency Coordinator should classify and declare an emergency before an Emergency Action Level (EAL) is exceeded if, in the EC's judgement, it is determined that the EAL will be exceeded.
- 3. Review and assess 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 Basis Document should be reviewed. Words contained in an EAL QR RAL that are bold face are either threshold values associated with that action level DR are words that are defined in the basis for that specific EAL/RAL. l
- 4. Identify and implement the referenced Attachment. j
[") 5. After classification and Attachment initiation, return to the ECG Section to review action levels that may result in {
escalation / deescalation of the emergency level. l D. Guidance for EMERGENCY /NON-EMERGENCY conditions discovered after-the-fact.
NOTE:
Plant emergency events that are in progress or that have occurred with ongoing consequences, effects, or corrective actions should not be considered "After-The-Fact" events and should therefore be classified and declared as an ongoing emergency event.
- 1. EMERGENCY CONDITIONS - if "After-The-Fact" (not on-going at the time of discovery) it is discovered that an event or condition occurred that exceeded an Emergency Action Level (EAL) but was not declared as an emergency, then an emergency declaration is
<~N NOT required. A non-emergency, One-Hour Report should be b initiated in accordance with ECG Section 11.6, After-The-Fact.
HCGS REV . 00 l
1
i ECG Section i Pg s of 9
!O I l
- 2. NON-EMERGENCY CONDITIONS - if After-The-Fact (regardless of l whether the event is on-going at the time of discovery) it is j discovered that an event or condition had occurred that should have resulted in the classification and implementation of a non-emergency I report (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,4 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 l report Attachment in the ECG should be implemented. ;
l l
E. Guidance concerning NRC COMMUNICATIONS during an emergency.
l 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 _QB 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.
f- 2. The NRC Data Sheet, along with the Initial Contact Message Form, l
{S) is the primary vehicle to ensure the NRC is kept informed. General Guidance on completing the event description portion of the NRC l
Data Sheet is provided in Attachment 5 of the ECG.
F. Guidance for VOLUNTARY / COURTESY REPORTING of Non-Emergency Events in accordance with NUREG 1022, Rev 1, voluntary reporting is encouraged.
PSE&G may make voluntary or courtesy NRC notification (11.10.1) 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 10CFR50.72 as defined in the ECG, then PSE&G should update the NRC with this information.
l l
LJ HCGS REV . 00
, ECG Section i i Pg s of 9 !
G. Guidance for Event Retraction if a ENS notification to the NRC was made as directed by the applicable ECG Attachment and it is later determined that the event or condition may not be reportable, the event may be retracted as follows:
- 1. Obtain both station General Manager's and Operations Manager's approval of any proposed retractions. -
- 2. Complete "page 1" of the NRC Data Sheet from the ECG Attachment which was implemented to make the original ;
notification. Event Description Section of NRC Data Sheet should '
explain the ationale for the retraction.
- 3. Contact the NRC OPERATIONS Center using the ENS and provide !
the "NRC Data Sheet" information. (Record on the "NRC Data !
Sheet" the name of the NRC Contact that received the retraction information).
- 4. 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. Guidance for Common Site Events:
Selected EALs (Unusual Event level only) and RALs have been designated l
as " Common Site" events. These events will be annotated with the words,
" Common Site" in the Action required portion of the EAL sections. These
, common site events need not be reported by both Salem and Hope Creek.
The reference ECG attachment will direct the SNSS 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 therefor will not be classified as common site events.
l O
l HCGS REV . 00
1 ECG Section ii Pg 1 of 5
( IIOPE CREEK GENERATING STATION l Emergency Action Levels and Reportable Action Levels Glossary of Acronyms Section li I
AC -
Alternating Current
. ADS -
Automatic Depressurization System APRM -
Average Power Range Monitor -
ARI -
Alternate Rod Insertion ATWS -
Anticipated Transient Without Scram BNE -
Bureau of Nuclear Engineering ( NJDEPE) -
CACS -
Containment Atmosphere Control System CEDE -
Committed Effective Dose Equivalent ;
CDE -
Committed Dose Equivalent CFR -
Code of Federal Regulations CIS -
Containment Isolation System CNTMT -
Containment CP -
Control Point j p CPM -
Counts Per Minute
() CR -
Control Room CREF -
Control Room Emergency Filter System CRIDS -
Control Room Integrated Display System CRD -
Control Rod Drive CSS -
Core Spray System DC -
Direct Current DAPA -
Drywell Atmosphere Post Accident (Radiation monitor) !
DEI -
Dose Equivalent Iodine j DEMA -
Delaware Emergency Management Agency DEPE -
NJ Department of Environmental Protection & Energy DID -
Direct Inward Dial (phone system)
EACS -
ESF Equipment Area Cooling System EAL -
Emergency Action Level EC -
Emergency Coordinator ECCS -
Emergency Core Cooling Systems EDG -
Emergency Diesel Generator EDO -
Emergency Duty Officer (g EMRAD -
Emergency Radio (NJ)
ENC -
Emergency News Center HCGS Rev. 00
ECG Section 11 Pg 2 of 5 ENS -
Emergency Notification System (NRC)
EOF -
Emergency Operations Facility EOP -
Emergency Operations Procedures EPA -
Environmental Protection Agency ERM -
Emergency Response Manager FC -
Euel Clad (Barrier) i FRVS -
Filtration, Recirculation, and Ventilation System FTS -
Federal Tele Communications System (NRC) -
GE -
General Electric ;
GE -
General Emergency GPM -
Gallons Per Minute HCLL -
Heat Capacity Level Limit HCGS -
Hope Creek Generating Station 1 HCTL -
Heat Capacity Temperature Limit HPCI -
High Pressure Coolant Injection l HTV -
Hardened Torus Vent HWCI -
Hydrogcq Water Chemical Injection l IC -
Initiating Condition
^ ICMF -
Initiai Contact Message Form IRM -
Intermediate Range Monitor l KV -
Kilovolt LAC -
Lower Alloways Creek LCO -
Limiting Condition for Operation LDE -
Lens Dose Equivalent LOCA -
Loss of Coolant Accident LPCI -
Low Pressure Coolant Injection LPZ -
Low Population Zone MEA -
Minimum Exclusion Area MET -
Meteorological MPH -
Miles Per Hour MRO -
Medical Review Officer MSIV -
Main Steam Isolation Valve MSIVSS -
Main Steam Isolation Valve Sealing System MSL -
Main Steam Line NAWAS -
National Attack Warning Alert System NETS -
Nuclear Emergency Telecommunications System NFPB -
Normal Full Power Background Os NJSP -
New Jersey State Police HCGS Rev. 00
ECG Section 11 Pg 3 of 5 Q
b NPV -
North Plant Vent NRC -
Nuclear Regulatory Commission NSSSS -
Nuclear Steam Supply Shutoff System CDCM -
Offsite Dose Calculation Manual OEM -
Office of Emergency Management (NJ)
OPCON -
Operating Condition OSC -
Operations Support Center PAG -
Protective Action Guidelines -
PC -
Primary Containment (Barrier)
PCIG -
Primary Containment Instrument Gas System PCIS -
Primary Containment Isolation System PSIG -
Pounds Square Inch Gauge RAL -
Reporting Action Level RC -
Reactor Coolant RCIC -
Reactor Core Isolation Cooling RCS -
Reactor Coolant System (Barrier)
RHR -
Residual Heat Removal (Containment Heat Removal) (
RMS -
Radiation Monitoring System n RPS -
Reactor Protection System V RPV -
Reactor Pressure Vessel RRCS -
Redundant Reactivity Control System SACS -
Safety Auxiliaries Cooling System SAE -
Site Area Emergency SBO -
Station Blackout SCP -
Security Contingency Procedure SDE -
Skin Dose Equivalent SDM -
Special Nuclear Material SNSS -
Senior Nuclear Shift Supervisor SPDS -
Safety Parameter Display System I SPV -
South Plant Vent SRM -
Source Range MNtor SRV -
Safety Relief Vane SSCL -
Station Status Checklist p SSWS -
Station Service Water System
(_) TAF -
Top of Active Fuel HCGS Rev. 00
I ECG e
Section 11 Pg 4 of 5 A NPV -
North Plant Vent NRC -
Nuclear Regulatory Commission [
NSSSS -
Nuclear Steam Supply Shutoff System ODCM -
Offsite Dose Calculation Manual
- OEM -
Office of Emergency Management (NJ)
OPCON -
Operating Condition OSC -
Operations Support Center PAG -
Protective Action Guidelines -
PC -
Primary Containment (Barrier)
PCIG -
Primary Containment Instrument Gas System PCIS -
Primary Containment Isolation System PSIG -
Pounds Square Inch Gauge RAL -
Reporting Action Level l RC -
Reactor Coolant RCIC -
Reactor Core Isolation Cooling RCS -
Reactor Coolant System (Barrier) i RHR -
Residual Heat Removal (Containment Heat Removal) ,
RMS -
Radiation Monitoring System :
p RPS -
\ RPV -
Reactor Pressure Vessel RRCS -
Redundant Reactivity Control System SACS -
Safety Auxiliaries Cooling System SAE -
Site Area Emergency SBO -
Station Blackout SCP -
Security Centingency Procedure SDE -
Skin Dese Equivalent SDM -
Shutdown Margin i SLC -
Special Nuclear Material SNSS -
Senior Nuclear Shift Supervisor SPDS -
Safety Parameter Display System SPV -
South Plant Vent SRM -
Source Range Monit r SRV -
Safety Relief Valve SSCL -
Station Status Checklist g SSWS -
Station Service Water System l V TAF -
Top of Active Fuel HCGS Rev. 00
ECG Section 11 Pg 5 of 5 TEDE -
Total Effective Dose Equivalent TSC -
Unusual Event UFSAR -
Updated Final Safety Analysis Report UHS -
United States Coast Guard VDC -
Volts Direct Current O
O HCGS Rev. 00
O O O 1.0 Fuel Clad Challenge "?!.Q" 1.1 RCS Activity "8' ' d '
Inititting Fuel Clad Degradation
'N Condition }
/
OPCON (1,2,3,4,5) ( 1,2,3,4 ) ( 1,2,3,4 ) ( 1,2,3 )
EAL# 1.1.1.a 1.1.1.b 1.1.1.c 1.1.2 i IF IF IF E l M i E l Reactor Coolant Valid OfTgas Pretreatment Valid Main Steam Line R I Sample Activity, Radiation Monitor Radiation Monitor G l > 4 pCi/gm (9RX621/ 9RX622) High liigh Alarm Condition E I Dose Equivalent fligh Alarm Condition (2 3 times Normal Full N l I-131 (3 2.2E+04 mrem /hr) Power Background) c !
V i ,
I I
A I c l l NN Il I ano
. determined is O
l > to be N I THEN Stuck Open i
THEN E l NOTE:
V Refer to Section 3.0, l . _ _ _ _ _ _ -
E i Fission Product Barrier Table g l prior to Event Classification I S l v v Actisn Referto Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT 1
2.0 RCS Challange "%E 2.1 RCS Leakage Initiating RCS Leakage Condition ,,
OPCON ( 1, 2, 3
) ( 1, 2, 3 ) ([ 1, 2,3 ) ( 1, 2, 3 )
EAL# 2.1.1.a 2.1.1.b 2.1.1.c 2.1.1.d i IF IF IF IF E l .
3g i Reactor Coolant System i Reactor Coolant System Reactor Coolant System "##"
- E Identified Leakage I
Pressure Boundary Leakage Unidentified Leakage "*'I ' "*E '
R > 25 epm '
l > 10 epm > 10 gpm Dual Seal Failure within G averaged over any ..
(Using 10 minute average) (Using 10 minute average) 0 minuta checognition l 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period N l c l Y :
! THEN A I C l T i I l 0 i NOTE:
N l Refer to Section 3.0,
~ ~ ~ ~ ~ - '
I Fission Product Barrier Table L l prior to Event Classification E
l V i e l -
i L i S l v
Action Refer to Attachment 1 Required UNUSUAL EVENT
~~
~
3.1.1 REACTOR WATER IEVEL 3.2.1 REACTOR WATER LEVEL POTTNTIAL l.OSS = 3 FTs POTENTIAL IDSS = 3 pts EAL # 3.1.1.a EAL # 3.1.1.b EAL # 3.2.1.a EAL # 3.2.1.b Reactor Water level Reactor WaterIzvel Reactor Water levei Reactor Water level REACHES - 161" (Top of CANNOT BE RESTORED 1EACHES -129" REACHES -161" (Top of Active Fuel) EXCLUDING AND MAINTAINED iKCLU_lXNQ intentional Active Fuel) EXCLUDING intentionallowenng of Reactor above 200"(Minimum Zero owenng of Reactor Water intentionallowering of Reacto Water level during an ATWS Injection RPV Water level) level during an ATWS Water level during an ATWS
-OR-3.2.2 RCS IEAK RATF1 DRYWEIL PRESSURE POTENTIALIDSS = 3 FTs EAL # 3.2.2.a EAL # 3.2.2.b
-OR- Unisolable RCS leak Rate Valid High Drywell Pressure
> 50 GPM E{SJDE _ Condition ( 2 1.68 psig)
Printary Contamment 3.1.2 DRYWELL ATMOSPHERE POST ACCIDENT (DAPA)
ANSTEC RADIATION LEVEL APERT 1 IRE POTFNTIAL IDSS = 0 FTs CARD ,,,,
Not Applicable EAL # 3.1.2 DAPA Radiation Monitor bg3o AV8{ table Of1 reading 2 5000 R/hr ertufe Cafd 3.2.3 RCS 11NE BREAK / CONTAINMENT BYPASS POTTNTIAL LOSS = 3 FTs bV EAL # 3.2.3.a EAL # 3.2.3.b Main Steam Line Break RCS Line Break OUTSIDE Primary Containment, Primary Containment, resulting in a Valid Isolation
-OR- resulting in an AUTOMATIC Signal for AEY one of the MSIV Isolation Signal following systems:
AND e NSSSS
- HPCI ;
ALL 4 Main Steam Lines have
LEAKAGE OUTSIDE Indication of CONTINUING l the Pnmary Contamment FIDW / LEAKAGE 3.1.3 RCS IODINE CONCENTRATION AFTER valve closure from the OUTSIDE the Pnmary Mam Contol Room has been Containment through the POTENTIAL I OSS = 0 FTs a".cmpted effected system AFTER ,
l Not Applicable EAL # 3.1.3 valve closure from the Reactor Coolant Sample Main Control Roorn has Activity > 300 pCi/gm Dose been attempted Equivalent I-131
-OR- -OR-3.1.4 EMERGENCY COORDINATOR JUDGEMENT 3.2.4 EMERGENCY COORDINATOR JUDGEMENT l EAL # 3.1.4 EAL # 3.2.4
/\EY condition, in the opinion of the EC. that indicates a AEY condition, in the opinion of the EC, that indicates a Potential loss (3 pts) or Loss (4 pts) of the Fuel Clad Barrier Potential loss (3 pts) or loss (4 pts) of the RCS Barrier FUEL CLAD BARRIER EAL# RCS BARRIER EAl#
POINT VALUE o / 3 / 4 (circle one) + POINT VALUE O / 3 / 4 (circle one)
"Pa$Oi e
3.3.1 REACTOR WAHR LEVEL '
POwNriu insS - i er -
FAL # 3.3.t Not Applicable FISSION PRODUCT A n y n i Reactor Water level CANNOT BE RESTORED AK 1 K ;
AND MAINTAINED !
above -200" (Minimum Zero injection RPV Water level)
_ APPLICABLE
-OR- OPERATIONAL :
3.3.2 DRYWEliPRESSUREl'"
POTENTIAL IDSS = 1 PT Q E IT O M A E ;
EAL # 3.3.2.a EAL # 3.3.2.b Supp Chamber press CANNOT Containment Failure as l EE MAINTAINED. below indicated by a rapid decrease in NOTE I 65psig Drywell pressure following an If the loss or Potential Loss is considered IMMINENT l
mcrease m pressure above (may occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), use judgement and classify as M P88g if the threshold is exceeded
~
Primary Containment 6ntainment is Vented by the . !
H2 concentration >4% and Emergency Operating Usage Instructions: l 02 concentration >$% Procedures (EOPs)
- 1. In the table to the left, review the Emergency Action Irvels of all columns and
-OR- identify which need further review.
3.3.3 DRYWELL ATMOSPHEPE POST ACCIDENT (DAPA)
RADIATION LEVEL 2. For each of the three barriers, determine the POTFNTIR IDSS = 1 PT EAL with the highest point value; enter that EAL # in the space provided at the bottom EAL # 3.3.3 Not Applicable of the column, and circle the corresponding D A Radi i e g 2o r int value. No more then one EAL should selected for each barrier.
-OR-3.3.4 RCS LINE BREAK / CONTAINMENT BYPASS 3. Use the tabulation section at the bottom of, the table and add the oint values circle /. for POTENTIAllOSS = 1 Pr the three barriers an enter the sum below:
EAL # 3.3.4.a EAL # 3.3.4.b RCS Line Break OUTSIDE isolation Signal for ANl 4. Classify based on the point value sum as Primary Containment, one of the following systems: follows; '
resultin in a Valid Isolation
$gna1
}o ,n h" f'h o NSSb(excluding Main hdSS e HPC; If the, su m is:
Classify as: Refer to
Attachment:
l 1
Steam Lines)
- RCIC l o HPCI 1, 2 UNUSUAL EVENT 1 o RCIC A_hll 3, 4 ALERT 2 Indication of 5,6,7,8 SITE AREA EMERGENCY 3 N_Q indication o CONTINUING CONTINUING 9,10 GENERAL EMERGENCY 4 FLOW / LEAKAGE OUTSIDE 1AKAGE the Primary Containment QO 01111 V /the Pnmary through the effected system Contamment through the CLASSIFICATION R valve closure from the effected system AFTER valve ain trol Room has been closure from the Main Control attempted Room has been 5. Implement the appropn.ate ECG attempted attachment per above chart.
-OR- 6. Continue to review the EALs on this Table i
1 3.3.5 EMERGENCY COORDINATOR JUDGEMENT for changes that could result in emergency escalation or deescalation. i PO SS-1 n EAL # 3.3.5
- N condition, in the opinion of the EC. that indicates a ntial less (1 pt) or loss (2 pts) of the Containment Barrier chop .
o; 1 CNTMT BARRIER EAL# Total (All 3 barriers)
POINT VALUE o / 1 / 2 (circle one) = Emergency Classification Points
i
[ "- []
(g) l D] L,)
4.0 Miscellaneous "f 5g 4.1 Emergency Coordinator Discretion
[ Other Conditions Exist h Other Conditions Exist Other Conditions Exist Which dXher Conditions Exist Which Which In the Judgement of the %1tich in the Judgement of the In the Judgement of the Emergency in the Judgement of the Emergency \
initiating Emergency Coordinator Warrant Emergency Coordinator Warrant Coordinator Warrant Declaration of Cond..ition Cocedinator Warrant Declaration off plaration oran Unusual Event Declaration of an Alert , a Site Area Emergency a General Emergency j OPCON (' All ) ( All ) ( All ) (_ All )
EAL# 4.1.1 4.1.2 4.1.3 4.1.4 l IF IF IF IF E i M
! Events are in progress or Events are in progress or Events are in progress or Events are in progress or E i have occurred which, in the have occurred which, in the have occurred which, in the have occurred which, in the
! judgement of the judgement of the judgement of the judgement of the l Emergency Coordinator, Emergency Coordinator, Emeigency Coordinator. Ernergency Coordinator, I indicate a indicate EITHER one of the indicate an indicate an l Potential Degradation following: Actual or likely major Actual or imminent C
y i of Plant Safety failure of plant functions subtantial core i , Plant safety systems needed for protection of degradation with the i THEN (more than one) are or the public A i potential for loss of may be degraded C l containment i . ANY Plant Vital T I
, i Structure is degraded or I THEN potentially degraded O I THEN AND L l E : Increased monitoring of I
y Safety Functions E is warranted l ,
L I THEN s l y v v v Action Refer to Attachment 1 Refer to Attachment 2 Refer to Attachment 3 Refer to Attachment 4 Required UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY
q p n b d U I!CCS ECG 5.0 Failure to Scram " " '
am o.
5.1 ATWS Fadure of ec Rmtor Nection Fadure of ec Rator Ntection System (m) ^
Failure of the Reactor Protection System (RPS)
System (RPS) to Successfully Complete to Successfully Complete a Reactor Scram Initiatin8 to Successfully Complete a Reactor Scram (Automatic and Manual) and there is mdication of an
. a Reactor Scram (Automatic and Manual)
Condition (Automatic or Manual) and Reactor Power is above 4% Extreme Challenge to the Ability to Cool the Core /
OPCON ([ 1,2 }) ( 1,2 ) ( 1, 2
) ( 1,2 })
EAL# 5.1.2.a 5.1.2.b 5.1.3 5.1.4 l IF IF E i M i An Automatic ANY Manually E l Reactor Scram initiated R 1 Condition exists Reactor Scram G I }
AND E I from the N ! An Automatic Control Room C Reactor Scram IS NOT l
Y (RPS) IS NOT successful successful l
A i ALL Reactor Scram attempts l
C l AND from the Control Room T I (RPS aILd ARI) DID NOT REDUCE and MAINTAIN I
O !
"" Reactor Power to < 4% -
"" " " * "E N t
- Reactor Water Level CANNOT BE l
AND MMMD NIW l
j y
- The combination of Suppression Pool 1 THEN Temperature and RPV Pressure y
I v CANNOT BE MAINTAINED E ,
I belowthe HCTL Curve 3
S 8 y _
y {THEN Action Refer to Attachment 2 Refer to Attachment 3 Refer to Attachment 4 Required ALERT SITE AREA EMERGENCY GENERAL EM ERGENCY
O O 6.0 Radiological Releases / Occurrences 4 "U S" 6.1 Gaseous Effluent Release Initiating 'Any Unplanned Release of Gascous Radioactivity t Any Unplanned Release of Gaseous Radioactivity to Condition r h""ij" fica for inu es ng 8' the Emironment that Exceeds 2 times the 10CFR20. Appendix B limits for 60 minutes or longer OPCON ( All ) ( All ) ( All ) ( All )
EAL# 6.1.1.a 6.1.1.b 6.1.1.c 6.1.1.d Field i cose IF Measured IF samme IF Alarm IF Assessment D se Rate Analysis Indicahons E l M l Dose Assessment Dose Rate measured Total gaseous ellluent release Valid High Alarm received from ANY one of the E indicates EITTER following Plant Ellluent RMS Channels:
l at the Protected sample analysis for ANY one R I ne of the following Area Boundary of the following indicates a e FRVS Noble Gas (Grid 1/3; 9RX680) l at the MEA or Wond concentration of:
G or beyond EXCEEDS as calculated on .05 mrem /hr
- NPV Noble Gas (Grid 1/3; 9RX590)
E l . FRVS:
l the SSCL: above normal N 1 background 2 1.13E-03 pCi/cc Total Noble Gas
- SPV Noble Gas (Grid 1/3; 9RX580)
C I
- TEDE 4-DaI dos
- 2 2.71E-07 pCi/cc I-131 I
I, 2 2.0E-01 mrem
- HTV Noble Gas (Grid 1/3; 9RX518)
I ' NPV:
- Thyroid-CDE Dose l 2.43E-04 pCi/cc Total Noble Gas AND A I 2 6.8E-01 mrem 2 5.81E-08 pCi/cc I-131 C based on Plant Vent l .
.SPV Total Plant Vent release rate EXCEEDS one of the fru nt sample analysis T I and not on a default 2 2.27E-05 pCi/cc Total Nobic Gas I
I I Noble Gas to lodine 2 5.44E-09 pCi/cc I-131
- 4.80E+03 pCi/m Total Noble Gas O I Ratio N
L i AND E Dose Assessment results NOT available l
V I I E I AND ,
L 3 Release is ongoing for > 60 minutes I
S I y
THEN Action Refer to Attachment 1 Required UNUSUAL EVENT
o q
,d pJ V 6.0 Radiological Releases / Occurrences "S"'5 6.1 Gaseous Effluent Release ,
Initiating ' Any Unplanned Release of Gascous Radioactivity i
"" " 8 UnlaM Rh of Gm Nimis m Condition "j'hi g I in s for in or the Environment that Exceeds 2 times the 10CFR20, Appendix B limits for 60 minutes or longe OPCON ( All ) ( All ) ( All ) ( All )
EAL# 6.1.1.a 6.1.1.b 6.1.1.c 6.1.1.d Field 1 oose IF uessured IF sam
- IF Alarm IF g Assessment Dose Rate Analysis k ni .J;vii.
l M Dose Assessment Dose Rate measured Gascous emuent release Valid High Alarm received from ANY one of the l indicates EITHER following Plant Emuent RMS Channels:
E , at the Protected sampic analysis for ANY one R I ne of the following Area Boundary of the following indicates a . FRYS Nobic Gas (Grid 1/3; 9RX680)
I at the MEA or Mond G or bevond EXCEEDS concentration of:
as calculated on .US mrem /hr
- NPV Nobic Gas (Grid 1/3; 9RX590)
E l the SSCL: . FRVS:
l above normal N l background 2 1.13E413 pCi/cc Total Nohic Gas
- SPV Nobic Gas (Grid 1/3; 9RX580)
- IITV Nobic Gas (Grid 1/3; 9RX518)
I, 2 2.0E411 mrem I
- NPV:
' Thyroid-CDE Dose l 2 2.43E-04 pCi/cc Total Noble Gas AND A I 2 6.NE4)I mrem 2 5.81E-08 pCi/cc l-131 -
C based on Plant Vent l . . p. Total Plant Vent release rate EXCEEDS
' " # "E "
d T or . def t 2 2.27E415 pCi/cc Total Nobic Gas I I Noble Gas to lodinc 2 5.44E-09 pCi/cc l-131
- 4.80E+03 pCi/sec Total Noble Gn O I Ratio
N l -_.
I L I AND E Dose Assessment results NOT available l
V I __J E ARO .
l ,
L Release is ongoit:g for > 60 minutes S l
' THEN y
Action Refer to Attachment 1 Required UNUSUAL EVENT
O O O 6.0 Radiological Releases / Occurrences "$1" 6.1 Gascous Effluent Release
. '^ny Unplanned Release of Gascous Radioactivity t^ '
3 Any Unplanned Release of Gascous Radioactivity to the Er..honment that Ex
( Enuronment that Exceeds 200 Times the Radiological 200 times the 10CFR20, Appendix B limits for 30 minutes or longer Con , ion Technical specincations for 15 minutes or longer , x _ _ _ ,
OPCON (~ All ) ( All [) ( All ) ( All ')
EAL# 6.1.2.a 6.1.2.b 6.1.2.c 6.1.2.d Field 1 oose IF Measwed IF samme IF Aiann IF Assessnent D se Rate Ana p in&aks E l M I Dose Assessment Dose Rate measured Total gaseous emuent release Valid I!igh Alarm received from ANY one of the E indicates EITilER at the Protected f 11 wing Plant Emuent RMS Channels:
l samp'c analysis for ANY one R I ne f the following Arca Boundary of the following indicates a e FRVS Nohic Gas (Grid 1/3; 9RX680)
I t the MEA or beyond or beyond EXCEEDS concentration of:
G I as calculated on 5 mrem /hr
- NPV Nohic Gas (Grid 1/3; 9RX590)
E the ssCL:
- FRVs:
l N i > 1.31E-01 pCi/cc Total Noble Gas
T, I
-> 2.0E+01 mHem
- IITV Nnble Gas (Grid 1/3; 9RX518)
I
- NPV: f I
- Thyroid-CDE Dose 2 2.43E-02 pCi/cc Total Noble Gas AND A i 2 6.NE+01 mrem 2 5.81EM pCi/cc I-131 - - ---
l C I based on Plant Vent Total Plant Vent release rate EXCEEDS i I
emuent sampic analysis
- sPV: EITIIERone of the following limits:
l l and NOTon a default 2 2.27E-03 pCi/cc Total Noble Gas I I Nobic Gas to lodinc 2 5.44E-07 pCi/cc I-131
- 4.80E+05 pCi/sec Total Nohic Gas O I Ratio 1.15E+02 pCi/sec l-131 (NPV & SPV ONLY)
N i l I I
g AND
~
E I AND Dose Assessment results NOT available.
Y I _ AND I
E Release is ongoing for > 15 minutes -
I -
b Release is ongoing for > 30 minntbs i THEN i 3 1 'THEN h
I Action Refer to Attachment 2 Required ALERT
(]
s 6.0 Radici:gOReleas:s/ V Occurrences O V ucostco 6.1 G:scous Efflu:nt Reins 2 no.og Page 3 of 4 x
Initiating [
r Boundary Dose Resulting from an Actual or Imminent Release of Gascous Radioactivity Exceeds 100 mrem Total Effectin Dose Equivalent (TEDE) j Condition (w or 500 mrem Thyroid CDE Dose for the actual or projected duration of the release
/
OPCON ( All ) ( All ) ( All ) ( All )
EAL# 6.1.3.a 6.1.3.b 6.1.3.c 6.1.3.d Field
' oose IF ueasured IF Field Sunrey IF Alarm IF
^55'5$ " "t DostR W AnaWs inddions E l M I Dose Assessment Dosc Rate measured Valid High Alarm received from ANY one of the
, _ Analysis of field E : indicates EITHER at the Protected survey samples at the following Plant Emuent RMS Channels:
R i one of the following Area Boundan- Protected Area Boundary I at the MEA or beyond or beyond EXCE5DS .
e FRVS Nobic Gas (Grid 1/3; 9RX680) ng:
- NPV Noble Gas (Grid 1/3; 9RX590)
- fg one th f I AND 2 4.36E+02 CCPM e SPV Noble Gas (Grid 1/3; 9RX580) l
- TEDE 4-Day Dose y 21.0E+02 mrem
- 2 3.85E-07 pCi/cc I-131 e IITV Noble Gas (Grid 1/3; 9RX518) l I Release is expected to e Thyroid-CDE Dose A > 5.0E+02 mrem AND C 8 bah on Plant Vent emuent sampic analysis and NOT on T I Total Plant Vent release rate EXCEEDS i a default Noble Gas to I 4.21E+07 pCi/sec Total Nohic Gas I lodine Ratio O I N l AND L Dose Assessment results NOT available.
l E AND l
V I Release is ongoing for > 30 minutes I
E -
1 l L i j S THEN l
l v
Action Refer to Attachment 3 l
l Requirett SITE AREA EMERGENCY
O O O 6.0 Radiological Releases / Occurrences "%"I!"
6.1 Gaseous Effluent Release Initiating Boundary Dose Resulting from an Actual or Imminent Reicase of Gascous Radioactivity Execeds 1000 mrem Total Effective Dose Equivalent (TEDE)
Condition or 5000 mrem Thyroid CDE Dose for the actual or projected duration of the release OPCON (} All ) ( All } ( All ) ( All )
EAL# 6.1.4.b 6.1.4.c 6.1.4.d 6.1 4.a- Field I Dose IF Measured IF rieid survey IF Alarm IF
^5 5'55'"'"' dos'Bo* ^"*N5'S '"dic8"o"5 E l M Dose Assessment Dose Rate measured Valid High Alarm received from ANY one of the l Analvsis of f~icid E : indicates EITilER at the Protected uncy hmples at the f0lIO"ing Plant Emuent RMS Channels:
R i one of the following Arca Boundary Protected Arca Boundary . FRVS Nohic Gas (Grid 1/3; 9RX680)
I t the MEA or beyond or beyond EXCEEDS ,
G I in ts NM as calculated on 1000 mrem /hr E ne f the following:
- NPV Noble Gas (Grid 1/3; 9RX590) l the SSCL:
AND
- 24.36E+03 CCPM e SPV Nobic Gas (Grid 1/3; 9RX580)
I
- TEDE 4-Day Dose y 21.0EH13 mrem
- 2 3.M5E-06 pCi/cc I-131 e llTV Nobic Gas (Grid 1/3; 9RX5 tM) l I Release is expected to
- Thyroid-CDE Dose ~
A 1 > 5.0E+03 mrem AND I bab on Plant Vent emuent C ~~~
l S mpic Analysis and NOT T Total Plant Vent release rate EXCEEDS
- on a default Noble Gas to I 4.21EH)M pCi/sec Total Nohic Gas I lodine Ratio O I _ . _ _ _
i AND I
L Dose Assessment results NOT available.
i _
E I AND V i Release is ongoing for > 30 minutes I
E l l L i I
S '
y THEN Action Refer to Attachment 4 Required GENERAL EMERGENCY
O O O 6.0 Radiological Releases / Occurrences "S"5" 6.3 In-Plant Radiation Occurrences
' Release of Radioactive Material or increases in Radiation Levels within the ]
initiating UnPl anned Increase in Plant Radiation facility that impedes operation of systems required to maintain Condition safe operations or to establish or maintain Cold Shutdown j
}
OPCON ( All ) ( All ) ( All )
EAL# 6.3.1.a 6.3.2.a 6.3.2.b 1 IF IF E l ,
M Unplanned increase in Unplanned Dose Rates l
E i radiation levels inside the > 15 mrem /hr R Protected Area in EITHER one of the following:
l G I 21000 times normal
- Main Control Room E as indicated by EITHER i
N I one of the following:
- Security Central Alarm Station (CAS)
C I y
I
- Permanent or portable i Area Radiation Monitors I
A I
- General Area Radiological Survey C l Unplanned Dose Rates T I 2 2000 mrem /hr I i AND in ANY area of the plant O I which requires ACCESS to N i > maintain plant safety i THEN functions (EXCLUDING the L i , Main Control Room and CAS)
E I V l E
l L i THEN S l v v Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
O O O 6.0 Radiological Releases / Occurrences "TJ.5f' 6.3 In-Plant Radiation Occurrences nMadng
. Unplanned Increase in Plant Radiation Condition OPCON (5 ) ( All )
EAL# 6.3.1.b 6.3.1.c i IF IF E l M i Uncontrolled water level Uncontrolled water level E l decrease in the decrease ln the R Reactor Cavity as Spent Fuel Pool G
l indicated by EITHER as indicated by E one of the following: Valid Fuel Pool Low l
- Visual Observation Level Alarm Condition C l Y
- Reactor Water Level l Shutdown Range Level AND
^ l Indicator IBBLI-R605 C l Visual Observation T i l 1 1 l
o i N l THEN E I V l E I .
L i S I v
Action Refer to Attachment 1 Required UNUSUAL EVENT i
U,~
6.0 Radi:Ingib, Rcirs:s/Occurrcncas O nms r-Rev.00 6.3 In-Plant Radiation Occurrences e.g3.r3 Initiating '
Condition Major Damage to Irradiated Ft.cl
[ Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel CW ( All )
( All ) ( All )
EAL# 6.3.2.c 6.3.2.d 6.3.2.e IF lp yp E l i '
M i Unplanned increase E I Major Damage to on ANY one of the Visual observation of l Irradiated Fuel R following Irradiated Fuel uncovered has occurred G l Area Rad Monitors or E I AND I by general area rad survey N
C Valid High Alarm > 20 e /hr:
l -
I received from
- Spent Fuel Storage Pool l ANY one of the A I following Area (9RX707)
C RMS channels:
- New Fuel Criticality i
- Refuel Floor Exhaust Storage Channel A g
' (9RX612)
O Channel A(9RX627)
N
- Refuel Floor Exhaust
- New Fuel Criticality l
Storage Channel B l Channel B (9RX628)
( 613) l
- Refuel Floor Exhaust E
y Channel C (9RX629) l E I L s l
S I
' THEN v
Action Refer to Attachment 2 Required ALERT
O O O liCGS LCO 7.0 Electr.ical Power ne.m 7.1 Loss of AC Power Capabilitics
' tess of All OITsite Power 'AC power capability to Vital Buses reduced to a Singic Power' Loss of All Offsite Power and All Onsite AC -
I n. . .itiating to vital Buses for greater Source for greater than 15 minutes such that any additional single Pourt to 4.16 KV Vital Buses during either in Cold )
Condition than is minutes j railure would result in a complete loss of all 4.16 KV Vital Buses Shutdown or Refueling for greater than 15 minutes -
OPCON ([ All ) ( 1, 2, 3 ) (4, 5, Defueled)
EAL# 7.1.1 7.1.2.a 7.1.2.b i IF lp ip E
M Unplanned Loss of Loss of 4.16 KV Vital Bus Power Sources ALL 4.16 KV Vital Buses E Power from Station are deenergized I (Offsite and Onsite) which results in the :
R Service Transformers availability of ONLY l (OPCON 4, 5, and G I I AX501 AND 1BX501 one 4.16 KV Vital Bus Power Sc,urce E to ALL (Offsite or Onsite)
U"'d "
l N I 4.16 KV Vital Buses C 8 Y I AND I
I A i > 15 minutes C have elapsed AND l
T I I THEN > 15 minutes have elapsed l
O I
' THEN N g I
L l E I v l E ,
l ,
L I s l v v Action Refer to Attachment 1 Refer to Attachment 2 Requirect UNUSUAL EVENT ALERT
n n (~
-V U s misrui <
7.0 Electr.ical Power n.= !
7.1 Loss of AC Power Capabilitics initiating ' i.~oss of AliUITsite Power and All Onsite AD Power ti' Prolonged Loss of All OITsite and All Vit i AC Buses during either Po .cr Operation, Condition (Qtartup or llot Shutdown for greatr than 15 minutes,. . _
Onsite AC Power to All Vital AC Buses OPCON (' 1, 2, 3 } ( 1, 2, 3 ') I 1, 2, 3 )
EAL# 7.1.3 7.1.4.a 7.1.4.b ;
E l --
M i ALL 4.16 KV Vital Huses E l are deenergized R I i
G 1 (OPCON 1. 2. 3 ONLY) l E I I
N i y t C l THEN v i i T AND AND I _ _ . _ _ _ _ _ _
I AN Itestoration of Power to C ** ""I T l [~ ~ at least one .
> 15 un.nutes have elapsed .* * ' " " *
, 4.16 KV Vitial Bus i i has occurred or is Imnu.nent i within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is. NOT likely l O I THEN N i THEN i
THEN L l E I v l I
E I I
L i I
S t v _ _ _ _
Action Refer to Attachment 3 Refer to Attachment 4 Required SITE AREA EMERGENCY GENERAL EMERGENCY t
O O O 7.0 Electrical Power ""lk" 7.2 Loss of DC Power Capabilities Initiating Unplanned Loss of All Vital 125 VDC Power during either ' Unplanned Loss of All Vital 125 VDC Power during either Power Condition Cold Shutdown or Refueling Mode for greater than 15 minutes Operation, Startup or Hot Shutdown for greater than 15 Minutes /
OPCON ( 4,5, Defueled ) ( I, 2, 3 )
EAL# 7.2.1 7.2.3 i IF IF E l M Unplanned degraded voltage condition for Unplanned degraded voltage condition for l
E I ALL Vital 125 VDC Buses, ALL Vital 125 VDC Buses, R such that voltage is < 108 VDC such that voltage is < 108 VDC l
G 3 (OPCON 4,5. and (OPCON 1,2,3 ONLY)
N l Defueled ONLY)
I AND AND
\ i I
> 15 minutes have elapsed > 15 minutes have elapsed l
C THEN THEN T l i l O I I
N i I
L i E
l V i E
l ,
L i S l i Action Refer to Attachment 1 l Refer to Attachment 3 Required UNUSUAL EVENT l SITE AREA EMERGENCY
__~ , .- _
O O O 8.0 System Malfunctions "ZET' 8.1 Loss ofIIeat Removal Capabililty
[ *
- Y **
- initiating [ Inability to Maintain the Plant inC idShutd ""
Complete Loss of Functions Needed to Achieve Cold Shutdown Conditions /
Condition \
" iI"t" h"NS"v'$i OPCON (4,5 ) (4,5 ) ( I,2,3 )
EAL# 8.1.2 8.1.3.a 8.1.3.b i IF IF IF E l M i Unplanned, Complete Loss of Reactor Water Level Loss of Main Condenser capabilities, as evidenced by E i ALL Technical Specification required REACHES -161" an inability to remove Decay IIeat from the Reactor R systems available to provide (Top of Active Fuel) l G i Decay lleat Removal functions AND E I N l i Loss f Torus capabilities as evidenced by EITIIER C i AND AND y one of the following:
3 An I RCS Temperature Entry into an Unsafe region of ANY of the A , UNCONTROLLED i has increased t I"II "I"8 C"'VCS C temperature
> 200 F Heat Capacity Temperature Limit (IICTL) Carve T l increase
- i (Excludmg a lleat Capacity Level Limit (HCLL) Curve is RAPIDLY monientag Pressure Suppression Pressure (PSP) Curve O I . approaching i increase >200 F SRV Tailpipe Level Limit Curve N 200 F I
with a (with NO heat g I heat removal ,,e,,y,g,c,;,, . Insufficient SRV capacity to reduce RPV pressure E function restored) l restored)
V l E
l , .
l THEN S ,
v v Action Refer to Attachment 2 Refer to Attachment 3 Required ALERT SITE AREA EMERGENCY
. p 11cos Eco Cs 8.0 Syst Mcifuncti:ns d Ein 8.2 Loss of Assessment Capabililty .
Initiating Unplanned Loss of All Onsite or OfTsite Communications Capabilities OPCON ( All ) ( All )
EAL # 8.2.1.a 8.2.1.b
, IF IF 1
E I I Unplanned loss of Unplanned loss of gy 0"3 0 E Ii communications communications R I as evidenced by the loss of as evidenced by the loss of Gl ALL of the following systems:
ALL of the following systems:
E i Nl
- Station Page (Gaitronics)
- Direct Inward Dial (DID)
C 1 Station Radio
- Nuclear Emergency Y I
- Direct inward Dial (DID) Telephone (NETS)
I
- Essex Phone (Centrex)
A i . NAWAS Cl T
- EMRAD i
g i
- FTS 2000 ol N II I
' l E i V I E I THEN t l i S I y Actirn Refer to Attachment 1 Required UNUSUAL EVENT
_ _ _ - _ _ _ _ _ _ _ . _ . _ _ _._m- _ ____ _ - _ _ _ _ _ , __ m _._ _ -r -
s- -. e4 manas-a
^
(d 8a0 Syste alfuncti:ns
(' ncostco r Lr2 8.2 Loss of Assessment Capabiliity .
[nitjaling Unplanned less of Mod or AI! Unplanned less of Most or All Control Room Inability to Monitor a Significant Transient in
.. Ar 1unciation or Indication in the Control Annunciators and a Signirant Transient is in Progress or Compensitory l Condition Room ror creater nan u minutes indicat= = Unmiwe P"S'*SS
,/
OPCON ( 1,2,3 ) (1,2,3) (1,2,3) (1,2,3)
EAL# 8.2.1.c 8.2.2.a 8.2.2.b 8.2.3 IF lp El Ml E I Unplanned Icss of> 75% Loss of> 75% of Main Control Room Overhead Annunciators of Main Control Room R Overhead Annunciators MD E I > A significant transient ** is in progress AND AND N I THEN BOTH of the AND C i y l v A significant following: BOTH of the following:
- 3 AND .
- CRIDS
- CRIDS g is in progress l *SPDS *SPDS 15 minutes are NOT AVAILABLE Cl have elapsed are NOT T I AVAILABLE AND I since the loss of ;
g O I Ollas '
THEN Main Control Room Indications are NOT I available to monitor ANY one of the I y following:
I
- RCS S atus L I 15 minutes have
- Reactivity Control E II THEN elapsed since the loss .ECCS ofOllAs .
E i,
- Containment Parameters L I THEN THEN s I y y I
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 SCRAM, LOAD REJECTION >25% POWER, ECCS INJECTION, THERMAL POWER OSCILLATION >10% .
O O O 8.0 System Malfuncti::ns ""*l,"
8.3 Loss of Control Room Ilabitibililty Initiating Main Control Room Evacuation ) Main Control Room Evacuation has been Initiated Condition has been Initiated ,/ and Plant Control cannot be established OPCON ( All ) ( All )
EAL# 8.3.2 8.3.3
' IF E l
! Main Control Room Evacuation j has been initiated G l E I I AND C THEN i _
Y I l Control of the plant CANNOT be established A I from outside the Main Control Room C l within 15 minutes T i THEN N l l
L !
E l
Y I E
l ,
L i S l v v Action Refer to Attachment 2 Refer to Attachment 3 Required ALERT SITE AREA EMERGENCY
O O O 8.0 System Malfunctions "?L" 8.4 Technical Specifications Initiating Inability to Reach Required Operational Condition Condition within Technical Specification Limits OPCON ([ 1, 2, 3
)
EAL# 8.4.1
' IF E l -
'I l Plant is NOT brought to the REOUIRED I Operational Condition within the i Technical Specification required time limit N :
C THEN l
Y :
i i
A I C l T I I l o i N l 1
t l E I v l E
l ,
L i I
S Action Refer to Attachment 1 Regiaired UNUSUAL EVENT
O O O 9.0 Hazards -Internal / External "T*.4' "
9.1 Security Threats "*'"'
initiating Confirmed Security Event Which ecu% wnt Resuhing Security Event in a Security Event in a in Loss of Ability to Reach a Indicates a Potential Degradation in the pgp g pj yg Level of Safety of the Plant Maintain Cold Shutdown ,
OPCON ( All ) ( All )' ( All ) ( All )
EAL# 9.1.1 9.1.2 9.1.3 9.1.4
' IF IF IF IF E l M Confirmed security threat directed Confirmed hostile intrusion or l Confirmed hostile ,,
Security cwnt resultmg m the E toward the station as evidenced by M intrusion or malicious acts malicious acts in Plant Vital Arcas R I one of the following: as esidenced by : actualloss of physical control as evidenced by ANY
- * "E G l
- Credible threat of malicious acts or one of the following:
- Discovery of an intruder (s),
I destructive device within the Protected armed and violent, within the
- Main Control Room E
- Discovery of an Area, resulting in SCP-5 implementation Plant Vital Area, resulting in
- Remote Shutdown Panel N l intruder (s), armed and i
- Credible intrusion or assault threat violent, within the SCP4 implementation C
y to the Protected Arca, resulting in Protected Area,
- Malicious acts or destructive l resulting in i SCP-5 implementation device discovered in a Plant Vital THEN I
- Attempted intrusion or assault to the SCP4 implementation Arca, resulting in SCP-10
- Hostage held implementation l Protected Area, resulting in SCP-7 or I SCP-11 implementation on-site in a non-vital T I area, resuh mg-in l
- Malicious acts attempted or discovered THEN
' I within the Protected Area, resulting in SCP-8 implementation O SCP-10 implementation l
I
- Ilostage/ Extortion situation that l threatens normal plant operations, THEN L i resulting in SCP-8 implementation E
- Destructive Desice discovered within l
V l the Protected Area, resulting in E I SCP-10 implementation ,
I I L I I THEN S
v v v t_
Action Refer to Attachment 24 Refer to Attachment 2 Refer to Attachment 3 Refer to Attachment 4 Required UNUSUAL EVENT (Common Site) ALERT SITE AREA EMERGENCY GENERAL EMERGENCY
- _ = _ _ _ _.
(^')
\_/
(3
%s (V3 9.0 Hazards -Internal / External "2'll 9.2 Fire Initiating Fire within the Protected Area Boundary Fire Affecting the Operability of Plant Safety Systems Condition Not Extinguished within 15 minutes of Detection Required to Establish or Maintain Safe Shutdown OPCON ( All ) ( All )
EAL# 9.2.1 9.2.1 9.2.2 i IF IF IF E l M i Valid Fire Alarm is received Report of a fire from Fire within ANY one of the following Plant Vital Structures:
I in the Main Control Room personnel at the scene E I
- Reactor Building R I I I G ! l AND e Control / Aux Building E Fire is within ANY one of the following Plant Structures
- l (EXCLUDING small fiers that have _N_Q potential to alTect Swm N i Safety Systems or Protected Area Permanent Plant Structures) .
C Senice/ Rad Waste Building y l
- Reactor Building l AND 1
- Turbine Building A I The Fire is of a magnitude that it SPECIFICALLY C
- Control / Aux Building l results in Damage to ANY one of the following:
T
- Senice Water Intake Structure I
l
- TWO OR MORE subsystems of a Safety System i
O I
- Scnicc/ Rad Waste Building
- MORE THAN ONE Safety System l
- Low Level Radwaste Interim Storage Facility
- Any Plant Vital Structure which renders the structure incapable of l
I Performing its Design Function L l l AND E I Fire is NOT extinguished within 15 minutes of AND I EITHER one of the following:
y Damaged Safety System (s) or Plant Vital Structure E
- Receipt of a Valid Fire Alann
- l is required for the prdsent Operational Condition L 1 g
- Report of a fire from the scene S ' THEN 4 THEN y Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
3 gv d- J '
9.0 Hazards -Internal / External "%'.5" 9.3 Explosion Explosion Affecting the Operability of Plant Initiating Natural and Destructive Phenomena Safety Systems Reqmred to Establish or Condition Affecting the Protected Arca Maintain Safe Shutdown j OPCON ( All ) ( All )
EAL# 9.3.1 9.3.2 5 IF IF E l M l Confirmed Explosion Confirmed Explosion within ANY one of the E within following Plant Vital Structures:
l R the Protected Area . Reactor Building j
G t I AND ' Control / Aux Building E
i N I
- Senice Water Intake Structure C l Report of visible damage to Plant y equipment or Protected Area
- Senice/ Rad Waste Building l Permanent Plant Structures i I AND A ' THEN C The Explosion is of a m.gnitude that it SPECIFICALLY l results in Damage to ANY one of the following:
T I 1
I e i TWO OR MORE subsystems of a Safety System O I
- MORE THAN ONE Safety System N l Any Plant Vital Structure which renders the structure l
b I incapable of performing its Design Function E I V AND l
E l
Damaged Safety System (s)or Plang Vital Structure is L 1 required for the present Operational Condition 3
S '
,, , ,THEN Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
O O 9.0 Hazards -Internal / External O""J' "** ' # '
9.4 Toxic Gases
" "8 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant di ion OPCON ( All ) ( All ) ( All ) ,
EAL# 9.4.1.a 9.4.1.b 9.4.1.c
' IF IF IF E l M I Notification by Local, County, or Uncontrolled Toxic Gas Uncontrolled Flammable Gas E i State Oflicials for the potential need release within the release within the Protected Area R 3 to EVACUATE in ANY area which Protected Area G non-essential personnel does uot normally require an that RESULTS in E 3 due to an Flammable Gas concentrations atmospheric survey N Offsite Toxic Gas release or Respiratory Protection for EXCEEDING
' enW .ofthe m y i AND I
i A I SNSS deems evacuation C l of non-essential personnel AND T I is required i l Routine Plant Operations are IMPEDED based O I on EITHER one of the following:
N l I
- Access restrictions caused by the uncontrolled release L l
- Personnel injuries have occurred as a result of the release E 3 V i e l THEN THEN, L I S l , y Refer to Attachment 24 Refer to Attachment 1 Action j UNUSUAL EVENT (Common Site) UNUSUAL EVENT Required
n V O V
O 9.0 Hazards -Interncl/ External V"E77 9.4 Toxic Gases ""* * # '
Initiating Release of Toxic or Flammable Gases Within a Facility Stmcture Which Jeopardizes Operation of Systems Condition Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Conditions ,
OPCON ( All ) ( All )
EAL# 9.4.2.a 9.4.2.b l 1 IF IF E l g i Uncontrolled Toxic Gas release within ANY one Uncontrolled Flammable Gas release within ANY one E of the following Plant Stmetures of the following Plant Stmetures l
R I
- Reactor Building
- Reactor Building i
G
- Turbine Building
- Turbine Building E I
. Control / Aux Building
- Control / Aux Building N I Service Water Intake Stmeture Service Water Intake Stmeture I
C y
I
- Service / Rad Waste Building
- Service / Rad Waste Building I AND AND 1
A l I Toxic Gas concentrations result in ANY one C of the following: Flammable Gas concentrations EXCEED l
T 50% of the LEL l . An IDLit atmosphere l 1
Plant personnel report severe adverse health reactions, O l including burning eyes, nose, throat, dizziness N
The Lower Toxicity Limit being EXCEEDED L l i
E AND I V i g i Plant personnel are unable to perform actions necessary to complete a Safe g
i Shutdown of the plant without appropriate personnel protection equiptnent l S l THEN Action Refer to Attachment 2 Required ALERT l
l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ \
O O O 9.0 Hazards -Internal / External "?J' i" 9.5 Scismic Event "** ' # '
l Initiating Natural and Destructive Phenomena Natural and Destnictive Phenomena Condition Affecting the Protected Area Affecting the Plant Vital Area ,
OPCON ( All [] ( All ) ( All )
EAI, # 9.5.1.a 9.5.1.b 9.5.2
} IF IF E i
! Seismic Event felt Valid Actuation of the I by personnel Seismic Trigger (> 0.0lg)
I within the has occurred as verified by the Protected Area SM A-3 Event Indicator (flag) l I _ being WIIITE j n Panel 10-C-673 in the C i y Upper Relay Room I
i A i I
C Valid Actuation of the Seismic Switch (> 0.lg)
T i has occurred as verified by EITHER one I l of the following:
O l AND N i ,
- Valid actuation ofMain Control Room 1
THEN L i AMBER Alarm light on the Seismic Switch E
- Power Supply Drawer is lit on Panel 10-C-673 l
V i in the Upper Relay Room l c l t L I THEN S l v v Action Refer to Attachment 24 Refer to Attachment 2 Required UNUSUAL EVENT (Common Site) ALERT i
- __ _ _-=- - __.
O O ^
O 9.0 Hazards -Internal / External "%'X
- 'd' 9.6 High Winds Initiating ,
Condition Natural and Destructive Phenomena Natural and Destructive Phenomena )
)
Affecting the Protected Area y Affecting the Plant Vital Area /
OPCON
( Ml ) ( MI ) ( Mi )
EAL# 9.6.1.a 9.6.1.b 9.6.2 i IF IF E l M Report of a Sustained wind speeds l
E I Tornado > 75 MPH R TOUCHING for 15 minutes, l
G I DOWN measured at E within the ANY elevation of l
N I Protected Area the Met Tower The Wind Speed is of a magnitude that it SPECIFICALLY C l results in Damage to ANY one of the following:
l
- TWO OR MORE subsystems of a Safety System A !
- MORE THAN ONE Safety System C
, Rendering ANY of the following stmetures incapable of performing g
i THEN its Design Function:
- Reactor Building O l "
i
- Control / Aux Building N
l
- Service Water Intake Structure g i
- Service /Radwaste Building E AND l
V i I
Damaged Safety System (s) or Plant Vital Structure E ,
g is required for the presend Operational Condition l
S l ,, ,
, THEN Refer to Attachment 24 Refer to Attachment 2 Action ALERT UNUSUAL EVENT (Common Site)
Required
l O O)
V (.
9.0 Hazards -Internal / External (v')"S *i' "8* ' # '
9.7 Flooding Initiating Intemal Flooding in Excess of Sump Handling Capability Internal Flooding Affecting the OperabilityorPlam Safety )
l Condition Affecting Safety Related Areas of the Plant Systems Required to Establish or Maintain Safe Shutdown ,/
i OPCON ( All ) ( All )
EAL# 9.7.1 9.7.2 !
' IF IF E l M Visual Observation ofIIncontrolled Flooding that confirms Visual Observation of Flooding within ANY one l
E I ANY one of the following: of the following Plant Vital Structures:
!
- Reactor Building Floor Levels above the Maximum
- Reactor Building Normal Floor Level (>l") referenced in EOP 103,
- Control / Aux Building E l I Secondary Containment Control . Service Water Intake Structure N
i
- Receipt of a SSWS Pump Room Flooded Alarm . Service / Rad Waste Building ;
C ;
y l
- Greater than 2" of water in ANY area that contains a i Safety System (s), not included above AND j
A i The Flooding is of a magnitude that it SPECIFICALLY 4 i
C results in Damage to ANY one of the following:
T l I
- TWO OR MORE subsystems of a Safety System l
O I . MORE TIIAN ONE Safety System N
- Any of the above listed Plant Vital Structures which renders l the structure incapable of performing its Design Function E AND l
V i E I Damaged Safety System (s) or Plant Vital Structure is required for '
the present Opefational Condition L !
S '
' ,, THEN Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
O O O 9.0 Hazards -Internal / External "%'.'4 9.8 Turbine Failure / Vehicle - Missile Impact Initiating Condition Natural and Destructive Phenomena Affecting Certain Structures Within the Protected Area Natural and Destmetive Phenomena Affecting Certain Structures Within the Plant Vital Area,/
)
OPCON ( 1,2,3 ) ( All ) ( All )
EAL# 9.8.1.a 9.8.1.b 9.8.2 i IF IF IF E l M Catastrophic damage to the Vehicle Crash / Vehicle Crash / Missile Impact with or within l
E Main Turbine as evidenced by ANY one of the following Plant Vital Structures:
I Missile Impact with or R EITHER one of the following:
l within ANY one of the
- Reactor Building i
- Main Turbine casing following Plant Structures:
- Control / Aux Building i penetration
- Reactor Building . Service Water Intake Structure
- Main Turbine / Generator . Turbine Building . Service / Rad Waste Building C I y I Damage potentially releasing
- Control / Aux Building i Lube Oil or Hydrogen Gas AND I
- Service Water Intake to the Turbm, e Building A I Structure The Vehicle Crash / Missile Impact is of a magnitude C i
- Service / Rad Waste Building that it SPECIFICALLY results in Damage to ANY l one of the following:
O l
- TWO OR MORE subsystems of a Safety System N l
- MORE THAN ONE Safety System i
- Any of the above Plant Vital Structures which renders L ! the stmeture incapable of performing its Design Function g i TMEls y AND l
E Damaged Safety Systgm(s) or Plant Vital Structure l
L i is required for the present Operational Condition b
y THEN Action Refer to Attachment 1 Refer to Attachment 2 Required UNUSUAL EVENT ALERT
l
\ .
iO l
l BASIS DOCUMENT 1
! HOPE CREEK ECG l ~(FINAL DRAFT)
SECTIONS 1 - 9 O EMERGENCY ACTION LEVELS (EALs)
NOTE: THIS IS A NEW DOCUMENT WHICH WILL BE SEPARATE FROM THE ECG AND USED AS A REFERENCE DOCUMENT.
BASIS DOCUMENT FILE: COVERS.ECG J
1 l
, HCGS EAURALTechnical Basis 1.0 Fuel Clad Challenge -
0 1.1 RCS Activity UNUSUAL EVENT - 1.1.1.a IC Fuel Clad Degradation EAL Reactor Coolant Sample Activity > 4 Ci/gm 3 Dose Equivalent I-131 OPERATIONAL CONDITION - 1,2,3,4,5 BASIS A Reactor Coolant sample analysis with specific activity in excess of the Technical Specification limit of 4 pCi/gm Dose Equivaleat Iodine-131 (DEI-131) is indicative of a degradation of the fuel clad, and is a precursor of more serious problems. This activity level is chosen instead of the 0.2 Ci/gm DEI-131 Technical Specification limit, under which i, operation is allowed to continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to accommodate short duration Iodine spikes following changes in thermal power. This EAL threshold does not use the term
" Valid", since Reactor Coolant Sample Activity of greater than 4 uCi/gm DEI- 131 can only occur as the result of fuel clad degradation and not as the result of a resin / chemical intrusion transient or HWCI System malfunction. Unusual Event declaration is warranted on]y when actual fuel clad degradation has occurred.
Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier.
ESCALATION CRITERIA i
Emergency Classification will escalate to an Aleit (1.1.2) when a sample analysis of Reactor Coolant activity exceeds 300 Ci/gm DEI-1-31 per EAL Section 3.1.3.
(-
%J EAL - 1.1.1.a Rev. 00 Page 1 of 2
. HCGS EAL/RALTechnical Basis c DISCUSSION -
/
()/ The Technical Speci6 cation limit on Reactor Coolant activity ensures that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a Main Steam Line failure outside the containment during i
steady state operation will not exceed a small fraction of the 10CFR100 limits. This limit accommodates Iodine Spiking, which frequently occurs following shutdowns, startups, rapid power changes and coolant depressurization. Iodine spikes are characterized by a rapid increase in Reactor Coolant Iodine concentration by as much as three orders of magnitude followed by a return to prespike concentrations. This spiking is a temporary excursion and is not caused by a sudden fuel failure. The Technical Srsvi6 cation limit of > 100/ EpCifgm is '
excluded from this EAL because this limit does not inclu:le Iodine Activity. l l
DEVIATION j NUMARC EAL SU 4.2 suggests that the Operating hkde Applicability for this EAL is ALL.
When the Reactor is defueled, the source term needed to achieve an RCS Activity of 4 uCi/gm Dose Equivalent I-131 is not available. Hence, this EAL is applicable in Operational Conditions 1,2,3,4 and 5.
REFERENCES NUMARC NESP-007, SU4.2 Technical Specification LCO 3.4.5 HC.OP-AB.ZZ-0100(Q), High Reactor Coolant Activity HC.OP-AB.ZZ-0203(Q) Main Steam Line High Radiation 10 CFR100 l
l l
' bD EAL - 1.1.1.a Rev.00 l Page 2 of 2 l
l
, HCGS EAL/RALTechnical Basis 1.0 Fuel Clad Challenge -
0 1.1 RCS Activity '
UNUSUAL EVENT - 1.1.1.b IC Fuel Clad Degradation EAL Valid Offgas Pretreatment Radiation Monitor (9RX621/ 9RX622)
High Alarm Condition (.;>_2.2E+04 mrem /hr)
OPERATIONAL CONDITION - 1,2,3,4 BASIS A Valid Offgas Pretreatment Radiation Monitor High alarm is indicative of a degradation of the fuel clad, and is a precursor of a more serious problems. The alarm is set at 2.2E+04 mR/hr, which ensures that the alarm will actuate prior to exceeding the Technical O Specification Offgas System Noble Gas Effluent Limit of 3.3E5 pCi/s. Valid is defined as the Offgas Pretreatment Radiation Monitor High Alarm actuatine soecifically due to fuel clad degradation, thus precluding unwarranted Unusual Event declaration as the result of a resin /
chemical intrusion transient, or HWCI System malfunction. Unusual Event declaration is j warranted only when actual fuel clad damage has occurred.
Barrier Analysis )
This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier.
ESCALATION CRITERIA Emergency Classification will escalate to an Alert (1.1.2) when a sample analysis of Reactor l Coolant activity exceeds 300 Ci/gm DEI-131 per EAL Section 3.1.3.
i l
i
' l
- O '
1
- V EAL - 1.1 Rev. ,J Page 1 of 2 l
l
. HCGS EAL/RALTechnical Basis i l
DISCUSSION l O The Offgas Pretreatment Radiation Monitors (9RX621/ 9RX622) monitor gamma radiation !
levels attributable to the non-condensable fission product gases produced in the reactor and transported with steam through the turbine to the condenser. This instrument takes a sample from the sample tap between the fourth and fifth holdup pipe of the Offgas system.
Restricting the gross radioactivity from the Main Condenser provides reasonable assurance that l the Total Body exposure to an individual at the exclusion area boundary will not exceed a f
small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently- ,
discharged directly to the environment without treatment.
Operating Experience at HCGS has demonstrated that Reactor coolant activity changes for i reasons other than fuel clad degradation can result in increasing Main Steam Line Radiation i l Monitors and Offgas Pretreatment Radiation Monitor. Such events (e.g. - resin intrusion) do l not require classification under this EAL.
4 DEVIATION NUMARC EAL SU 4.1 suggests that the Operating Mode Applicability for this EAL is ALL. J In Operational Condition 5 and Defueled, the MSIVs will be closed, thus rendering the Offgas Pretreatment Radiation Monitors unavailable for detection of increased RCS Activity. Hence, this EAL is applicable in Operational Conditions 1,2,3 and 4.
REFERENCES NUMARC NESP-007, SU4.1 Technical Specifications; Table 3.3.7.1 (5); LCO 3.11.2.7 HC.OP-AB.ZZ-0100(Q), High Reactor Coolant Activity HC.OP-AB.ZZ-0203(Q) Main Stea:n Line High Radiation HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response OE-6144, Resin Intrusion 10 CFR100 l
1
[
iO EAL - 1.1.1.b Rev.00 Page 2 of 2
i HCGS EAL/RALTechnical Basis 1.0 Fuel Clad Challenge -
0 1.1 RCS Activity UNUSUAL EVENT - 1.1.1.c IC Fuel Clad Degradation I l
EAL 1
Valid Main Steam Line Radiation Monitor High High Alarm Condition l (13 times Normal Full Power Background)
OPERATIONAL CONDITION - 1,2,3,4 l l
BASIS ,
' A Valid Main Steam Line Radiation Monitor High High alarm (1 3 times normal full power background) is indicative of degradation of the fuel clad and may be a precursor of more serious problems. Valid is defined as the Main Steam Line Radiation Monitor High High O Alarm actuating soecifically due to fuel clad degradation, thus precluding unwarranted Unusual Event declaration as the result of a resin / chemical intrusion transient, or HWCI System malfunction. Unusual Event declaration is warranted onb when actual fuel clad degradation has occurred. Reaching the High High Alarm on M of the 4 Main Steam Line Radiation
. Monitor channels, as determined by receipt of M one of the following, due to fuel clad degradation, warrants Unusual Event declaration.
Overhead Annunciator C6-B2, MN STM LINE RAD HI HI OR INOP CRIDS Point D2121, MN STM LINE HI HI RAD / INOP - W CRIDS Point D2122, MN STM LINE HI MI RAD / INOP - X CRIDS Point D2123, MN STM LINE HI HI RAD / INOP - Y CRIDS Point D2124, MN STM LINE HI HI RAD / INOP - Z Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier.
!O
! EAL - 1.1.1.c Rev.00 Page 1 of 3
i HCGS EAL/RALTechnical Basis :
ESCALATION CRITERIA -
O Emergency Classification will escalate to an Alert (1.1.2) when a sample analysis of Reactor Coolant activity exceeds 300 uCi/gm DEI-131 per EAL Section 3.1.3.
DISCUSSION The Main Steam Line Radiation Monitor Channels (9RX509, 9RX510, 9RX511, 9RX512) monitor gamma radiation levels at the Main Steam Lines. A High High alarm is indicative of a release of gap activity to the coolant but is not indication of a major failure of the fuel clad.
A Valid Main Steam Line Radiation High High alarm condition requires a manual Reactor r Scram and Main Steam Isolation Valve closure to reduce and isolate the potential source of the radioactivity release.
The terminology used for the 3 times Normal Full Power Background threshold differs ;
between the Main Control' Room Overhead Annunciators and the Radiation Monitoring System (RM-11). As a result, specific monitor channels are not included in the EAL. Overhead Annunciators use the terminology of "High High" for this threshold, where the RM-11 uses the terminology of "High" for the same threshold. For the purpose of this EAL, the High High setpoint terminology used by the Overhead Annunciators is used, though the same indications are available on the following RM-11 Channels: ;
Main Steam Line " Channel" A (Grid 1/4; 9RX509)
Main Steam Line " Channel" B (Grid 1/4; 9RX510)
Main Steam Line " Channel" C (Grid 1/4; 9RX511)
Main Steam Line " Channel" D (Grid 1/4; 9RX512)
In addition, the Main Steam Line Radiation Monitor Numac Drawers can be used to trend changes in Main Steam Line Radiation I_evels.
A rapid power reduction from full power may cause the Main Steam line Radiation Monitors to momentarily increase to 1.5 times normal full background readings. This is due to the response time of the HWCI Hydrogen Flow Controller and the transport time from the Hydrogen Injection point (Secondary Condensate Pumps).
Operating Experience at HCGS has demonstrated that Reactor coolant activity changes for reasons other than fuel clad degradation can result in increases in Main Steam Line Radiation Monitors and Offgas Pretreatment Radiation Monitors. Such events (e.g. - resin intrusion) do not require classification under this EAL.
DEVIATION NUMARC EAL SU 4.1 suggests that the Operating Mode Applicability for this EAL is ALL.
In Operational Condition 5 and Defueled, the MSIVs will be closed, thus rendering the Main Steam Line Rad Monitors unavailable for detection of increased RCS Activity. Hence, this fl V EAL is applicable in Operational Conditions 1,2,3 and 4.
EAL - 1.1.1.c Rev. 00 Page 2 of 3
. HCGS EAL/RALTechnical Basis REFERENCES -
O NUMARC NESP-007, SU4.1 HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-0011(Qi, Annunciator Response Procedures, Window C6-B2 HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HCGS Technical Specifications 3/4.3, Instrumentation Technical Specifications, LCO 3.11.2.7 OE-6144, Resin Intrusion 10CFR100 O
O EAL - 1.1.1.c Page 3 of 3
. HCGS EAURALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity ALERT - 1.1.2 ,
IC Fuel Clad Degradation l EAL -
^
i i
ANY one of the following: !
- Reactor Coolant Sample Activity >4 Ci/gm Dose Equivalent I-131
. Valid Offgas Pretreatment Radiation Monitor (9RX621/ 9RX622) High Alarm l Condition (.2.2.2E+04 mrem /hr) '
- ' Valid Main Steam Line Radiation Monitor High High Alarm Condition (.2.3 times Normal Full Power Background)
E :
ANY SRV is determined to be Stuck Open OPERATIONAL CONDITION - 1,2,3 BASIS Indication of Fuel Clad Degradation coincident with AHX SRV determined to be Stuck Open is indicative of a Loss of the RCS Barrier, as fission products are being transported directly to the Suppression Pool, thus compromising the integrity of the RCS Barrier. Hence, Alert i declaration is warranted. In the event an SRV is Stuck Open with NQ indications of Fuel Clad degradation, an emergency declaration is HQI warranted, since an open SRV is within the analyzed design envelope of the plant and does not, by itself, represent a degradation in the level of plant safety. An SRV is considered to be Stuck Open when the SRV can not be reclosed by operator action within 2 minutes of ANY spurious, automatic or manual actuation.
A Stuck Open SRV SHOULD NOT be considered as an Unisolable RCS Leak > 50 GPM, as the consequences of a Stuck Open SRV discharging to the Suppression Pool are different than i an Unisolable RCS Leak exceeding 50 GPM that is discharging into the Drywell Air Space.
A Stuck Open SRV by itself requires a 1 Hour Report if a Unit Shutdown (Manual Reactor i
Scram) is initiated to comply with Technical Specification or a 4 Hour Report if the SRV is reclosed within the Technical Specification limits, due to the ESF actuation.
l l
EAL - 1.1.2 Rev. 00 Page 1 of 2 l
l
. HCGS EAL/RALTechnical Basis Barrier Analysis l RCS Barrier has been lost.
ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0
- 1 DISCUSSION !
A Stuck Open SRV discharging Reactor Coolant to the Suppression Pool does not represent the same challenge to the RCS and Primary Containment as an Unisolable RCS Leak i discharging into the Drywell. The consequences of a Stuck Open SRV do not represent a significant precursor to further plant degradation, as plant design (Pressure Suppression ability of the Torus) and the Abnormal Operating Procedure for a Stuck Open SRV (directing a Manual Reactor Scram within 2 minutes if the SRV can not be closed), minimize the consequences of the event. In contrast, an Unisolable RCS Imk represents a situation where there is concern for " break propagation", which could lead to a significantly larger uncontrolled for loss of RCS inventory. Hence, a Stuck Open SRV must be coincident with Fuel Clad Degradation for the RCS Barrier to be considered lost.
' D DEVIATION
[G l None REFERENCES l
NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #7" f HC.OP-AB.ZZ-0121(Q), Failed Open Safety / Relief Valve l
l l EAL - 1.1.2 l Rev.00 Page 2 of 2
. HCGS EAL/RALTechnical Basis f- 2.0 RCS Challenge -
2.1 RCS Leakage UNUSUAL EVENT - 2.1.1.a / 2.1.1.b IC RCS Leakage EAL _
EITHER one of the following:
- Reactor Coolant System Presssure Boundary Leakage > 10 gpm (Using 10 minute average)
- Reactor Coolant System Unidentified Leakage > 10 gpm (Using 10 minute average)
OPERATIONAL CONTITION - 1, 2, 3 BASIS RCS Pressure Boundary and Unidentified Leakage exceeding 10 gpm is indicative of possible degradation of the RCS and may be a precursor of a more serious condition. RCS Operational Leakage addressed by these 2 EALs is soecifically RCS leakage into the Drywell. Leakage into the Drywell that is confirmed to HD1 be RCS Leakage, i.e. a leaking Drywell Cooling Coil, does not warrant classification under this EAL. These types of RCS Operational Leakage, exceeding their respective EAL thresholds, should be classified as an Unusual Event, regardless of whether or not the leak has been isolated, since the EAL thresholds exceed the Technical Specification limit. Classification should be based on the 10 minute average and not an instantaneous value, to assure accurate event classification.
The value of 10 gpm for RCS Pressure Boundary and Unidentified Leakage was set higher than the Technical Specification limit of 0 and 5 gpm respectively, to allow time to implement corrective actions (including plant shutdown) prior to exceeding the threshold.
Only operating conditions in which there is fuel in the reactor coolant system and the system is pressurized are specified.
!O EAL - 2.1.1.a / 2.1.1.b Rev. 00 Page 1 of 3
HCGS EAL/RALTechnical Basis l Barrier Analysis -
lO t This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier.
l
! ESCALATION CRITERIA L
f Emergency Classification will escalate to an Alert when either Unisolable RCS Leak Rate
- j. -
exceeds 50 gpm or Drywell Pressure exceeds 1.68 PSIG per EAL Section 3.2.2 l
DISCUSSION f
Allowable leakage rates from the Reactor Coolant System are based on predicted and i experimentally observed behavior of cracks in pipes. Utilizing the leak before break methodology, it is anticipated that there will be indication (s) of minor reactor coolant system
- l. boundary leakage prior to a fault escalating to a major leak or a system rupture. Detection of low levels of leakage while pressurized allows for implementation of mitigative actions and i permits monitoring for catastrophic failure or rupture precursors. l l The limit for Unidentified and Pressure Boundary Leakage is set to a lower value, than Identified Leakage due to concern over " break propagation" resulting from an Unidentified or Pressure Boundary Leak (Small Break), that could potentially lead to a significantly larger loss i
O O
of inventory. Identified leakage occurs when there is degradation or failure of a mechanical joint. Pipe " break propagation" is thus not an issue.
Instrumentation available via the Radiation Monitoring System (RM-11) to determine RCS Leakage into the Drywell includes:
- (9AX313) Drywell Equipment Drain Sump (DLD EQPT) Monitor
. (9AX314) Drywell Floor Drain Sump (DLD FLR) Monitor
- (9AX317) lower Drywell Air Condensate Coolers (DLD CCM LOW) Monitor e (9AX318) Upper Drywell Air Condensate Coolers (DLD CCM UP) Monitor
- (9AX319) Drywell Sumps (DLD SMS) Monitor e (9AX320) Drywell Air Condensate Coolers Summation (DLD CCM SUM) Monitor Redundant Instrumentation for Drywell I2ak Detection is available on panel 10-C-604 located
. in the back of the Main Control Room.
Technical Specification required actions based on this leak rate may require a plant shutdown i
l and subsequent depressurization, unless the source of the leak can be located, identified, l and/or stopped.
DEVIATION None
! EAL - 2.1.1.a / 2.1.1.b Rev.00 !
Page 2 of 3 l
. HCGS EAURALTechnical Basis g REFERENCES b NUMARC NESP-007, SUS NUMARC Questions and Answers, June 1993, " General Question #12" NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #11" HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-Oll6 (Q), Containment Isolation and Recovery From An Isolation ,
HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control I
HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control l HC.OP-GP.ZZ-0005 (Q), Drywell I.eakage Source Detection HCGS Technical Specifications, LCO 3.4.3.2 O
l 1
EAL - 2.1.1.a / 2.1.1.b Rev. 00 l Page 3 of 3
i HCGS EAURALTechnical Basis 2.0 RCS Challenge 2.1 RCS Leakage UNUSUAL EVENT - 2.1.1.c IC RCS I2akage EAL _
i 1
l Reactor Coolant System Identified leakage > 25 gpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period OPERATIONAL CONDITION - 1, 2, 3 l BASIS i
RCS Identified leakage exceeding 25 gpm is indicative of possible degradation of the RCS ]
and may be a precursor of a more serious condition. RCS Operational Leakage addressed by :
I Q
V this EAL is soecifically RCS leakage into the Drywell. Leakage into the Drywell that is confirmed to not be RCS Leakage, i.e. a leaking Drywell Cooling Coil, does not warrant 1 classification under this EAL. Identified Izakage should ONLY be classified as an Unusual '
Event, when the leak rate exceeds 25 gpm when averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, regardless of whether or not the leak has been isolated. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average is included as part of the EAL threshold to provide consistency with the Technical Specification limit for Identified Leakage.
t Only operating conditions in which there is fuel in the reactor coolant system and the system is l pressurized are specified.
l Barrier Analysis l This event does not reach the threshold for the loss of the RCS Barrier, but does affect j that barrier.
i ESCALATION CRITERIA Emergency Classification will escalate to an Alert when either Unisolable RCS Leak Rate
- exceeds 50 gpm or Drywell Pressure exceeds 1.68 PSIG per EAL Section 3.2.2 O EAL - 2.1.1.c Rev.00 Page 1 of 3
. HCGS EAL/RALTechnical Basis DISCUSSION O Allowable leakage rate: from the Reactor Coolant System are based on predicted and experimentally observed behavior of cracks in pipes. Utilizing the leak before break.
methodology, it is anticipated that there will be indication (s) of minor 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 allows for implementation of mitigative actions and l
permits monitoring for catastrophic failure or rupture precursors.
l
! The limit for Unidentified and Pressure Boundary 12akage is set to a lower value, then --
Identified Leakage due to concern over " break propagation" resulting from an Unidentified or Pressure Boundary Leak (Small Break), that could potentially lead to a significantly larger loss of inventory. Identified leakage occurs when there is degradation or failure of a mechanical
! joint. Pipe " break propagation" is thus not an issue.
l Instrumentation available via the Radiation Monitoring System (RM-11) to determine RCS leakage into the Drywell includes:
L * (9AX313) Drywell Equipment Drain Sump (DLD EQPT) Monitor
- (9AX317) Lower Drywell Air Condensate Coolers (DLD CCM LOW) Monitor l * (9AX318) Upper Drywell Air Condensate Coolers (DLD CCM UP) Monitor l * (9AX319) Drywell Sumps (DLD SMS) Monitor
- (9AX320) Drywell Air Condensate Coolers Summation (DLD CCM SUM) Monitor Redundant Instrumentation for Drywell Leak Detection is available on panel 10-C-604 located in the back of the Main Control Room.
l Technical Specification required actions based on this leak rate may require a plant shutdown and subsequent depressurization, unless the source of the leak can be located, identified, and/or stopped.
DEVIATION NUMARC EAL SU5 suggests that exceeding an RCS Identified Leakage limit of 25 gpm warrants the declaration of an Unusual Event because it may be a precursor to a more serious condition. The Hope Creek Technical Specification limit for RCS Identified Leakage is 21 GPM averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> oeriod. The plant is within Technical Specification as long as this limit is not exceeded and hence an Unusual Event is not warranted until the limit is l
exceeded. This philosophy is consistent with that contained in NUMARC EAL SU2, which i only requires declaration of an Unusual Event when the plant is outside the Technical Specification. RCS Pressure Boundary and Unidentified Leakage that exceed the NUMARC EAL threshold will be classified as an Unusual Event, as this leakage exceeds the Technical Specification limit.
4 EAL - 2.1.1.c l Rev. 00 l Page 2 of 3 I
< 1 l
l HCGS EAIJRALTechnical Basis ;I 1
In addition, NUMARC EAL SU5. appears to apply specifically to those plants that do not l allow for averaging of RCS Identined Leakage over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. Furthermore, NUMARC Questions and Answers Document , June 1993, " General Question #12", addresses j those cases where the Technical Soeci6 cation LCO has been exceeded and the required Action l section has been entered (i.e. 4 Hours to identify and reduce the leakage below the limit). The l EAL threshold for RCS Identified Leakage does not consider this time for Unusual Event l declaration. The Q&A also states that the EAL for RCS Identified I2akage has been i significantly raised from 10 to 25 gpm at some plants. Since the Hope Creek Technical Specification limit is already set at 25 gpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the EAL should not be more limiting than the Technical Specifications. ,
l i
REFERENCES NUMARC NESP-007, SUS NUMARC Questions and Answers, June 1993, "Gewal Question #12" NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #11" l HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling i HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control i HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control l HC.OP-GP.ZZ-0005 (Q), Drywell leakage Source Detection ;
HCGS Technical Specifications, LCO 3.4.3.2 F
i i
I i
lO EAL - 2.1.1.c Rev.00 Page 3 of 3 e______.___ . _ _ _ . _. _. , _
, HCGS EAL/RALTechnical Basis 2.0 RCS Challenge 2.1 RCS Leakage UNUSUAL EVENT - 2.1.1.d IC RCS Leakage EAL _
Successful Isolation of a Reactor Recirc Pump Dual Seal Failure within 10 minutes of recognition OPERATIONAL CONTITION - 1,2, 3 BASIS Successful Isolation of a Reactor Recire Pump Dual Seal Failure within 10 minutes of recognition is classified as an Unusual Event, due to the significance of the event. Even A though the consequences of a successfully isolated Recirc Pump Dual Seal failure are minor, V with no possibility for " break propagation", an Unusual Event is warranted due to the multiple failures of mechanical joints that allowed the discharge of a significant quantity of Reactor Coolant (>50 GPM) directly into the Drywell Air Space.
Successful is defined as indication of ALL of the following within 10 minutes of recognition of the Recirc Pump Dual Seal failure.
- Recirc Pump Suction and Discharge Valves have closed
- RWCU Suction Valve from the Recire Imop has closed
- Recirc Pump Seal Purge Water Valve have closed
- Drywell Pressure and Temperature has begun to decrease
- RCS Leakage has begun to decrease 10 minutes was determined to be a reasonable amount of time to isolate the pump and monitor for the effectiveness of the actions.
Only operating conditions in which there is fuel in the reactor coolant system and the system is pressurized are specified.
O v
EAL - 2.1.1.d Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis Barrier Analysis -
This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier.
ESCALATION CRITERIA Emergency Classification will escalate to an Alert if ten minutes elapse prict to successful 1 isolation or Drywell Pressure exceeds 1.68 PSIG per EAL Section 3.2.2 l DISCUSSION Prompt recognition of a Recire Pump Dual Seal failure by the operating crew will allow for implementation of actions to isolate the leakage source in accordance with Abnormal Operating ;
Procedures. The design of the Recirc Pump seal limits the magnitude of the identified leakage I for this event to 60 gpm due to the presence of a breakdown bushing. As a result, RCS inventory will not be significantly effected. The ability to monitor the leak rate is limited to 50 l gpm, the upper limit of the Drywell Leak Detection Instrumentation. Drywell Pressure is not expected to reach the High Drywell Pressure Scram setpoint for this event, provided that the isolation was successfully completed within 10 minutes. l l
DEVIATION None .
REFERENCES NUMARC NESP-007, SU5 NUMARC Questions and Answers, June 1993, " General Question #12" NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #11" i HC.OP-AB.ZZ-0112 (Q), Recirculation Pump Trip j HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling )
HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control ;
HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control l HC.OP-GP.ZZ-0005 (Q), Drywell Leakage Source Detection HCGS Technical Specifications, LCO 3.4.3.2 l O EAL - 2.1.1.d Rev.00 Page 2 of 2 a
HCGS EAl)RALTechnical Basis g 3.0. Fission Product Barriers
\J 3.1 Fuel Clad Barrier 3.1.1.a IC Potential loss of Fuel Clad EAL _
Reactor Water Level REACHES - 161" (Top of Active Fuel), EXCLUDING intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONTITION - 1, 2, 3 BASIS Reactor Water 12 vel reaching -161" (Top of Active Fuel - TAF), excluding intentional lowering of Reactor Water level during an ATWS, results in an inability to maintain adequate core cooling by core submergence, causing a Potential IAss of the Fuel Clad
(]
V Barrier. Without core submergence, the integrity of the fuel clad barrier is in jeopardy.
Appropriate classification under this EAL is based on reachine Reactor Water Level of -
161" (instead of being able to restore and maintain above -161") due to the potentially severe consequences of a loss of core submergence. Reactor Water Level reaching this threshold results from either a LOCA exceeding available makeup capacity or a Total Loss of High Pressure injection capability.
In addition, during an Anticipated Transient Without Scram (ATWS), it is possible that operator actions will be taken to intentionally lower Reactor Water I2 vel to between -161" and -190", for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Section 5.0 Barrier Analysis Fuel Clad Barrier has been potentially lost ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0.
l (d EAL - 3.1.1.a Rev.00 Page 1 of 3 l
HCGS EAURALTechnical Basis DISCUSSION O Core Submergence is the preferred method of maintaining adequate core cooling. When t Reactor Water level decreases to below TAF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad fission product barrier can no longer be l
considered intact. While the Emergency Operating Procedures provide contingencies to i
l establish adequate core cooling when Reactor Water Level drops below TAF (Steam Cooling l with or without injection), these actions are designed to be an alternative method of providing l
adequate core cooling while actions are taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel clad damage and a significant release of fission products to the Reactor coolant. Sustained core uncovery can also result in a breach of the Reactor Vessel due to core melt material interaction with the RPV.
A less of Core Submergence will occur when the rate of inventory loss is greater than the rate of inventory makeup from High Pressure injection sources. This condition can occur as the result of the following events / sequences (excluding intentional lowering of Reactor Water Level during an ATWS).
A LOCA will cause Reactor Water Level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leakrate.
A Loss of High Pressure injection sources without the presence of a LOCA will also j
result in Reactor Water Level decreasing to TAF, due to continued Reactor Steam I
Flow without makeup.
Either of these events / sequences results in a challenge to the Fuel Clad Barrier when Reactor Water Level reaches TAF due to core uncovery, hence classification at this threshold is l
appropriate. However, for both these sequences, Low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shuto'ff head of the pumps. Reactor Depressurization will occur either due to the LOCA or Manual initiation of Emergency Depressurization when Reactor Water Level reaches -161", provided injection systems are '
available. This will allow for restoration of Reactor Water Level and re-establishment of Core Submergence. Failure of these systems to restore and maintain Reactor Water Level above -
200" will require escalation.
DEVIATION
.i
- None
!O EAL - 3.1.1.a Rev.00 Page 2 of 3 I- _
HCGS EAURALTechnical Basis
- l l
REFERENCES O NUMARC NESP-0007, FC2 l
HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram l HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP-EO.ZZ-0207 (Q)-FC, Level / Power Control i BWR Owner's Group Emergency Procedure Guidelines, Rev. 4 l l
l l
l l
i lO 1
l l
i l0 EAL - 3.1.1.a Rev.00 Page 3 of 3
__ - -* -a
l I ,
HCGS EAL/RALTechnical Basis 3.0 Fusion Product Barriers 3.1 Fuel Clad Barrier l
I 3.1.1.b IC Ioss of Fuel Clad ML _
i Reactor Water Level CANNOT BE RESTORED AND MAINTAINED above -200" (Minimum Zero Injection RPV Water Level) i OPERATIONAL CONDITION - 1, 2, 3 BASIS Inability to restore and maintain Reactor Water Level above - 200" (Minimum Zero Injection RPV Water Level), results in a loss of adequate core cooling by all mechanisms, causing a Loss of the Fuel Clad Barrier. Without adequate core cooling, the integrity of the fuel clad
/ barrier can no longer be assured. Appropriate classification under this EAL is based on the failure ofinjection systems to restore and mainhin Reactor Water Level above > -200", ;
following a condition that causes level to decrease below the threshold. - For example, a large i l
break LOCA is expected to cause Reactor Water Level to momentarily decrease below -200",
due to the response time of Low Pressure ECCS. As these systems initiate and commence i J
injection to the Reactor, water level will begin to increase and should be able to be maintained above -200". In this case, classification under this EAL is not appropriate as plant systems have performed their intended design function and will eventually restore adequate core cooling by core submergence. However, in the event that IAw Pressure ECCS and alternate injection system, as defined in the EOPs are in a degraded condition (i.e., Station Blackout, L
ECCS Suction Strainer plugging, etc.) and Reactor Water Level can not be restored and ;
! maintained above -200", then classification under this EAL should occur due to the potential for release of energy to the containment from imminent fuel failure.
a Barrier Analysis Fuel Clad Barrier has been lost.
!O I EAL - 3.1.1.b Rev. 00 Page i of 3 r
HCGS EAL/RALTechnical Basis g ESCALATION CRITERIA N~] Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0.
DISCUSSION Core submergence is the preferred method for maintaining adequate core cooling. The failure to reestablish Reactor Water Level above -161", the Top of Active Fuel (TAF), for an exteraded period of time could lead to a significant of fuel damage. With Reactor Water Level below TAF, but above the Minimum Zero Injection RPV Water Level (-200"), adequate core cooling occurs due to the cooling effects of steam generated in the covered portion of the core flowing through the uncovered portion (Steam Cooling). The Minimum Ze;c Injection RPV Water level is defined in the Emergency Openting Procedures. This method of cooling precludes any fuel clad temperaturu in the uncovered portion of the core from exceeding 1800*F. As Reactor Water Level drops below -200" with no injection available, this method of cooling becomes inadequate. Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of caergy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time. Reactor Water level remaining below TAF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public O
v protective actions.
Ample time should % # owed for Low Pressure ECCS and alternate injection systems to restore Reactor Watzi Level prior to entry into this classification. The time basis for deciding whether or not Reactor Water can be maintained > -200" should be based on the rate of reactar depressurization, the availability of low pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as Reactor Water level trend, injection flow rates, containment parameter trends, and low pressure injection system operability should also bc ; ansidered.
In the event, Reactor Water Ixvel can not be restored > -200", containment flooding will be required by the EOPs. This will attempt to flood the containment as a means of flooding the RPV, and use a flooded containment as a heat sink for the nuclear fuel.
l DEVIATION None n
EAL - 3.1.1.b Rev.00 Page 2 of 3
, HCGS EAURALTechnical Basis REFERENCES d NUMARC NESP-0007, FC2 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control !
HC.OP-EO.ZZ-0201 (Q)-FC, Alternate level Control HC.OP-EO.ZZ-0207 (Q)-FC, level / Power Control HC.OP-EO.ZZ-0208 (Q)-FC, Primary Containment Flooding ,
BWR Owners Group Emergency Procedure Guidelines, Revision 4 l t
f s
O ,
i i
t l
i l
O EAL - 3.1.1.b Rev. 00 Page 3 of 3
HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier l 3.1.3 l l
l IC Loss of Fuel Clad l EAL _
Reactor Coolant Sample Activity .>._ 300 Cl/gm Dose Equivalent I-131 OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Coolant sample analysis with specific activity greater than or equal to 300 Ci/gm Dose Equivalent I-131 (DEI-131) indicates fuel clad damage due to significant clad heating or mechanical stress, causing a Loss of the Fuel Clad Barrier. This threshold is well above the
/3 activity level that could occur as the result of Iodine spiking. The use of the term " Valid" as a V qualifier for event classification is not required, since Reactor Coolant Activity of this magnitude can only occur as the result of fuel clad damage. This activity level corresponds to approximately 3.8% fuel clad damage.
Barrier Analysis i
Fuel Clad Barrier has been lost. l l
ESCALATION CRITERIA Emergency Classification will escalate based w the Potential Loss or Loss of additional Fission Preduct Barriers per EAL Section 3.0.
1 l
1 O EAL - 3.1.3 Rev.00 Page 1 of 3
1 HCGS EAURALTechnical Basis l
DISCUSSUN - '
O The percentage of Fuel Damage that corresponds to an RCS Activity of 300 Ci/gm DEI-131 is calculated as follows (for purposes of this calculation, cc and gm are considered equivalent):
Dose Factors (RG-1.109) 1-131 = 4.39E-3 I-132 = 5.23E-5 I-133 = 1.04E-3 I-134 = 1.37E-5 I-135 = 2.14E-4 ,
Total core inventory (HCGS-UFSAR, table 12.2-135). This table gives 50% inventory, so table values are multiplied by 2.0.
1-131 = 8.64E7 Ci 1-132 = 1.29E8 Ci 1-133 = 1.99E8 Ci I-134 = 2.32E8 Ci 1-135 = 1.81E8 Ci l
< - Reactor Water Volume = 13000 cubic feet (HCGS-UFSAR, table 12.3-2) . l l
Clad Release Fraction for iodines = 0.02 (Table 4.1, NUREG-1228) ,
i l The activity of each isotope in the clad would then be:
I-131 = 8.64E7(0.02) = 1.73E6 Ci I-132 = 1.29E8(0.02) = 2.58E6 Ci I-133 = 1.99E8(0.02) = 3.98E6 Ci 1-134 = 2.32E8(0.02) = 4.64E6 Ci i i I-135 = 1.81E8(0.02) = 3.62E6 Ci l :
These activities are equivalent to 2.89E6 Ci DEI-131 i
~ + ~ ~
DF1- 131 =
4.93E- 3 i
i j O I EAL - 3.1.3 Rev.00 Page 2 of .3
l HCGS EAURALTechnical Basis Calculating the equivalent concentration: -
O Conc =
2.89E6 Ci(lE6pCi / Ci)
= 7.94E3 Ci/cc 13000 cf(2.8E4 cc / cf) which represents the 100% clad damage concentration.
300 Ci/cc DEI-131 is then equivalent to: _.
300 pCi / cc
= 3.78%
7.94E3 pCi / cc This is rounded to 3.8%.
DEVIATION s.
None l
l REFERENCES
, NUMARC NESP-007, FCI HC.OP-AB.ZZ-0100(Q), High Reactor Coolant Activity l
f HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation 1 HCGS Technical Specification LCO 3.4.5 NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power l
Plant Accidents, Table 4.1 l Reg. Guide 1.109, Table E-9 ;
HCGS-UFSAR, Table 12.2-135 and Table 12.3-2 !
4 i
O EAL - 3.1.3 4
Rev.00
- Page 3 or 3 1
. HCGS EAL/RALTechnical Basis 3.0. Fission Product Barriers (qJ 3.1 Fuel Clad Barrier 3.1.4 IC Potential less or Loss of Fuel Clad EAL _
AMX condition, in the opinion of the EC, that indicates a Potential Imss (3 pts) or Loss (4 pts) of the Fuel Clad Barrier ,
i OPERATIONAL CONDITION - 1, 2, 3 BASIS l
This EAL allows the Emergency Coordinator to address any condition that effects the integrity of the Fuel Clad Barrier that is not already covered elsewhere in the Fission Product 1 Barrier Table. A complete loss of the ability to monitor the Fuel Clad Barrier should be
' considered as a " Potential Loss" of that barrier.
J Barrier Analysis Fuel Clad Barrier has been potentially lost or lost.
ESCALATION CRITERIA Emergency Classification will escalate based on the potential loss or loss of additional Fission Product Barriers per EAL Section 3.0.
DISCUSSION None DEVIATION None REFERENCES NUMARC NESP-007, FC5 ;
EAL - 3.1.4 l Rev.00 l Page 1 of 1
I HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers p
V 3.2 RCS Barrier 4
)
i 3.2.1.a t
I
- EAL _
l 1
Reactor Water Ixvel REACliES -129", EXCLUDIN_Q intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONDITION - 1,2,3 l
BASIS Reactor Water Izvel reaching -129", excluding intentional lowering of Reactor Water Level during an ATWS, indicates th.'t the inventory loss from the RCS exceeds the capacity of '
f available High Pressure injection sources. Below this threshold, a challenge to maintaining
- \ Adequate Core Cooling by core submergence exists, based on Reactor Water Level continuing to decream, thus a Potential Loss of the RCS Barrier exists. Without core submergence, the integrity of the Fuel Clad would be in jeopardy. Appropriate classification under this EAL is
- based on reachine Reactor Water Level of-129" (instead of being able to restore and maintain above -129"), due to the challenge that exist to core submergence. Reactor Water Level l
reaching this threshold results from either a LOCA exceeding available makeup capacity or a Total Loss of High Pressure injection capability.
In addition, during an Anticipated Transient Without Scram (ATWS), it is possible that operator action will be taken to intentionally lower Reactor Water Level to below -129". for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Section 5.0.
Barrier Analysis RCS Barrier has been potentially lost.
EAL - 3.2.1.a Rev.00 Page 1 of 3
HCGS EAL/RALTechnical Basis ESCALATION CRITERIA -
! O _
Emergency Classification will escalate based on the Potential Loss or less of additional Fission Product Barriers per EAL Section 3.0.
i DISCUSSION L
Core Submergence is the preferred method of maintaining adequate core cooling. When
! Reactor Water level decreases to -129", a significant challenge to continued core submergence exists. The threshold for this EAL corresponds to the initiation setpoint I for the low pressure Emergency Core Cooling Systems (ECCS).
L Reactor Water level reaching -129" occurs when the rate of inventory loss is greater than the rate of inventory from High Pressure injection sources. This condition can occur as the result of the following events / sequences (excluding intentional lowering of Reactor Water level during an ATWS).
A LOCA will cause Reactor Water level to reach -129" when the LOCA is the result of a large, break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leak rate.
f\ A less of High Pressure injection sources without the presence of.a LOCA will also result in Reactor Water Level decreasing to -129" , due to continued Reactor Steam
- Flow without makeup.
Either of these events / sequences results in'a potential challenge to the RCS Barrier when Reactor Water level reaches -129", hence classification at this threshold is appropriate.
However, for both these sequences, low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shutoff head of the pumps. Reactor Depressurizatica will occur either due to the LOCA or Manual initiation of Emergency Depressuization when Reactor Water Level reaches -161", provided injection systems are available, This will allow for restoration of Reactor Water Level and re-establishment of Core Submergence.
DEVIATION None D
L lO EAL - 3.2.1.a Rev.00 Page 2 of 3 J
HCGS EAIJRALTechnical Basis REFERENCES -
[
NUMARC NESP-0007, RC5 HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB ZZ-0200 (Q), Reactor Low Water Level HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HCGS Technical Specifications LCO 3/4.3, Instrumentation '
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HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers -
G O 3.2 RCS Barrier 3.2.1.b IC Loss of RCS EAL Reactor Water Level REACHES -161" (Top of Active Fuel), EXCLUDING intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Water Level reaching -161" (Top of Active Fuel - TAF), excluding intentional lowering of Reactor Water Level during an ATWS, results in an inability to maintain adequate core cooling by core submergence, causing a Loss of the RCS Barrier. Without core s
submergence, the integrity of the fuel clad barrier is in jeopardy. Appropriate classification under this EAL is based on reachine Reactor Water Level of -161" (instead of being able to restore and maintain above -161") due to the potentially severe consequences of a loss of core submergence. Reactor Water Level reaching this threshold results from either a LOCA ;
exceeding available makeup capacity or a Total Loss of High Pressure injection capability. ;
In addition, during an Anticipated Transient Without Scram (ATWS), it is possible that operator actions will be taken to intentionally lower Reactor Water Level to between -161" and -190", for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Section 5.0 Barrier Analysis
- RCS Barrier has been lost.
! ESCALATION CRITERIA l
Emergency Classification will escalate based upon the Potential Loss or Loss of additional
- Fission Product Barriers per EAL Section 3.0.
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EAL - 3.2.1.b f'
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HCGS EAL/RALTechnical Basis DISCUSSION .
Core Submergence is the preferred method of maintaining adequate core cooling. When Reactor Water 12 vel decreases to below TAF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad barrier can no longer be considered intact. While the Emergency Operating Procedures provide contingencies to establish adequate core cooling when Reactor Water Level drops below TAF (Steam Cooling with or without injection), these ,
actions are designed to be an alternative method of providing adequate core cooling while actions are taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel clad damage and a significant release of fission products to the Reactor coolant. i Sustained core uncovery can also result in a breach of the Reactor Vessel due to core melt material interaction with the RPV.
A IAss of Core Submergence will occur when the rate of inventory loss is greater than the rate of inventory makeup from High Pressure injection sources. This condition can occur as the result of the following events / sequences (excluding intentional lowering of Reactor Water I2 vel during an ATWS).
A LOCA will cause Reactor Water Level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leak rate.
A Loss of High Pressure injection sources without the pxsence of a LOCA will also result in Reactor Water Ixvel decreasing to TAF, due to continued Reactor Steam Flow without makeup.
Either of these events / sequences results in a challenge to the Fuel Clad Barrier when Reactor Water Ixvel reaches TAF due to core uncovery, hence classification at this threshold is appropriate. However, for both these sequences, Low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shutoff head of the pumps. Reactor Depressurization will occur either due to the LOCA or Manual initiation of Emergency Depressurization when Reactor Water Level reaches -161", provided injection systems are available. This will allow for restoration of Reactor Water Level and re-establishment of Core Submergence.
DEVIATION l
None ;
EAL - 3.2.1.b Rev. 00 Page 2 of 3
HCGS EAURALTechnical Basis REFERENCES ,
NUMARC NESP-0007, RC4 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP-EO.ZZ-0207 (Q)-FC, Level / Power Control BWR Owner's Group Emergency Procedure Guidelines, Rev. 4 F
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l HCGS EAL/RALTechnical Basis
! 3.0 Fission Product Barriers .
!O lV 3.2 RCS Barrier ,
l 3.2.2.a l-IC Potential Loss of RCS i EAL Unisolable RCS Leak Rate .;>_ 50 GPM INSIDE Primary Containment t l OPERATIONAL CONDITION - 1, 2, 3 BASIS 1
i Unisolable RCS Ink Rate exceeding 50 GPM, inside Primary Containment is indicative of a potential loss of the RCS. An unisolable leak rate of this magnitude is significant due to the potential for further break propagation, resulting in a much higher loss of inventory with an i I
inability to isolate the leak source. As such, this threshold is considered a Potential Iess of the RCS. Leakage just above the 50 GPM threshold is well within the capacity of normal and )
emergency injection systems and is not a significant concern for core uncovery. However,50 GPM is the minimum leak rate that would be classified under this EAL, with the maximum i being equivalent to the leak rate that would result in either Reactor Water Level reaching -129" or Drywell Pressure reaching 1.68 PSIG, since these two conditions are obviously more recognizable to Control Room personnel, than an existing leak rate.
Specifying an unisolable RCS leak as part of the threshold for this EAL, precludes classifying events such as an isolable Reactor Recirculation Pump dual seal failure under this EAL.
I Barrier Analysis RCS Barrier has been potentially lost.
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! ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0.
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, O EAL - 3.2.2.a i Rev. 00 i Page 1 of 2 4
HCGS EAL/RALTechnical Basis DISCUSSION O It is important to recognize that the unisolable RCS leak rate established in this EAL is inside the Primary Containment. The inability to isolate the leak would eventually lead to a High Drywell Pressure (> 1.68 PSIG) actuation of RPS, ECCS and PCIS. The actuation would lead to an isolation of the Drywell Floor and Equipment Drain sumps, complicating efforts to further identify and quantify any changes in the existing leak rate. In addition, monitoring of the leak rate could be limited by reaching the upper range (50 GPM) of the Drywell Leak Detection channels (9AX313 - Equipment, 9AX314- Floor Drain).
For leakage outside Containment, since quantification of the leak rate is much more difficult due to the physical size of the Reactor Building, receipt of a Valid isolation signal has been established as the threshold for classification of this type ofleakage.
t DEVIATION
- None i REFERENCES NUMARC NESP-007, RC1 NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #11" O HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0201(Q), Drywell High Pressure / Loss of Drywell Cooling HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response HC.OP-EO.ZZ-0100(Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101(Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102(Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103(Q)-FC, Secondary Containment Control HC.OP-GP.ZZ-0005(Q), Drywell Leakage Source Detection O EAL - 3.2.2.a Rev.00 Page 2 of 2
HCGS EAL/RALTechnical Basis 3.0. Fission Product Barriers 3.2 RCS Barrier 3.2.2.b IC loss of RCS EAL _
Valid High Drywell Pressure Condition ( .;>_ I.68 psig)
OPERATIONAL CONDITION - 1, 2, 3 BASIS A Valid High Drywell Pressure Condition (.2.1.68 PSIG) is indicative of the release of high energy Reactor Coolant from the RCS into the Drywell and hence is considered a less of the RCS Barrier. Valid is defined as the High Drywell Pressure condition soecifically due to RCS leakage into the Drywell, ensuring that event classification under this EAL is truly reflective of a degraded RCS Barrier. This precludes unwarranted event declaration as the result of system i malfunc+. ions, including a loss of Drywell Cooling or inadvertent Drywell makeup. Indication of an RCS leak should be positively determined by observing Primary Containment l
l parameters, including Drywell Pressure and Temperatare trends, Drywell Equipment and
! Floor Drain sump levels, DAPA Radiation levels, atmospheric pressure, Torus Pressure, and the status of Drywell Cooling systems.
An isolable Reactor Recirculation Pump dual seal failure should not result in Drywell Pressure reaching the threshold for this EAL, hence classification under this EAL should not occur.
Barrier Analysis RCS Barrier has been lost.
ESCALATION CRITERIA 4
! Emergency Classification will escalate based on the Potential Imss or Loss of additional Fission Product Barriers per EAL Section 3.0.
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EAL - 3.2.2.b Rev.00 Page 1 of 2
HCGS EAL/RALTechnical Basis DISCUSSION O RCS Ixakage into the Drywell exceeding 50 GPM is substantially greater than the RCS leakage thresholds established in EAL Section 2.1.1, and represents further degradation of the RCS barrier. Inability to isolate the RCS leakage would eventually result in a High Drywell Pressure (> 1.68 PSIG) actuation of RPS, ECCS and PCIS. The actuation would lead to an isolation of the Drywell Floor and Equipment Drain sumps, complicating efforts to further identify and quantify any changes in the leak rate. In addition, monitoring of the leak rate could be limited by reaching the upper range (50 GPM) of the Drywell 12ak Detection channels (9AX313 - Equipment, 9AX314 - Floor Drain).
There are multiple Control Room indicators and alarms which can be used to determine the presence of a High Drywell Pressure condition. Overhead Annunciators will alarm at 1.5 PSIG and 1.68 PSIG. Plant automatic response to a High Drywell Pressure condition includes: a reactor scram, ECCS initiation, trip of the drywell cooling fans and isolation of the cooling water to the drywell. These actuations inay mask the trend in drywell pressure.
For example, the scram will result in less heat being added to the containment and the cooling water isolation will result in no heat being removed.
Actions initiated as part of increasing drywell pressure condition include investigation of the source of the increased leakage into the drywell. maximizing drywell cooling and venting the Drywell (if release criteria can be satisfied). These actions are designed to control and relieve increasing drywell pressure.
DEVIATION None REFERENCES NUMARC NESP-0007, RC2 NUMARC Questions and Answers, June 1993, " Fission Product Barrier Question #11" HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram l HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control l .HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-GP.ZZ-0005 (Q), Drywell Leak Source Detection Hope Creek Appendix A based on NEDO-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications LCO 3/4.3, Instrumentation O EAL - 3.2.2.b Rev.00 Page 2 of 2
, HCGS EAL/RALTechnical Basis l
,_s 3.0 Fission Product Barriers -
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3.2 RCS Barrier 3.2.3.a IC Potential Loss of RCS EAL l Main Steam Line Break OUTSIDE Primary Containment , resulting in an AUTOMATIC MSIV Isolation Signal AND ALL 4 Main Steam Lines have been successfully isolated based on HQ indication of CONTINUING FLOW / LEAKAGE OUTSIDE the
- Primary Containment AFTER valve closure from the Main Control Room has been attempted
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V OPERATIONAL CONDITION - 1, 2, 3 l i BASIS ,
A Main Steam Line Break outside the Primary Containment, resulting in an automatic MSIV Isolation Signal, could result in dose consequences offsite from a " puff" release in excess of 10 l millirem, based on design basis accident analysis, even if MSIV closure occurs within design limits. Hence this condition is classified as a Potential Loss of the RCS Barrier. Classification under this EAL is specifically for a Main Steam Line Break outside the Primary Containment, as evidenced by a rapid change in Main Steam Line Flow and Steam Tunnel Temperature, that results in automatic isolation with no indication of continuing leakage. Valve Packinn leaks that result in elevated Steam Tunnel temperatures do not require classification under this EAL.
A manual actuation of NSSSS or manual MSIV closure PRIOR to exceeding the setpoints that would result in ar automatic isolation of the MSlV should not result in a " puff" release exceeding 10 millirem, and thus should not be classified under this EAL. Verification that continuing leakage does not exist, ensures that any potential release will not significantly exceed the 10 CFR100 limits. This EAL is specific to a break outside the Primary
- Containment, since a break outside represents a potential challenge to Primary Containment l Integrity due to the Containment Bypass condition that would exist until MSIV closure
- occurred . Failure to completely isolate the effected Main Steam Line(s) as determined by i %)
- EAL - 3.2.3.a Rev.00 Page 1 of 3 1
HCGS EAL/RALTechnical Basis valve position and indication of continuing leakage would result in an additional Loss of the l i Primary Containment Barrier.
G In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any Main Steam Line that did not completely isolate. Valve closure is defined as the closure of AN_y valve from the Main Control Room associated with the effected Main Steam Line(s),
that did not completely isolate. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the effected Main Steam Line(s), then event classification under this EAL is warranted due to the consequences of the event previously discussed. This includes Motor Operated Valves tt12t are not controlled by the isolation logic, but are manually controlled from the Main Control Room. (i.e. Main Steam Stop Valves I ABHV-3631 A/B/C/D). In the event the effected Main Steam Line(s) can not be isolated, escalation of the classification will be required.
Barrier Analysis RCS Barrier has been potentially lost ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional barriers per EAL section 3.0.
U DISCUSSION The Main Steam System is associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, Appendix A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment and are not open or potentially open to the environment.
These systems represent an extension of the RCS Barrier beyond the Primary Containment.
Positive identification of a Main Steam Line Break outside the Primary Containment can be based on receipt of the following Overhead Annunciators:
NSSSS ISLN SIG - STM TNL TEMP HI (C8-C4)
NSSSS ISLN SIG - MN STM FLOW HI (C8-B4) j MSIV CLOSURE (C5-B3) as well as the following indications:
MSIV TRIP LOGIC TRIPPED Rapid changes in Main Steam Line Flow and Steam Tunnel Temperatures pd EAL - 3.2.3.a Rev.00 Page 2 of 3
HCGS EAL/RALTechnical Basis DEVIATION This EAL is being maintained in the Fission Product Barrier Table for ease of use by the operators. It has been categorized as a " Potential loss" since the RCS leak is successfully isolated and an alert classification will still be made as a result of the potential loss of RCS.
REFERENCES l
NUMARC NESP-007, RC1 NUMARC Question and Answer, June 1983, " Fission Product Barrier- BWR" Question-J4 ;
10 CFR50, App. A, GDC 55 10 CFR 100 HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0114(Q), IAss of Primary Containment Integrity HC.OP-AB.ZZ-Oll6(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-0011(Q), Annunciator Response Procedures, Window C6 l
HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control l
HC.OP-EO.ZZ-0102 (Q)-FC. Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control l
HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications, LCO 3/4.3
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. HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers -
3.2 RCS Barrier 3.2.3.b IC less of RCS EAL _
RCS Line Break OUTSIDE Primary Containment, resulting in a Valid Isolation Signal for ANY one of the following systems:
- NSSSS
- RCIC AND '
Indication of CONTINUING FLOW / LEAKAGE OUTSIDE the Primary Containment A through the effected system AFTER valve closure from the Main Control Room
,V has been attempted i l
OPERATIONAL CONDITION - 1, 2, 3 l BASIS I l
l An RCS Line Break outside Primary Containment that results in a Valid Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment Isolation valves to maintain RCS and Primary Containment integrity under abnormal
, conditions. A failure of these isolation valves to isolate directly allows Reactor Coolant to be released outside the Primary Containment (Containment Bypass), resulting in a Loss of RCS and Loss of Containment. An RCS Line is AEr line that communicates directly with the Reactor. An RCS Line Break with indication of continuing flow is classified under this EAL, due to the continuing discharge of Reactor Coolant outside the Primary Containment along l
with a potential for further " break propagation". This is the only condition that warrants
- classification under this EAL.
P Valid is defined as the isolation signal snecifically being the result of an RCS Line Break,
- thus ensuring that the RCS discharge is of significant magnitude to pose a threat to the
- pI integrity of the RCS Barrier. This precludes unwarranted Event Classification as the result of
\v condition that result in limited leakage with no potential for " break propagation", including EAL - 3.2.3.b Rev.00 Page 1 of 3 2
HCGS EAL/RALTechnical Basis valve packing leaks outside Primary Containment and RWCU Pumo Seal Leaks. In addition, isolation signal generated from known failures in other systems, that do not result in Reactor l Coolant discharging outside the Primary Containment do not warrant Event Classification under this EAL either. Examples of such failures include a high temperature isolation resulting )
from a loss of ventilation or cooling water, spurious actuation during I&C surveillance testing {
l or a low Reactor Water Level Condition due to a bss of High Pressure injection capability. :
1 In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any l systems that did not completely isolate, prior to event classification. Valve closure is defined l as the closure of ANX valve from the Main Control Room in the system (s) that did not~
completely isolate. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the effected system, then classification under this EAL is not warranted. This includes Motor Operated Valves that are not control by the isolation logic, but are manually controlled from the Main Control Room. Effected system is defined as the system that is providing the flowpath outside the Primary Containment.
Barrier Analysis RCS Barrier has been lost ESCALATION CRITERIA (O
/
Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0.
DISCUSSION 1
NSSSS isolations, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, Appendix A , General Design Criteria
- 55. These systems form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an i extension of the RCS boundary beyond the Primary Containment, and a potential release path )
from the RCS to the environment. Without a completed isolation, continuing flow / leakage l represents a situation where Reactor Coolant is discharging outside the Primary Containment, including areas in the Reactor Building addressed in the EOPs.
Indication of continuing flow / leakage includes: flow indication through isolated lines,
, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room ;
levels in spaces associated with affected lines, as well as increases in Plant Vent Effluent
{
levels.
4 lO EAL - 3.2.3.b Rev. 00 Page 2 of 3 4
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HCGS EAL/RALTechnical Basis DEVIATION b I This EAL is being considered a loss of the reactor coolant boundary since actuation oflisted isolation system indicate a leak of significant magnitude, and an isolation failure. The classification for exceeding this EAL remains consistent with NUM ARC guide lines.
REFERENCES NUMARC NESP-007, RC1 10 CFR50, App. A, GDC 55 10 CFR 100 HC.OP-SO.SM-0001(Q), Isolation Systems Operation HC.OP-AB.ZZ-0114(Q), Loss of Primary Containment Integrity HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-0011(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control i HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications LCO 3/4.3, Instrumentation HCGS UFSAR, Section 6.2.4.3.1 L
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. HCGS EAL/RALTechnical Basis I
g 3.0. Fission Product Barriers U
3.2 RCS Barrier 3.2.4 IC Potential Loss or Loss of RCS EAL _
AMX condition, in the opinion of the EC, that indicates a Potential Loss (3 pts) or Loss (4 pts) of the RCS Barrier 1 OPERATIONAL CONDITION - 1, 2, 3 BASIS l
This EAL allows the Emergency Coordinator to address any condition that effects the integrity of the RCS Barrier that is not already covered elsewhere in the Fission Product Barrier Table.
A complete loss of the ability to monitor the RCS barrier should be considered as a " Potential
{x Loss" of that barrier.
Barrier Analysis RCS Barrier has been potentially lost or lost.
ESCALATION CRITERIA Emergency Classification will be escalate based on the Potential Loss or Loss of additional i barriers per EAL section 3.0.
DISCUSSION None DEVIATION None REFERENCES NUMARC NESP-007, RC6 EAL - 3.2.4 I I
Rev.00
' Page 1 of 1 l
. HCGS EAL/RALTechnical Basis 3.0. Fission Product Barriers q'J 3.3 Containment Barrier 3.3.1 IC Potential Loss of Containment l
~
EAL 1
Reactor Water Level CANNOT BE RESTORED AND MAINTAINED above -200" (Minimum Zero Injection RPV Water Level)
OPERATIONAL CONDITION - 1, 2, 3 BASIS i l
Inability to restore and maintain Reactor Water Level above - 200" (Minimum Zero Injection RPV Water Level), results in a loss of adequate core cooling by all mechanisms, causing a O
O Potential Imss of the Fuel Clad Barrier. Without adequate core cooling, the integrity of the Containment is being challenged and can no longer be assured. Appropriate classification under this EAL is based on the failure of injection systems to restore and_ maintain Reactor i Water Level above -200", following a condition that causes level to decrease below the threshold. For example, a large break LOCA is expected to cause Reactor Water I.evel to i momentarily decrease below -200", due to the response time of Low Pressure ECCS. As these systems initiate and commence injection to the Reactor, water level will begin to increase and l
should be able to be maintained above -200". In this case, classification under this EAL is not appropriate as plant systems have performed their intended design function and will eventually j restore adequate core cooling by core submergence. However, in the event that Ixw Pressure j ECCS and alternate injection system, as def' m ed in the EOPs are in a degraded condition (i.e.,
Station Blackout, ECCS Suction Strainer plugging, etc.) and Reactor Water Level can not be restored and maintained above -200", then classification under this EAL should occur due to
! the Potential Loss of Containment from the release of energy to the containment from
! imminent fuel failure.
i Barrier Analysis Primary Containment Barrier has been potentially lost. )
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EAL - 3.3.1 Rev.00 i Page 1 of 3 4
HCGS EAL/RALTechnical Basis
, ESCALATION CRITERIA -
U Emergency Classification will escalate based upon the Potential Loss or Loss of addWonal Fission Product Barriers per EAL Section 3.0.
DISCUSSION Core submergence is the preferred method for maintaining adequate core cooling. The failure to reestablish Reactor Water Level above -161", the Top of Active Fuel (TAF), for an extended period of time could lead to significant fuel damage. With Reactor Water Level below TAF, but above the Minimum Zero Injection RPV Water I2 vel (-200"), adequate core cooling occurs due to the cooling effects of steam generated in the covered portion of the core flowing through the uncovered portion (Steam Cooling). The Minimum Zero Injection RPV Water Level is defined in the Emergency Operating Procedures. This method of cooling precludes any fuel clad temperature in the uncovered portion of the core from exceeding 1800*F. As Reactor Water Level drops below -200" with no injection available, this method of cooling becomes inc.dequate. Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of energy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time, Reactor Water Level l remaining below TAF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public i- protective actions.
(J')
Ample time should be provided for Low Pressure ECCS and alternate injection systems restore Reactor Water Level prior to entry into this classification. The time basis for deciding whether or not Reactor Water can be maintained > -200" should be based on the rate of reactor depressurization, the availability of low pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as Reactor Water Level trend, injection flow rates, containment parameter trends, and low ,
pressure injection system operability should also be considered.
In the event, Reactor Water Level can not be restored > -200", containment flooding will be required by the EOPs. This will attempt to flood the containment as a means of flooding the RPV, and use a flooded containment as a heat sink for the nuclear fuel.
DEVIATION None D
(V EAL - 3.3.1 Rev.00 Page 2 of 3
[
HCGS EAL/RALTechnical Basis REFERENCES O NUMARC NESP-0007, PC4 l HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram l HC.OP.EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP.EO.ZZ-0201 (Q)-FC, Alternate Level Control '
HC.OP.EO.ZZ-208 (Q)-FC, Primary Containment Flooding BWR Owners Group Emergency Procedure Guidelines, Revision 4 O
I O EAL - 3.3.1 Rev.00 Page 3 of 3
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. HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers -
l 3.3 Containment Barrier 3.3.2.a IC Potential Loss of Containment EAL _
Supp Chamber press CANNOT BE MAINTAINED below 65 psig OR Primary Containment 112 concentration >4% and O2 concentration >5%
OPERATIONAL CONDITION - 1, 2, 3 BASIS Contain' ment venting required by the EOPs indicates a degrading condition in containment and is implemented in an effort to preclude containment failure. Venting is required before Suppression Chamber pressure reaches 65 PSIG or Hydrogen concentration reaches the Lower Explosive Limit (LEL = 4%) and Oxygen concentration reaches 5%. Exceeding these parameters creates the potential for an unisolable breach of the primary containment, which could result in an uncontrolled, unmonitored, and untreated release of radioactivity to the l environment. This EAL represents a Potential Loss of Containment, since containment i
! venting is required due to Containment parameters potentially exceeding their design limits. j l
The magnitude of any radiological release is dependent upon events leading to the requirement i l for emergency venting, including a loss of the RCS and a loss of the Fuel Clad Barriers.
A Downcomer failure, by itself, does not represent a Loss of the Primary Containment l
Barrier. This failure does, however, render the Primary Containment inoperable per the l Technical Specification, as Primary Containment integrity has been compromised. A Downcomer failure combined with a large break LOCA will likely result in a Potential Loss of l Primary Containment under this EAL if Containment pressure can not be maintained below 65 PSIG and Containment Venting is required.
l Barrier Analysis
- Primary Containment Barrier has been potentially lost.
l EAL - 3.3.2.a Rev.00 Page 1 or 2 l
1
.- HCGS EAL/RALTechnical Basis l
i ESCALATION CRITERIA -
O Emergency Classification will escalate based on the Potential Loss or Loss of additional l Fission Product Barriers per EAL Section 3.0.
DISCUSSION Venting of the Primary Containment is initiated to preserve containment integrity under l- accident conditions. Primary Containment venting is required when Suppression Chamber cannot be maintained below 65 psig, which is well above the maximum pressure expected to l
be present in the Primary Containment during a design basis Loss of Coolant iccident j
( (LOCA). Primary Containment venting is also required based on hydrogen concentrations ;
exceeding 4%. H concentrations 2 in excess of 6.0 % requires Emergency Depressurization' - ;
and subsequent containment venting. Venting is continued until either H2concentration has j been reduced to <6.0% or O 2levels have been reduced to <5.0%. Venting with elevated . l hydrogen concentration conditions ensures that containment failure resulting from a hydrogen detonation or deflagration does not occur.
The elevated hydrogen in the containment may result from excessive zircaloy-water reaction l occurring following a LOCA. Additionally, hydrogen and oxygen gas may be introduced into
!' the containment environment from long term disassociation of water in the Suppression Chamber.
EOP procedural guidance in these cases is provided to vent the Primary Containment
(. regardless of off-site dose consequences. Although radiological releases resulting from venting l - containment may exceed EPA limits, a controlled, monitored, and isolable release is preferred l to a potential uncontrolled, unmonitored radiological release that would result from a failure of l containment.
i L DEVIATION None REFERENCES NUMARC NESP-0007, PCl, PC2 HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control
)
HC.OP-EO.ZZ-0318 (Q)-FC, Containment Venting BWR Owners Group Emergency Procedure Guidelines, Revision 4 f- l i
1 Rev. 00 I Page 2 of 2 ,
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. HCGS EAL/RALTechnical Basis 3.0. Fission Product Barriers 3.3 Containment Barrier l 3.3.2.b IC Loss of Containment EAL P
Containment Failure as indicated by a rapid decrease in Drywell pressure following an increase in pressure above 1.68 psig D.R Containment is Vented by the Emergency Operating Procedures (EOPs) !
OPERATIONAL CONDITION - 1, 2, 3
,G BASIS U
Containment failure indicated by a rapid decrease in Drywell pressure following a significant increase in Drywell pressure is indicative of a Loss of the Containment barrier. This EAL specifically represents a Loss of Containment , whereby a unisolable breach of the ;
containment structure has occurred. Conditions that result in a decrease in Drywell pressure following a pressure rise that are not the direct result of a Containment failure do not warrant :
classification under this EAL. These events include the initiation of Drywell Sprays, the re-establishment of Drywell Cooling, Containment Venting as required by the EOPs, and anticipated Drywell pressure decrease due to ambient losses.
Containment Venting is a controlled loss of containment. This venting is performed for the purpose of preventing an unisolable, unmonitored radiological release of containment gases.
I A Downcomer failure, by itself, does not represent a Loss of the Primary Containment Barrier. This failure does, however, render the Primary Containment inoperable per the Technical Specification, as Primary Containment integrity has been compromised. A Downcomer failure combined with a large break LOCA will likely result in a Potential Loss of Primary Containment under EAL 3.3.2.a if Containment pressure can not be maintained j below 65 PSIG and Containment Venting is required.
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V EAL - 3.3.2.b i Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis Barrier Analysis -
- Primary Containment Barrier has been lost.
ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or loss of additional Fission Product Barriers per EAL Section 3.0.
DISCUSSION -
Appropriate classification under this EAL occurs as the result of a Containment failure.
Drywell pressure reaching 1.68 psig indicates that there is a significant release of reactor coolant to the containment. Unless this source ofleakage is isolated or the Reactor is depressurized, Drywell pressure would not be expected to decrease in a rapid manner.
Other indications such as Reactor Building Area Radiation Monitors (ARMS) radiation levels, Reactor Building area temperatures, Reactor Building floor and sump levels, Plant Effluent radiation levels, and containment isolation status should be used to confirm the loss of I
containment integrity if possible. Reactor Building to Torus vacuum breaker status should be monitored to ensure that this pathway does not result in a loss of containment integrity.
. DEVIATION l None REFERENCES NUMARC NESP-0007, PCI !
HC.OP-AB.ZZ-0114 (Q), Loss of Primary Containment Integrity I HC.OP-AB.ZZ-0116 (Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure / Loss of Drywell Cooling HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control l HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control i HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control I' BWR Owners Group Emergency Procedure Guidelines, Revision 4
/N U EAL - 3.3.2.b Rev.00 Page 2 of 2
- . HCGS EAL/RALTechnical Basisw l l
3.0. Fission Product Barriers O) u 3.3 Containment Barrier 3.3.3 IC Potential IAss of Containment l 1
- EAL _
1 1
DAPA Radiation Monitor reading A 28000 R/hr l l
OPERATIONAL CONDITION - 1, 2, 3 l l
BASIS l
Drywell Atmosphere Post Accident (DAPA) monitor reading 128000 R/hr indicates significant fuel damage, well in excess of the level corresponding to the loss of the RCS and Fuel Clad barriers. This threshold corresponds to approximately 20% fuel clad damage.
Regardless of whether or not containment is challenged, this amount of activity in l J containment, if released, could have severe consequences and it is prudent to treat this 1
- condition as a Potential Loss of containment.
l Barrier Analysis Primary Containment Barrier is potentially lost. ;
l ESCALATION CRITERIA i l Emergency Classification will escalate based on the Potential loss or Loss of additional Fission Product Barriers per EAL Section 3.0. :
DISCUSSION NUREG-1228, " Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", states that releases of severe magnitude are not possible if plant systems l
function as designed, and any accident with a release of 20% or greater of the gap region mi'st j be considered severe.
Using attachment 2 of EPIP 205H,10% clad damage is represented by a DAPA reading of q 1.4E4 R/hr at 0.1 hrs after shutdown (the most conservative). This is shown on the NJ EAL - 3.3.3 l
Rev. 00 Page 1 of 2
I
. HCGS EAURALTechnical Basis
~
attachment as the 1 % TID line. Extrapolating to 20% clad damage gives a reading oi 2.8E4 -
R/hr. -
Exceeding a DAPA reading of 28000 R/hr should meet the criteria for declaration of a General Emergency.
DEVIATION 1
None !
- 1 REFERENCES i NUMARC NESP-007, PC3 NUREG-1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents EPIP 205H, TSC - Post Accident Core Damage Assessment 0
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1 - - - - _ . . _ .
. HCGS EAL/RALTechnical Basis
- 3.0 Fission Product Barriers -
I i j 3.3 Containment Barrier l 3.3.4.a l IC Potential Loss of Containment 1
EAL __
i I
- RCS Line Break OUTSIDE Primary Containment, resulting in a Valid Isolation Signal for ANY one of the following systems
l l
- NSSSS (excluding Main Steam Lines)
,
- HPCI i
!
- RCIC !
l AND l
l EQ indication of CONTINUING FLOW / LEAKAGE OUTSIDE the j Primary Containment through the effected system AFTER valve closure l G from the Main Control Room has been attempted i
OPERATIONAL CONDITION - 1, 2, 3 l
! BASIS 1
i An RCS Line Break outside Primary Containment that results in a Valid Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment
! Isolation valves to maintain RCS and Primary Containment integrity under abnormal
- conditions. Successful closure of the required isolation valves that results in HQ indication of
! continuing FLOW / LEAKAGE is classified under this EAL as an Unusual Event, due to the l significance of an RCS line break outside the Primary Containment for one of the systems listed in the EAL. An RCS Line is ANY line that communicates directly with the Reactor. A l Main Steam Line Break with successful isolation is excluded from this EAL, since it is covered under EAL 3.2.3.a. An RCS Line Break with indication of successful isolation is the enh condition that warrants classification under this EAL.
Valid is defined as the isolation signal snecifically being the result of an RCS Line Break, thus ensuring that the RCS discharge is of significant magnitude to pose a threat to the (q /
integrity of the Primary Containment Barrier. This precludes unwarranted Event Classification as the result of condition that result in limited leakage with no potential for " break EAL - 3.3.4.a Rev.00 Page 1 of 3
. HCGS EAL/RALTechnical Basis propagation", including valve oackine leaks outside Primary Containment and RWCU-Pump Scal Leaks. In addition, isolation signal generated from known failures in other systems, that do not result in Reactor Coolant discharging outside the Primary Containment do not warrant Event Classification under this EAL either. Examples of such failures include a high temperature isolation resulting from a loss of ventilation or cooling water, spurious actuation during I&C surveillance testing or a low Reactor Water Level Condition due to a Loss of High Pressure injection capability. l In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any systems that did not completely isolate, prior to event classification. Valve closure is defined as the closure of ANY valve from the Main Control Room in the system (s) that did not completely isolate. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the e rfected system, then event classification under this EAL is warranted, due to the consequences of the event previously discussed. This includes Mctor Operated Valves that are not control by the j isolation logic, but are manually controlled from the Main Control Room. Effected system is defined as the system that is providing the flowpath outside the Primary Containment. In the event the effected system (s) can not be isolated, escalation of the classification will be required.
Barrier Analysis O Primary Containment Barrier has been potentially lost O
ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0.
DISCUSSION NSSSS isola: ions, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, Appendix A , General Design Criteria
- 55. These systerns form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment.
Indication of continuing flow / leakage includes: flow indication through isolated lines, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room levels in spaces associated with affected lines, as well as increases in Plant Vent Effluent levels.
O EAL - 3.3.4.a Rev. 00 Page 2 of 3
_ _ _ _____ _.. _.__ _ ._ .._ . - _- _ .._ _ _ _ _ m ._ .. .
. HCGS EAL/RALTechnical Basis DEVIATION -
l None REFERENCES NUMARC NESP-007, PC5 10 CFR50, App. A, GDC 55 .
HC.OP-SO.SM-0C '(Q), Isolation Systems Operation -
HC.OP-AB.ZZ-f !QQ), I.oss of Primary Containment Integrity HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation l HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation l HC.OP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-00ll(Q), Annunciator Response Procedures, Window Ce .
HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 ;
HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram "
HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control ,
! HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control !
!. HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications, LCO 3/4.3 HCGS UFSAR, Section 6.2.4.3.1 I
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, HCGS EAURALTechnical Basis ;
l 1
3.0 Fission Product Barriers -
O 3.3 Containment Barrier l 3.3.4 b )
l IC Loss of Containment l
EAL _
Isolation Signal for AN_Y one of the following systems:
l
- NSSSS l
\
AND Indication of CONTINUING FLOW / LEAK AGE OUTSIDE the Primary Containment ;
r'~N through the effected system AFTER valve closure from the Main Control Room has d been attempted OPERATIONAL CONDITION - 1, 2, 3 BASIS l An Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment Isolation valves to maintain RCS and Primary Containment integrity under abnormal conditions. A failure of these isolation valves to isolate directly allows the
(
transport of Reactor Coolant or containment atmosphere to outside the Primary Containment
! (Containment Breach or Bypass), resulting in a Loss of Containment. This EAL addresses two condition under which RCS is being transported OUTSIDE the Primary Containment. The first condition is associated with an Isolation signal being generated as the result of an RCS l
Line Break with a failure of the isolation valves to close. In this condition, a ABNORMAL l FLOWPATH exists for RCS to be discharged directly outside the Primary Containment. The
- second condition is associated with the failure of both Inboard and Outboard Isolation valves to FULLY close following an Isolation signal. In this coaJition, a flow path from containment atmosphere to areas outside of the Primary Containment exists.
O V'
EAL - 3.3.4.b Rev. 00 l Page 1 of 3 l
l
. HCGS EAL/RALTechnical Basis In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any systems that did not completely isolate, prior to event classification. Valve closure is defined as the closure of ANY valve from the Main Control Room in the system (s) that did not completely isolate. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the effected system, then Unusual Event declaration is not warranted. This includes Motor Operated Valves that are not control by the isolation logic, but are manually controlled from the Main Control Room. Effected system is defined as the system that is providing the flowpath outside the Primary Containment.
Barrier Analysis Primary Containment has been lost.
1 ESCALATION CRITERIA Emergency Classification will escalate % sed on the Potential Loss or Loss of additional I Fission Product Barriers per EAL Sem 3.0.
1 DISCUSSION l PCIS Isolations are associated with systems having lines that are either: 1) connect directly to
( the Primary Containment atmosphere and penetrate the Primary Containment; or 2) penetrate k the Primary Containment and are neither part of the RCS boundary and are not connected directly to the Primary Containment atmosphere (e.g. RACS, Chilled Water). Isolation ,
requirements for these lines are covered in 10CFR50, App. A , General Design Criteria 56 l and 57 respectively. This event, therefore, may potentially connect the RCS or the Primary Containment atmosphere to the environment. Without a completed isolation, continuing flow / leakage represents a release path from the RCS or Primary containment to the l
environment. !
1 NSSSS isolations, as well as HPCI and RCIC steam line isolations, are associated with systems i that are part of the RCS boundary and penetrate the Primary Containment. Isolation l requirements for these lines are covered in 10CFR50, App. A , General Design Criteria 55.
These systems form a closed loop apEidg the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment. Without a completed isolation, continuing leakage i represents a Primary System discharging outside the Primary Containment (Containment Bypass), including areas in the Reactor Building addressed in the EOPs.
Indication of continuing flow / leakage includes: flow indication through isolated lines, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room f
\
levels in spaces associated with affected lincs, as well as increases in Plant Vent Effluent levels.
EAL - 3.3.4.b Rev.00 Page 2 of 3
i HCGS EAURALTechnical Basis I
The isolation valve status of all isolation groups is monitored for quick reference on SPDS, to be backed up by operator observation of valve status.
DEVIATION-NUMARC Primary Containment Barrier Example Flowchart (PC2) suggests that for the
" Containment Isolation Valve States after Containment Isolation Signal" EAL, a failure of both valves in any one line to close AND downstream pathway to the environment exists be l included as a threshold for classification of an Unusual Event. In order to include the condition l where the Inboard Valve fails to close and an RCS Line Break exists between the Primary Containment wall and Outboard Valve, the condition that both valves fail to close is NOT being included in the EAL. Indication of continuing flow / leakage OUTSIDE the Primary l Containment will provide an adequate threshold for Event Classification, since both isolation ;
I valves must be open for continuing leakage Outside the Primary Containment, except as noted ;
above.
REFERENCES l
NUMARC NESP-007, PC2 10CFR50, App. A, GDC 55,56,57 10 CFR 100 ;
HC.OP-SO.SM-0001(Q), Isolation Systems Operation :
HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation :
HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response l HC.OP-AR.ZZ-0011(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram l HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control l HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control l HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HCGS Technical Specifications LCO 3/4.3, Instrumentation HCGS UFSAR Sections 6.2.4.3.1, 6.2.4.3.2, 6.2.4.3.3 l
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I
, HCGS EAL/RALTechnical Basis l
l l , 3.0 Fission Product Barriers -
3.3 Containment Barrier i 3.3.5 IC Potential IAss or Loss of Containment Barrier l EAL _
AN_X condition, in the opinion of the EC, that indicates a Potential IAss (1 pt) or Loss (2 pts) of the Containment Barrier OPERATIONAL CONDITION - 1, 2, 3 BASIS I
This EAL allows the Emergency Coordinator to address any condition that effects the integrity j of the Containment Barrier that is not already covered elsewhere in the Fission Product Barrier l
/~N Table.
U A complete loss of the ability to monitor the Containment Barrier should be considered as a "PotentialIess" of that barrier.
Barrier Analysis Containment Barrier has been potentially lost or lost.
ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional barriers per EAL section 3.0.
DISCUSSION None DEVIATION None I REFERENCES NUMARC NESP-007. PC6 EAL - 3.3.5 Rev.00 Page 1 of 1
. HCGS EAL/RALTechnical Basis 4.0 EC Discretion -
O 4.1 Emergency Coordinator Discretion '
UNUSUAL EVENT - 4.1.1 IC Other Conditions Exist Which In the Judgement of the Emergency Coordinator Warrant Declaration of an Unusual Event
~
EAL Ever.ts are in progress or have occurred which, in the judgement of the Emergency l
Coordinator, indicate a Potential Degradation of Plant Safety i l
i l
OPERATIONAL CONDITION - All
)
BASIS )
l Emergency Coordinator judgement to declare an Unusual Event , based on the determination that the Potential Degradation of Plant Safety exists, should be implemented ONLY when g 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 follov ing examples are by no means all inclusive and are not intended 1 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 judgement for Fission Product Barriers is found on the l Fission Product Barrier Table, Section 3.0.
ESCALATION CRITERIA Emergency Coordinator Judgement EAL - 4.1.1 Rev.00 l Page 1 of 2
1
. HCGS EAURALTechnical Basis DISCUSSION -
O 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 < 10 mrem TEDE. Refer to Section 6 of the ECG if a Radiological release is ongoing.
DEVIATION 1
None REFERENCES <
NUMARC NESP-0007, HUI.3, HU5, Section 3.7. .
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.- HCGS EAL/RALTechnical Basis l
4.0 EC Discretion -
O 4.1 Emergency Coordinator Discretion 1 ALERT - 4.1.2 IC Other Conditions Exist Which In the Judgement of the Emergency Coordinator Warrant Declaration of an Alert EAL Events are in progress or have occurred which, in the judgement of the Emergency Coordinator, indicate EITHER one of the following:
Plant safety systems (more than one) are, or may be degraded ANY Plant Vital Structure is degraded or potentially degraded AND Increased monitoring of Safety Functions is warranted OPERATIONAL CONDITION - All BASIS Emergency Coordinatorjudgement to declare an Alert, based on the determination that Plant Systems are, or may be degraded, should be implemented ONLY when conditions are not i 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 plant functions is warranted to better determine the plants actual safety status than an Alert classification may be appropriate.
Barrier Analysis Additional guidance on EC judgement for Fission Product Barriers is found on the Fission Product Barrier Table, Section 3.0.
ESCALATION CRITERIA Emergency Coordinator Judgement DISCUSSION EAL - 4.1.2 l Rev.00 l Page 1 of 2
- HCGS EAURALTechnical Basis Dose consequences for an Alert, if a radiological release was ongoing, would only be a small fraction of the EPA Protective action Guideline (PAG) plume exposure level, i.e.,10 to 100 mrem TEDE. Refer to ECG Section 6 if a radiological release is ongoing.
1 l DEVIATION None ,
REFERENCES NUMARC NESP-0007, HA6, HAl.4, Section 3.7.
! EPA-400 r t l
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( . HCGS EAL/RALTechnical Basis 4.0 EC Discretion -
l 4.1 Emergency Coordinator Discretion
- SITE AREA EMERGENCY - 4.1.3 IC Other Conditions Exist Which in the Judgement of the Emergency Coordinator Warrant Declaration of a Site Area Emergency
$ EAL Events are in progress or have occurred which, in the judgement of the Emergency Coordinator, indicate an Actual or likely major failure of plant functions needed for the protection of the public OPERATIONAL CONDITION - All BASIS i
4 73 Emergency Coordinator judgement to declare a Site Area Emergency, based on the V
determination that the potential exists for an uncontrolled radiological release or the source ,
term available in the Containment atmosphere could result in Site Boundary dose rates in excess of 100 mrem /hr, should be implemented ONLY when conditions are not explicitly I 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 judgement for Fission Product Barriers is found on the
. Fission Product Barrier Table, Section 3.
ESCALATION CRITERIA s Emergency Coordinator Judgement l
- DISCUSSION l' 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
'(n) within the Site Boundary. However, plume exposure levels of 100 to < 1000 mrem TEDE EAL - 4.1.3 Rev.00 Page 1 of 2
1
)
)
.- HCGS EAURALTechnical Basis may be possible offsite and levels > 1000 mrem TEDE could be experienced onsite. Refer to ECG Section 6 if a radiological release is ongoing.
DEVIATION None REFERENCES l NUMARC NESP-0007, HS3, Section 3.7. -
l EPA-400 l
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. HCGS EAL/RALTechnical Basis 4.0 EC Discretion 4.1 Emergency Coordinator Discretion GENERAL EMERGENCY - 4.1.4 IC Other Conditions Exist Which In the Judgement of the Emergency Coordinator Warrant Declaration of a General Emergency
~
EAL Events are in progress or have occurred which, in the judgement of the Emergency Coordinator, indicate an Actual or imminent substantial core degradation with the potential for loss of containment i
OPERATIONAL CONDITION - All BASIS p Emergency Coordinator judgement to declare a General Emergency , based on the I,d determination that the potential for an uncontrolled radionuclide 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, ar.d permits maximum offsite intervention time.
Barrier Analysis This EAL is intended for EC judgement for declaration at the General Emergency level. Additional guidance on EC judgement for Fission Product Barriers is found on the Fission Preduct Barrier Table, Section 3.0.
ESCALATION CRITERIA N/A 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
l 4
. HCGS EAURALTechnical Basis
! DEVIATION -
O None REFERENCES NUMARC NESP-0007, HG2, Section 3.7.
EPA-400 l
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HCGS EAL/RALTechnical Basis 7, 5.0 Failure to Scram -
5.1 ATWS ALERT - 5.1,2.a / 5.1.2.b IC Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram (Automatic or Manual)
EAL 1
l EITHER one of the following conditions: l
- An Automatic Reactor Scram Condition exists AND An Automatic Reactor Scram (RPS)
IS NOT successful e ANY Manually Initiated Reactor Scram (RPS) from the Control Room IS NOT l I
successful OPERATIONAL CONDITION - 1, 2
(]
V l
BASIS Failure of the RPS to successfully complete a Reactor Scram (automatic or manual) represents I
a significant degradation in plant safety, as the primary reactivity control system has failed to perform its design function. The intent of this EAL is to classify events in which either an automatic at manual RPS signal fails to initially complete a successful scram when required, even if a subsequent manual or automatic scram is successful . The failure of RPS to complete a successful scram, is the bases for Alert declaration under this EAL. A Successful scram (RPS automatic or RPS manual), as it relates to this EAL, results in a Control Rod configuration by which the Reactor will remain shutdown under all conditions without boron.
The three criteria that satisfy this condition are :
- 1) All Control Rods are inserted to position 02 or beyond (Maximum Suberitical Banked Withdrawal Position)
- 2) All Control Rods but one being full inserted.
- 3) Reactor Engineering has determined that the Reactor will remairi Shutdown under all conditions without Boron In addition, for a manual scram to be considered successful, it must be attempted from the p Reactor Control console. In the event that ARI completes a successful Scram following a failure of automatic or manual RPS, the declaration of an Alert is still warranted, due to the V
EAL - 5.1.2.a / 5.1.2.b Rev.00 Page 1 of 3
1
, HCGS EAIJRALTechnical Basis l
failure of RPS. An inability to physically place the Reactor Mode Switch in the SHUTDOWN
,)
(V position, (i.e. broken key) does not constitute an RPS failure, since the RPS logic has not failed.
j Barrier Analysis l
This event does not reach the threshold for the loss of Fuel Clad or RCS Barriers, but !
conditions exist that could lead to a potential loss of those barriers. {
J ESCALATION CRITERIA -
Emergency Classification will escalate to a Site Area Emergency (5.1.3) when a failure of both automatic or manual scram functions occurs, with Reactor power remaining .2.4%.
l DISCUSSION The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is " fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical / pneumatic failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure). _
l l
The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System j -
(RRCS) provides an automatic backup function for an electrical / pneumatic failure of the RPS.
i A successful scram due to ARI following a failure of the RPS would still be classified under this EAL because of the potentially serious consequences of an RPS failure.
! Confirmation indications of an RPS failure to complete a successful scram include control l room annunciators, control rod positions, APRM power and downscale indicating lights, ;
IRM/SRM power level, SRM period, and control rod position indication.
A manual scram is defined as any set of actions by the reactor operator (s) at the reactor controls which causes control rods to be rapidly inserted into the core via the RPS in an attempt to place the reactor in a subcritical condition (i.e. mode switch to shutdown, manual j scram push buttons). This EAL addresses only those manual scram attempts that are initiated l I
from the Control Room control panels.
A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram can result in the design limits of the nuclear fuel being compromised. RPS is designed to l
automatically detect and generate a reactor scram signal when a Technical Specification l Limiting Safety System Setting (LSSS) is reached or exceeded. If an LSSS is exceeded without an automatic scram, consideration must be given to the possibility that a Technical Specification Safety Limit may have been exceeded.
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V i EAL - 5.1.2.a / 5.1.2.b Rev 00 Page 2 of 3 i
. HCGS EAURALTechnical Basis DEVIATION -
O NUMARC EAL SA2 suggests that an Alert classification be based only on a failure of an automatic RPS scram followed by a successful manual RPS scram from the control room, with EAL SS2 escalating to a Site Area Emergency if a manual scram (RPS or ARI) fails to reduce Reactor Power below 4%.
The Alert threshold is set so that unsuccessful manual RPS scrams from the control room, as well as unsuccessful automatic RPS scrams via RPS would be classified at the Alert level.
This will cover those situations in which a manual RPS scram is attempted in anticipation of a continually degrading plant condition (i.e. degrading Main Condenser Vacuum). In addition, this threshold will also address those situations where a manual scram is required by procedure. (i.e. stuck open SRV, Main Steam Line Hi Hi Radiation, Dual Reactor Recirc Pump trip, Power Oscillations) and the manual scram is not successful. In either case, Alert declaration is appropriate when the RPS fails to perform its intended design function.
The SAE threshold is set to include automatic and manual failure (for the reasons stated above), with resulting power.;>_.4% as suggested in NUMARC EAL SS2 bases.
By defining a " Successful" scram as control rod being positioned such that the Reactor will remain Shutdown under all conditions, partial scrams that result in Reactor Power below 4%
would be classified as an Alert, whether automatically or manually initiated.
REFERENCES NUMARC NESP-0007, SA2 NUMARC Questions and Answers, June 1993, " System Malfunctions Question #7" HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0207 (Q)-FC, Level / Power Control BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Defm' itions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation O EAL - 5.1.2.a / 5.1.2.b Rev.00 Page 3 of 3
HCGS EAURALTechnical Basis 5.0 Failure to Scram -
5.1 ATWS SITE AREA EMERGENCY - 5.1.3 s
IC Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram (Automatic and Manual) and Reactor Power is above than 4%.
~
EAL EITHER one of the following conditions:
- An Automatic Reactor Scram Condition exists AND An Automatic Reactor Scram (RPS)
IS NOT successful e ANY Manually Initiated Reactor Scram (RPS) from the Control Room IS NOi successful AND
- ALL Reactor Scram attempts from the Control Room (RPS and ARI) DID NOT REDUCE and MAINTAIN Reactor Power to.14%
OPERATIONAL CONDITION - 1, 2 -
BASIS Failure of the RPS to successfully complete a Reactor Scram (automatic and manual) represents a significant degradation in plant safety, as the primary reactivity control system has failed to perform its design function. In addition, failure of subsequent Reactor Scram attempts (both RPS and ARI) to reduce Reactor Power to less than 4%, represents a potential challenge to the ability to provide continued heat removal from the Reactor. Thus, conditions exist that could lead to an imminent loss or potential loss of both the Fuel Clad and RCS Barriers. The intent of this EAL is to classify events in which bEh automatic a0d manual RPS signals fail to complete a successful scram when required, and subsequent actions using ARI fails to i reduce Reactor Power to less than 4%. The failure of RPS and ARI to complete a successful scram with Reactor Power remaining above 4% is the bases for SAE declaration under this EAL. A Successful scram (RPS Automatic or Manual), as it relates to this EAL, results in a Control Rod configuration by which the Reactor will remain shutdown under all conditions without boron injection. The three criteria that satisfy this condition are :
w EAL - 5.1.3 ;
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, HCGS EAL/RALTechnical Basis
- 1) All Control Rods are inserted to position 02 or beyond (Maximum Suberitical Banked Withdrawal Position) l
('") 2) All Control Rods but one being full inserted.
- 3) Reactor Engineering has determined that the Reactor will remain Shutdown under all conditions without Boron In addition, for a manual scram to be considered successful, it must be attempted from the Reactor control console. In the event that ARI completes a successful Scram following a failure of automatic or manual RPS, the declaration of an SAE is not warranted.
Barrier Analysis This event does not reach the threshold for the loss of Fuel Clad or RCS Barriers, but conditions exist that could lead to an imminent loss or potential loss of those barriers.
ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency (5.1.4) when Reactor Water Level can not be maintained >-190", or Suppression Pool Temperature and Reactor Pressure can not be maintained below the HCTL.
DISCUSSION l V(}
l The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is " fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical / pneumatic failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure).
! The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System (RRCS) provides an automatic backup function for an electrical / pneumatic failure of the RPS.
! A failure of ARI to reduce Reactor Power to .s_4% following a failure of the RPS is classified under this EAL because of the potentially serious consequences of a failure of RPS and ARI to reduce Reactor Power.
Confirmation indications of an RPS failure to complete a successful scram include control room annunciators, control rod positions, APRM power and downscale indicating lights, IRM/SRM power level, SRM period, and control rod position indication.
A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram can result in the design limits of the nuclear fuel being compromised. RPS is designed to automatically detect and generate a reactor scram signal when a Technical Specification Limiting Safety System Setting (LSSS) is reached or exceeded. If an LSSS is exceeded 7 without an autom.itic scram, consideration must be given to the possibility that a Technical (V Specification Safety Limit may have been exceeded.
EAL - 5.1.3 Rev.00 Page 2 of 3 l
HCGS EAL/RALTechnical Basis
~
t Emergency Operating Procedures (EOPs) establish Reactor Power > 4% coincident with a scram condition as the initiating condition for various actions in response to an ATWS. If the i Reactor is isolated (MSIVs closed), the heat generated is transferred to the Primary Containment, thus potentially threatening the integrity of Primary Containment . In an attempt to preclude this condition, EOP guidance includes restoration of the Main Condenser as a heat !
sink, provided there is no indication of gross fuel failure or a main steam line break. EOP guidance also includes methods of alternate reactivity control, including the use of Standby Liquid Control (SLC), alternate control rod insertion, and intentional lowering of Reactor Water 12 vel to control Reactor Power.
DEVIATION None REFERENCES 1 NUMARC NESP-0007, SS2 l NUMARC Questions and Answers, June 1993, " System Malfunctions Question #7" l HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV' Jontrol HC.OP-EO.ZZ-0207 (Q)-FC, Level / Power Control l O BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Definitions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation l
l O EAL - 5.1.3 Rev.00 Page 3 of 3
. HCGS EAL/RALTechnical Basis 5.0 Failure to Scram -
> tq 5.1 ATWS GENERAL EMERGENCY - 5.1.4 1C Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram (Automatic and Manual) and there is indication of an Extreme Challenge to the Ability to Cool the Core ,
EAL EITHER one of the following conditions:
- An Automatic Reactor Scram Condition exists M An Automatic Reactor Scram (RPS)
IS NOT successful
V e ALL Reactor Scram attempts from the Control Room (RPS gnd ARI) DID NOT REDUCE and MAINTAIN Reactor Power to 14%
AND EITHER one of the following:
- Reactor Water Level CANNOT BE MAINTAINED > -190"
- The combination of Suppression Pool Temperature and RPV Pressure CANNOT BE MAINTAINED below the HCTL Curve OPERATIONAL CONTITION - 1,2 BASIS l
Failure of the RPS to successfully complete a Reactor Scram (Automatic and Manual) represents a significant degradation in plant safety, as the primary reactivity control system has i p(f failed to perform its design function. In addition, failure of subsequent scram attempts (ARI)
EAL - 5.1.4 Rev.00 Page 1 of 4
HCGS EAL/RALTechnical Basis to reduce Reactor Power to less than 4%, resulting in an inability to MAINTAIN Reactor Water 12 vel above -190" or Suppression Pool Temperature and Reactor Pressure below the Heat Capacity Temperature Limit (HCTL), represents an imminent loss or potentialloss of all !
three fission product barriers. The inability to MAINTAIN Reactor Water Level above -190" was chosen based on the condition that core cooling is extremely challenged. This threshold )
corresponds directly to a decision step contained in EOP 207, Level / Power Control (Step LP- l 18), which requires a determination be made if Reactor Water Level can be MAINTAINED ,
above -190". For cases where Reactor Water Level CAN NOT BE M AINTAINER > -190",
a General Emergency declaration is warranted. The intent of this EAL is to classify those ATWS events that result in a challenge to the integrity of these barriers.
A Successful scram (RPS Automatic and Manual), as it relates to this EAL, results in a Control Rod configaration by which the Reactor will remain shutdown under all conditions without boron. The three criteria that satisfy this condition are :
- 1) All Control Rods are inserted to position 02 or beyond (Maximum Suberitical Banked Withdrawal Position)
- 2) All Control Rods but one being full inserted.
- 3) Reactor Engineering has determined that the Reactor will remain Shutdown under all conditions without Boron Barrier Analysis This event reaches the threshold for either a loss or potential loss of all three Fission Product Barriers.
ESCALATION CRITERIA N/A DISCUSSION l
The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is " fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical / pneumatic failure) or a failure of the Control Rod Drive system to permit the l control rods to insert (hydraulic failure). The Alternate Rod Insertion (ARI) function of the l Redundant Reactivity Control System (RRCS) provides an automatic backup function for an l electrical / pneumatic failure of the RPS.
Confirmation indications of an RPS failure to complete a su ressful scram include control room annunciators, control rod positions, APRM power ar ~ downscale indicating lights, j IRM/SRM power level, SRM period, and control rod position indication.
- O EAL - 5.1.4 Rev.00 Page 2 of 4
I ,
HCGS EAL/RALTechnical Basis A failure of the Reactor Protection System (RPS) to initiate and complete a reactor scram can result in the design limits of the nuclear fuel being compromised. RPS is designed to f
! automatically detect and generate a reactor scram signal when a Technical Specification l Limiting Safety System Setting (LSSS) is reached or exceeded. If an LSSS is exceeded i without an automatic scram, consideration must be given to the possibility that a Technical Specification Safety Limit may have been exceeded.
Emergency Operating Procedures (EOPs) establish Reactor Power > 4% coincident with a l scram condition as the initiating condition for various actions in response to an ATWS. If the l Reactor is isolated (MSIVs closed), the heat generated is transferred to the Primary -
l Containment, thus potentially threatening the integrity of Primary Containment. In an attempt to preclude this condition, EOP guidance includes restoration of the Main Condenser as a heat sink, provided there is no indication of gross fuel failure or a main steam line break. EOP l guidance also includes methods of alternate reactivity control, including the use of Standby Liquid Control (SLC), alternate control rod insertion, and intentional lowering of Reactor l Water I2 vel to cc uol Reactor Power.
During these actions, adequate core cooling is accomplished by maintaining Reactor Water I2 vel above -190". Although this is below the Top of Active Fuel (Loss of Core Submergence), maintaining Reactor Water I2 vel above -190" will ensure sufficient steam flow from the covered portion of the core to preclude Fuel Clad Temperatures in the ,
J uncovered portion of the core from exceeding 1500 Degrees F. This is refereed to as the l Minimum Steam Cooling RPV Water Level. Inability to maintain this level may result in damage to the fuel. 1 l
i The EOPs require the initiation of SLC before Suppression Pool Temperature reaches 110 )
Degrees F. This threshold is refereed to as the Boron Injection Initiation Temperature, and is 1 defined as the highest Suppression Pool Temperature at which initiation of boron injection will result in injection of the Hot Shutdown Boron Weight before Suppression Pool Temperature exceeds the Heat Capacity Temperature Limit (HCTL).
Actions required by the EOPs when Reactor Water Level can not be maintained above l -190" or the HCTL is exceeded include the initiation of Emergency Depressurization.
DEVIATION None l
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HCGS EAL/RALTechnical Basis 1
REFERENCES -
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NUMARC NESP-0007, SG2 l NUMARC Questions and Answers, June 1993, " System Malfunctions Question #7" l HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control l
HC.OP-EO.ZZ-0207 (Q)-FC, Level / Power Control l BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Definitions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation -
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. HCGS EAURALTechnical Basis l
lA 6.0 Radiological Releases / Occurrences 6.1 Gaseous Effluent Release l
UNUSUAL EVENT - 6.1.1.a i
- IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 2 I
! Times the Radiological Technical Specifications for 60 minutes or longer
~
l EAL l
Dose Assessment indicates EITHER one of the following ai the MEA or beyond as calculated on the SSCL: l
- TEDE 4-Day Dose of 12.0E-01 mrem !
e Thyroid-CDE Dose of A 6.8E-01 mrem I based on Plant Vent efnuent sample analysis and N_QI on a default Noble Gas to Iodine Ratio l I AND l l
Release is ongoing for 160 minutes OPERATIONAL CONDITION - All BASIS Dose Assessment at or beyond the MEA exceeding the EAL threshold, can result from a l 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 I 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 comolv 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 l magnitude. The use of dose assessment based EALs is therefore preferred over the use of Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default Meteorological data is used. As long as dose assessment is available, this EAL should be used in place of EAL 6.1.1.d.
EAL - 6.1.1.a Rev.00 Page 1 of 3
i
. HCGS EAL/RALTechnical Basis t .
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 decla ed as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer.
Barrier Analysis N/A ESCALATION CRITERIA l
l Emergency Classification will escalate to an Alert when the effluent release concentration ;
I increases to 200 times the Technical Specification limit.
DISCUSSION l 1
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 she), the associated site boundary dose rate would be 0.057 mrem /hr.
TEDE 4-Day MEA Dose Rate = (500 mrem /3r )(2)( 5)=0.057 mrem /hr I gyggg,,,7 1 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 l
= 0.2 mrem. I 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 I associated site boundary dose rate would be 0.17 s' Rem /hr. l l
Thyroid-CDEMEA Dose Rate = (1500 mrem /yr)(2)(.5)=0.17 mrem /hr 1 8766hrIpr The Thyroid-CDE Dose is cased 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 0.17 mrem /hr*4 hours = 0.68 mrem. l DEVIATION None l
EAL - 6.1.1.a Rev.00 Page 2 of 3
l
. HCGS EAURALTechnical Basis l
l REFERENCES' -
l NUMARC NESP-007, AUI.4 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents l NUMARC Draft White Paper, 7-25-94, 9-10-94.
l Technical Specification 3.11.2.1 l
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. MCGS EAURALTechnical Basis s 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 j EAL ,
1 l
Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS
.05 mrem /hr above normal background AND Release is ongoing for j>_60 minutes OPERATIONAL CONDITION - All BASIS I
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 l 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 N/A O
L.)
EAL - 6.1.1.b Rev. 00 Page 1 of 2
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, HCGS EAL/RALTechnical Basis )
l ESCALATION CRITERIA - 1 O Emergency Classification will escalate to an Alert when effluent release concentration increases to 200 times the Technical Specification limit.
DISCUSSION
)
Prorating the 500 mrem /yr criterion for: time (8766 hr/yr); the 2 x Tech. Spec. multiplier; and, ArtificialIsland's Allocation Factor of 0.5 (50% per site), the associated site boundary (MEA) dose rate would be 0.057 mrem /hr.
Protected Area Boundary Dose Rate = (500 mrem /yr)(2)(.5)=0.57 mrem /hr
. gy55g,y y, This is rounded to .05 mrem /hr DEVIATION None j REFERENCES NUMARC NESP-007, AU1.3 Off-Site Dose Calculation Manual, Section 2.0 - Geseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.
Technical Specification 3.11.2.1 1
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EAL - 6.1.1.b Rev.00 Page 2 or 2
. HCGS EAL/RALTechnical Basis l
6.0 Radiological Releases / Occurrences -
t b
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 10CFR20, Appendix B limits for 60 minutes or longer EAL Gaseous effluent release sample analysis for ANY one of the following indicates a concentration of:
- FRYS:
A 1.13L03 Ci/cc Total Noble Gas l A 2.71LO7 pCi/cc I-131
- NPV:
A_2.43604 Ci/cc Total Noble Gas 15.81LO8 Ci/cc I-131
(
- SPV: l A 2.27L05 Ci/cc Total Noble Gas l A_5.44LO9 pCi/cc I-131 l l
AND Dose Assessment results HQIavailable AND i . Release is ongoing for A 60 minutes l
l OPERATIONAL CONDITION - All !
l BASIS Gaseous effluent release sample analysis exceedin3 the EAL threshold for any of the plant !
, vents listed (FRVS, NPV, SPV), can result from a Gaseous Radiological Release in excess of l 2 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 - 6.1.1.c Rev.00 l Page 1 of a l
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i l ,
HCGS EAL/RALTechnicai Basis EAL is NOT based on a specine 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.
l The final integrated dose is very low and is not the primary concern. Classification is based on l an ongoing release that does not comoly with a license condition. The HTV is not included under this EAL since there are no provisions for collecting a HTV grab sample. Unplanned is denned as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit.
It is not intended that the release be averaged over 60 minutes, but exceed 2 times the l 10CFR20, Appendix B limit for 60 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer.
Barrier Analysis l N/A ESCALATION CRITERIA Emergency Classification will escalate to an Alert when the effluent release concentration increases to 200 times the 10CFR20, Appendix B limit.
DISCUSSION Refer to Basis Section for EAL 6.1.1.d for the 10CFR20, Appendix B Noble Gas and Thyroid
- committed Dose release rate calculations.
Calculation of the threshold sample concentrations are as follows:
FRVS Noble Gas Sample Concentration = 4.80E+ 03 Ci/ sec =1.13E-03 Cilcc 472x9000cfn I
1.15E + 00 Ci / sec FRVS 1-131 Sample Concentration = =2.71E-07pCilcc 472r9000cfn NPVNoble Gas 4.80E + 03 Ci / sec Sample Concentration = =2.43E-04pCi/cc 472r4.19E + 04cfn 1.1SE + 00/ Ci / see NPVl-131 Sample Concentration = = 5.8l E-08pCilcc 472x4.19E + 04cfn l
O EAL - 6.1.1.c Rev.00 Page 2 of 3
HCGS EAL/RALTechnical Basis SPV Noble Gas Sample Concentration = 4.80E + 03pci/ see =2.27E-05pCilcc 472x4.48E + 05cfin 1.15E + 00 Ci/ sec SPVl-131 Sample Concentration = =5.44E-09pCi/cc 472x4.48E + 05cfin Where: 472 = conversion factor (28,317 cc/ft3 x 1 min./60 sec.)
9000 cfm = FRVS Vent Flow (maximum) 4.19E+04 cfm = NPV Vent Flow (maximum) -.-
l 4.48E+05 cfm = SPV Vent Flow (maximum)
The noble gas release rate of 4.80E+03 Ci/sec is obtained by multiplying the 10CFR20, Appendix B limit release rate of 2.40E+03 Ci/see times 2.
i The iodine release rate of 1.15E+00 Ci/sec is obtained by multip'ying the 10CFR20, Appendix B limit release rate of 5.75E-01 pCi/see times 2. l t
l l DEVIATION The value for EAL 6.1.1.c is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate coula 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 i
6.1.1.c would not be used unless EAL 6.1.1.a / Dose Assessment) can not be used to I (v determine the classification, if any, due to the potential uncertainty of this " default" EAL.
l 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, AU1.2, AUI.1, AUI.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
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HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences
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l l 6.1 Gaseous Effluent Release UNUSUAL EVENT - 6.1.1.d l
l IC Any Unplanned Release of Gaseous Radioactivity to the Environment that E ceeds 2
- times the 10CFR20, Appendix B limits for 60 minutes or longer
~
EAL j Valid High Alarm received from ANY one of the following Plant Effluent RMS l Channels:
l . FRVS Noble Gas (Grid 1/3; 9RX680) l
. NPV Noble Gas (Grid 1/3; 9RX590) l . SPV Noble Gas (Grid 1/3; 9RX580)
. IITV Noble Gas (Grid 1/3; 9RX518)
AND l Total Plant Vent release rate EXCEEDS EITHER one of the following limits:
. 4.80E+ 03 Ci/sec Total Noble Gas l
. 1.15E+ 00 Ci/sec I-131 (NPV & SPV ONLY) i AND Dose Assessment results N_QT available AND Release is ongoing for .;>_60' minutes OPERATIONAL CONDITION - Ali BASIS Valid High alarm and effluent release rate values exceeding the EAL threshold, can result from a Gaseous Radiological Release in excess of 2 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 m 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
, EAL - 6.1.1.d l Rev. 00 Page 1 of 4 l
-.- - -. =- - .- - - -.- - --- . - . -. - - - . - .... - - - - . . . . .
l . HCGS EAL/RALTechnical Basis l
l primary concern. Valid is defined.as the High alarm actuating Soecifically due to a Gaseous I Release Ning 10 CFR 20, Appendix B limits, thus precluding unwarranted event i declaration as the result of spurious actuation. Classification is based on an ongoing release that does not comoly 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 j applicable permit The EAL value for Total Plant Vent release rate was determined using default X/Q values from the ODCM which provides a less accurate method of evaluation release magnitude then using dose assessment with real time meteorological data. For that reason, this EAL should not
- - be utilized if Dose Assessment is available. Dose Assessment will take in account actual l meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available than refer to EAL 6.1.1.a for classification.
It is not intended that the release be averaged over 60 minutes, but exceed 2 times 10 CFR20, ,
Appendix Blimits 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.
l Barrier Analysis N/A ESCALATION CRITERIA t Emergen'.:y Classification will be escalate to an Alert when the effluent release concentration l increases to 200 times the 10CFR20, Appendix B limits.
DISCUSSION-l The release rate thresholds for this EAL are obtained by multiplying the Technical Specification release rates of 2.4E+03 Ci/sec and 5.78E-01 Ci/sec, for Noble Gases and Iodine-131 respectively, times 2. Total Noble Gas release rate is the summation of all plant vent release rates.
This EAL includes Iodine Release Rates for the NPV and SPV, since these vents have an Iodine monitor. Determination of the Iodine Release Rate from the Iodine monitor is accomplished by multiplying the Iodine reading (in uCi/cc) by the applicable vent flow rate, and 472 (Conversation factor). Iodine Release rates for FRVS and the HTV are excluded t since these vents do not include an Iodine detector. The SPDS Total Iodine Offsite Release
- Rate does not provide useful information because this is based on a default value of 1000 times less than the Total Noble Gas Offsite Release Rate, which could be grossly inaccurate.
EAL - 6.1.1.d Rev.00 Page 2 of 4 l
1
. HCGS EAL/RALTechnical Basis Release rates for FRVS and the HTV are not included since these vents do not have an Iodine detector. A gaseous effluent sample is needed to accurately quantify the Iodine Release rate.
O/ The SPDS Total Iodine Offsite Release Rate should not be used, as this is based on a default value of 1000 times less than the Total Noble Gas Offsite Release Rate. The 10CFR20, Appendix B limits are based on ODCM calculations.
10CFR20, Appendix B Calculation for Noble Gas uCi/Second = (l00 mRemIyear)*(Al/ccation Factor)
~
WHERE: uCl/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, ODCM X/Q = 2.67E-06sec/m' ODCM DRCF = 7.80E+03 mrem /yr/uCi/m 3 Allocation Factor = 5.00E-01 2.40E+03uCi/Second =
(2.67E - 06 sec/ m') * (7.80E + 03m Re m / yr / pCi / m')
2.40E+03uCi/Second *2 = EAL value.
4.80E+03pCi/sec is the EAL value.
10CFR20, Appendix B Calculation for Thyroid Committed Dose uCi/Second = 50 mrem / year * (Allocation Factor)
WHERE: uCi/Second = Total Iodine 131 release rate from Salem (Unit 1 or 2) or Hope Creek (all Vents; NPV, SPV, FRVS and HTV).
ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3, EAL - 6.1.1.d Rev. 00 Page 3 of 4
1
. HCGS EAURALTechnical Basis ODCM DRCF = is the most limiting potential pathway (inhalation, child, thyroid I-131) dose rate conversion factor in mrem / year /uCi/m3 ,
ODCM X/Q = 2.67E-06 ODCM DRCF THY = 1.62E+07 mrem /yr/uCi/m3 Allocation Factor = 5.00E-01 ,
1 5.78E-01uCi/Second = (50 mrem /vear) * (5.00E-01) l (2.67E-06 sec/m 3 )* (1.62E+07 mrem /yr/uci/m3 )
5.78E-01uCi/Second
- 2 = EAL value.
1.15E+'.)0pCi/sec is the EAL value. !
1 DEVIATION l The value for EAL 6.1.1.d is based on one meteorological case and one isotopic mixture found l in the ODCM. A radiological release based on this specific release rate could produce a i TEDS Dose which would require an Alert classification or not meet the Unusual Event !
classification, depending on the meteorological conditions and the isotopic mixture. EAL ,
6.1.1.d would not be used unless EAL 6.1.1.a (Dose Assessment) can not be used to l determine the classification, if any, due to the potential uncertainty of this " default" EAL ;
Two times the 10CFR20, Appendix B limits for noble gas and Iodine 131 are being used for pd this EAL, due to concerns that the State of New Jersey have 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 l 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. )
i REFERENCES i
- NUMARC NESP-007, AUI.1, AUI.4 i HC.OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity I '
HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.
Technical Specification 3.11.2.1 o
- O EAL - 6.1.1.d Rev.00 Page 4 of 4 r., ,
._______.----_____i+ __, .m i is e--. .. m -,-g - n ,ei.,r. - ,+ in , 7 g + _----,m.+.g-m----
__ %+.,9_
i l
HCGS EAL/RALTechnical Basis l 6.0 Radiological Releases / Occurrences -
! C')
\/ 6.1 Gaseous Effluent Release 1
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 -
i i
Dose Assessment indicates EITHER of the following at the MEA or beyond as l calculated on the SSCL:
- TEDE 4-Day Dose of A 2.0E+01 mrem; i e Thyroid-CDE Dose of A 6.8E+01 mrem based on Plant Vent effluent sample analysis and NOT on a default Noble Gas to i Iodine Ratio i AND Release is ongoing for A 15 minutes l
OPERATIONAL CONDITION - All BASIS Dose Assessment at or beyond the MEA exceeding the EAL threshold , can result from a Gaseous Radiological Release in excess of 200 times Technical Specifications. This condition l 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. 'Onplanned is defined as
! any release for which a radioactive discharge permit was not prepared, or a release that l exceeds the conditions on the applicable permit.
i l 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 l
Page 1 of .1 l
, HCGS EAURALTechnical Basis O It is not .atended 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 N/A ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when the effluent release concentration increases to a level that would cause a 100 mrem dose at the Protecid 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.
TEDE 4-Day MEA Dose Rate = ( )(200)(05)=5.7 mrem /hr This is rounded to 5.0 mrem /hr. I 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*4 hours = 20 mrem.
~
Prorating the 1500 mrem /yr criterion for the Thyroid-CDE Dose: time (8766 hr/yr); the 200 j x Tech. Spec. multiplier; and, Artifi.ial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 17 mrem /hr.
lhyroid-CDEMEA Dose Rate = (1500 mrem /yr)(200)(.5)=0.17 mrem /hr gyggy,jy, 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 duration. Therefore 17 mrem /hr*4 hours = 68 mrem.
DEVIATION None EAL - 6.1.2.a l Rev.00 l
Page 2 of 3 l
.____ -. _ . . _ _ . . . - . . _ . . . _ _ - . _ _ _ = . . _ _ . _ . . _ . _ _ . _ . _ _ _ . _ _ _ _ _ _ . . _ . _ , _ . _ . _
i j . HCGS EAURALTechnical Basis i REFERENCES -
O NUMARC NESP-007, AA1.4 j Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents
]' NUMARC Draft White Paper, 7-25-95, 9-10-94.
j Technical Specification 3.11.2.1 i
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O EAL - 6.1.2.a Rev. 00 Page 3 of 3 )
. HCGS EAURALTechnical Basis 6.0 Radiological Releases / Occurrences -
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6.1 Gaseous Effluent Release ALERT - 6.1.2.b l
l IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 Times Radiological Technical Specifications for 15 minutes or longer EAL l Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 5 mrem /hr AND Release is ongoing for A 15 minutes i
l l
OPERATIONAL CONDITION - All l
, 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 1 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 i i on the applicable permit.
It is not intended that the release be averaged over 15 minutes, but exceed 200 times the I Technical Specification limit for 15 minutes or longer. In addition, it is intended that the j event be declared as soon as it is determined that the release will exceed 200 times the limit for l 15 minutes or longer.
Barrier Analysis N/A EAL - 6.1.2.b Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis ESCALATION CRITERIA -
O b' 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.
l l DISCUSSION l
Prorating the 500 mrem /yr criterion for: time (8766 hr/yr); the 200 x Tech. Spec.
multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 5.7 mrem /hr.
Protected Area Boundary Dose Rate = ( 500 mrem /yr )(200)(.5)=5.7 mrem /t r '
i gyggy,,y, This is rounded to 5.0 mrem /hr DEVIATION !
l None REFERENCES l NUMARC NESP-007, AA1.3 l Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents i NUMARC Draft White Paper, 7-25-94, 9-10-94.
l Technical Specification 3.11.2.1 )
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!O EAL - 6.1.2.b Rev. 00 Page 2 of 2
. HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences i
6.1 Gaseous Effluent Release ALERT - 6.1.2.c i
IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 l l
Times the 10CFR20, Appendix B limits for 30 minutes or longer
~
EAL Total gaseous effluent release sample analysis for ANY one of the following indicates a concentration of:
- FRVS:
11.13L01 Ci/cc Total Noble Gas 12.71605 pCi/cc I-131
- NPV:
12.43L02 Ci/cc Total Noble Gas
_ 15.81606 Ci/cc I-131 1
, U . SPV:
1 2.27 & O3 Ci/cc Total Noble Gas ,
1 5.44607 Ci/cc I-131 l l
i AND 1 1
Dose Assessment results HQI available AST l
Release is ongoing for 130 minutes l
l OPERATIONAL CONDITION - All BASIS Total gaseous effluent release sample analysis exceeding the EAL threshold for any of the plant vents listed (FRVS, NPV, SPV), can result from a Gaseous Radiological Release in excess of 200 times 10CFR20, Appendix B limitsTechnical Specifications. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated EAL - 6.1.2.c Rev. 00 Page 1 of 3
- - - ~ - - _ - _ - _ . . . - . - - _ __ - - - - - . . . _ - - . . - . . - .
. HCGS EAL/RALTechnical Basis offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose rate, but O rather on the loss of plant control implied by a radiological release of this magnitude that was not isolated within 30 minutes. The final integrated dose is very low and is not the primary concern. Classification is based on an ongoing release that does not comply with a license condition. The HTV is not included under this EAL since there are no provisions for collecting a HTV grab sample. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit.
l It is not intended that the release be averaged over 30 minutes, but exceed 200 times the i 10CFR20, Appendix B limit for 30 minutes or longer. In addition, it is intended that the event l be declared as soon as it is determined that the release will exceed 200 times the limit for 30
! minutes or longer. i
! 1 Barrier Analysis N/A-l ESCALATION CRITERIA +
l l
Emergency Classification will escalate to a Site Area Emergency whea effluent release ;
concentration increases to a level that would cause a 100 mrem dose or Thyroid-CDE of ;
500 mrem for I-131 at the Protected Area Boundary.
' DISCUSSION ,
Refer to Basis Section for EAL 6.1.2.d for the 10CFR20, Appendix B Noble Gas and Thyroid :
Committed Dose release Rate Calculations. ;
Calculation of the threshold sample concentrations are as follows: )
l FRVS Noble Gas Sample Concentration = 4.80E + 05 Ci/ sec = 1.13E-01pCilcc 472x9000cfm FRVS 1-131 Sample Concentration = 1.15E + 02 Ci /=2.71E-05 see Cilcc 472x9000cfm NPVNoble Gas Sample Concentration = 4.80E + 05 Ci / see =2.43E-02pCilcc 472x4.19E + 4cfm NPVl-131 Sample Concentration = 1.15E + 02 Ci / sec =5.81E-06 Cilcc 472x4.19E + 04cfm O EAL - 6.1.2.c Rev.00 l
Page 2 of 3
. HCGS EAL/RALTechnical Basis p SPVNoble Gas Sample Concentration = 4.80E + 05 Ci / sec- =2.27E-03 Cilcc g 472x4.48E + 5cfm SPVl-131 Sample Concentration = 1.15E + 02 Ci/ sec =5.44E-07 Ci/cc 472x4.48E + 05cfm Where: 472 = conversion factor (28,317 cc/ft3 x 1 min./60 sec.)
9000 cfm = FRV3 Vent Flow (maximum) 4.19E+04 cfm = NPV Vent Flow (maximum) 4.48E+05 cfm = SPV Vent Flow (maximum)
The noble gas release rate of 4.80E+05 Ci/sec is obtained by multiplying the 10CFR20, Appendix B limit release rate of 2.40E+03 Ci/see times 200.
The iodine release rate of 1.15E+02 Ci/sec is obtained by multiplying the 10CFR20, Appendix B limit release rate of 5.75E-01 Ci/sec times 200.
DEVIATION The value for EAL 6.1.2.c is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate could produce a TEDE Dose which would require a General Emergency classification or not meet the Alert classification, depending on the meteorological conditions and isotopic mixture. EAL 6.1.2.c c' would not be used unless EAL 6.1.2.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this " default" EAL.
Two hundred times the 10CFR20, Appendix B limit noble gas and Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey had pertaining to this EAL and based on the above mentioned uncertainties.
The time limit has been increased from 15 minutes to 30 minutes, to allow additional time to perform dose assessment, since the threshold fore this EAL is only 20% of the value allowed per NESP-007 and we do not wish to use this " default" EAL, unless absolutely necessary.
REFERENCES NUMARC NESP-007, AA1.2, AAl.1, 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 l
O V
EAL - 6.1.2.c Rev.00 l Page 3 of 3 l
HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences O 6.1 Gaseous Effluent Release ALERT - 6.1.2.d IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 Times 10 CFR20, Appendix B Limits for 30 minutes or longer
~
EAL Valid High Alarm received from ANY one of the following Effluent RMS Channels:
. FRVS Noble Gas (Grid 1/3; 9RX680)
. NPV Noble Gas (Grid 1/3; 9RX590)
. SPV Noble Gas (Grid 1/3; 9RX580)
. IITV Noble Gas (Grid 1/3; 9RX518)
AND Total Plant Vent release rate EXCEEDS EITHER one of the following limits:
g . 4.80E+05 C1/sec Total Noble Gas :
1.15E+ 02 Ci/sec I-131 (NPV & SPV QNLY)
AND Dose Assessment results NQT available l AND Release is ongoing for .>_30 minutes OPERATIONAL CONDITION - All BASIS l Valid High alarm and effluent release rate values exceeding the EAL threshold, 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, l
resulting in elevated offsite dose rates. The threshold for this EAL is NOT based on a specific n 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
()
EAL - 6.1.2.d Rev.00 Page 1 of 4 i
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i
. HCGS EAL/RALTechnical Basis
! 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 l
f]
C 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 whici.
l a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the l applicable permit..
l The EAL value for Total Plant Vent release rate was determined using default X/Q values l from the ODCM which provides a less accurate method of evaluation release magnitude then i using dose assessment with real time meteorological data. For that reason, this EAL shotrid not be utilized if Dose Assessment is available. Dose Assessment will take in account actual
! meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available than refer to EAL 6.1.2.a for classification.
The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPV, FRVS and HTV noble gas readings.
It is not intended that the 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.
/ \
'U Barrier Analysis N/A ESCALATION CRITERIA l
l 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 j Boundary l DISCUSSION The release rate thresholds for this EAL are obtained by multiplying the 10CFR20, Appendix B Limit release rates of 2.4E+03 pCi/sec and 5.78E-01 pCi/sec for Noble Gases and I-131 respectively, times 200. Total Noble Gas release rate is the summation of all plant vent release rates.
l This EAL includes an Iodine Release rate for NPV & SPV since these vents have Iodine monitors. Determination of the Iodine Release Rate from the Iodine monitor is accomplished by multiplying the Iodine reading (in pCi/cc) by the applicable vent flow rate and 472 p)
'v (conversion factor). Iodine Release Rates for FRVS and HTV are excluded since these vents EAL - 6.1.2.d Rev.00 Page 2 of 4
I I I
l .
HCGS EAL/RALTechnical Basis do not include an lodine detector. The SPDS Total Iodine offsite Release Rate does not
,(3 provide useful information, since this based on a default valve of 1000 times less than the
!V Total Noble Gas offsite Release Rate , which could be grossly inaccurate. A gaseous effluent sample is needed to accurately quantify the lodine Release Rate.10CFR20, Appendix B Limit 1 Calculation for Noble Gas l uCi/Second = 100 mrem / year * (Allocation Factor) l (ODCM X/Q) * (ODCM DRCF) i l WHERE: uCi/Second = Total Noble Gas Release Rate from Salem (Unit 1 & Urrit 2) ;
I or Hope Creek (all Vents; NPV, SPV, FRVS, and HTV) l l
which would result in a TEDE Dose Rate of 50 mrem / year.
1 ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3, l
, ODCM DRCF = Site Specific (Salem or Hope Creek) dose rate l conversion factor in mrem / year /uCi/m3 ,
ODCM X/Q = 2.67E-06 ODCM DRCF = 7.80E+03 mrem /yr/uci/m3 Allocation Factor = 5.00E-01 2.40E+03 uCi/Second = (100 mrem /vear) * (5.00E-01) l (2.67E-06 sec/m 3
) * (7.80E+03 mrem /yr/uCi/m3 )
2.40E+03 uCi/Second
- 200 = EAL value.
4.80E+05pCi/sec = EAL value l
10CFR20, Appendix B Limit Calculation for Thyroid Committed Dose pCi/Second = 50 mrem / Year * (Allocation Factor)
Where: Ci/Second = Total Iodine 131 release rate from Salem (Unit 1 or 2) or Hope Creek (all Vents; NPV, SPV, FRVS,and HTV).
ODCM x/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in Sec/m'.
ODCM DRCF = is the most limiting potential Pathway (inhalation, child thyroid 1-131) dose rate conversion factor in mrem / year /pCi/m'.
EAL - 6.1.2.d Rev.00 Page 3 of 4 l
l
, HCGS EAURALTechnical Basis ODCM X/ Q = 2.67E-06 ODCM DRCF THY = 1.62E+07 mrem /yr/pCi/m' Allocation Factor = 5.00E-01 5.78E-01 (pCi/Second = (50 mrem / year) * (5.00E-01) 3 (2.67E-06Sec/m ) * (1.62E+07 mrem /vr/uCi/mh 5.78E-Oluci/sec
- 200 = EAL value.
1.15E+02uCi/sec = EhL 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 Dose which would require a General Emergency classification or not meet the Alert classification, depending on the meteorological conditions and the isotopic mixture. EAL 6.1.2.d would not be used unless EAL 6.1.2.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this " default" EAL.
Two hundred times the 10CFR20, Appendix B limits of 100 mrem / year noble gas and 50 Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey had pertaining to this EAL and based on the above mentioned uncertainties.
The time limit has been increased from 15 minutes to 30 minutes, to allow additional time to l 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 l NUMARC NESP-007, AAl.1, AA1.4 l . OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.
Technical Specification 3.11.2.1 l
lO EAL - 6.1.2.d Rev.00 i
Page 4 of f
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- = _ - . . _
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. HCGS EAL/RALTechnical Basis
- 6.0 Radiological Releases / Occurrences -
l0
'V 6.1 Gaseous Efiluent Release SITE AREA EMERGENCY - 6.1.3.a l
l IC Boundary Dose Resulting from an Actual or Inmiinent Release of Gaseous l 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 i
l Dose Assessment indicates ELTHER one of the following at the MEA or beyond as l calculated on the SSCL:
l TEDE 4-Day Dose of A 1.0E+02 mrem Thyroid-CDE Dose of 15.0E+02 mrem based on Plant Vent effluent sample analysis and M.
Q I on a default Noble Gas to Iodine Ratio i
OPERATIONAL CONDITION - All BASIS l
l The TEDE 4-Day Dose of 100 mrem corresponds directly to the NUMARC dose of 100
( ' mrem.
- The Thyroid-CDE Dose of 500 mrem corresponds directly to the NUMARC dose of 500 .
- mrem. !
Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default Meteorological data is used. Imminent is defined as expected to occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Barrier Analysis N/A i
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. HCGS EAL/RALTechnical Basis i i
ESCALATION CRITERIA -
l Emergency Classification escalates to a' General Emergency when actual or projected doses exceed EPA Protective Action Guidelines. I DISCUSSION I ,
l This value provides a desirable gradient (one order of magnitude) between the Site Area j Emergency and General Emergency classifications. No site a' location factor (.5) is used in this calculation due to the assumption that releases of this magnitude will be from one sitt.
! i 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. 4
)
DEVIATION '
NONE !
REFERENCES
{
! NUMARC NESP-007, AS1.3 .
p EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear l
d Incidents NUMARC Draft White Paper, 7-25-94, 9-10-94 l
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HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
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SITE AREA EMERGENCY - 6.1.3.b i IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem Total Effective Dose Equivalent (TEDE) or 500 l mrem Thyroid CDE Dose for the actual or projected duration of the release
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EAL Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 100 mr/hr -
AND Re. lease is expected to continue for A 15 minutes
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OPERATIONAL CONDITION - All BASIS An actual dose rate of 100 mrem /hr which is expected to continue for 115 minutes indicates a substantial radiological release which could exceed the 10CFR20 annual average population exposure limit of 100 mrem TEDE, using the assumption of a one hour release duration.
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 N/A i ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when actual or projected doses 4 exceed EPA Protective Action Guidelines.
DISCUSSION An actual dose of 100 mrem Total Effective Dose Equivalent (TEDE) is based on the 10CFR20 annual average population exposure limit. Unless otherwise indicated, the conversion from whole body dose to TEDE is 1:1. Measured dose rates will be taken at the (O
EAL - 6.1.3.b Rev.00 Page 1 of 2
HCGS EAL/RALTechnical Basis Protected Area Boundary, and a .2.15 minute release duration threshold will be applied to be consen ative.
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 O
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. HCGS EAIJRALTechnical Basis 6.0 Radiological Releases / Occurrences -
O 6.1 Gaseous Effluent Release SITE AREA EMERGENCY - 6.I.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 CDE Dose for the actual or projected duration of the release --
EAL Analysis of field survey samples at the Protected Area Boundary indicates EITHER one of the following:
e > 4.36E+02 CCPM e > 3.85E-07 Ci/cc I-I3I I
l OPERATIONAL CONDITION - All O
V BASIS l 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-140N with an HP260 probe attached. The Iodine-131 field survey sample concentration threshold is based on I-131 dose conversion factors from EPA-400. The thresholds are based on a Thyroid-CDE dose rate of 500 l mrem /hr for I-131. 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 />. l Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines.
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, HCGS EA1)RALTechnical Basis 6.0 Radiological Releases / Occurrences -
6.1 Gaseous 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 CDE Dose for the actual or projected duration of the release EAL Analysis of field survey samples at the Protected Area Boundary indicates EITHER one of the following:
e >4.36E+02 CCPM e > 3.85E-07 Ci/cc I-131 OPERATIONAL CONDITION - All BASIS The Corrected Counts per Minute (CCPM) value is based on reading (s) obtained using a radiation count rate meter such as a RM-14 or E-140N with an HP260 probe attached. The Iodine-131 field survey sample concentration threshold is based on I-131 dose conversion factors from EPA-400. The thresholds are based on a Thyroid-CDE dose rate of 500 mrem /hr thyroid for I-131. 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 N/A ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines.
O EAL - 6.1.3.c Rev.00 Page i of 3 i
. HCGS EAURALTechnical Basis DISCUSSION -
/^~N O The release sample concentration calculations are as follows.
The sample concentration is calculated using the I-131 Dose Conversion Factor from EPA-400:
Solving the following equation for pCi/cc:
mrem /hr = ( Ci/cc)(Dose Conversion Factor)
Then; 500 mrem /hr 1-131 Sample Concentration = (1.30E+09 mrem / Ci/cc/hr)=3.85E-07pCi/cc Where 1.30E+09 mrem /pCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4 and includes the EPA-400 breathing rate .
f The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, l and factors for using an RM-14 or E-140N with an HP260 probe.
Solving the following equation for CCPM:
Ci/cc = ceru (Detector Emciency)(Collection Emciency)(Conversion Factor DPM to pCiXVolume . 3h XConversion Factor ec to A3 )
Then; CCPM =(3.85E-07pCi/cc)(0.9)(2.22E+06DPM/pCi)(2.00E-03CCPM/DPM)
(10ft')(2.832E+04cc/ft') = 4.36E+02 CCPM i Where:
2.00E-03 = DetectorEpciency- CCPM/DPM 0.9 (or 90%) = Collection Efciency 2.22E+06 = Conversionfactor - DPM/pCi 10ft3= Volume 2.832E+04 = 3 Conversionfactor- cc toft i CCPM = Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe.
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, . HCGS EAL/RALTechnical Basis l
l DEVIATION -
l l l None i REFERENCES
- NUMARC NESP-007, ASI.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear j Incidents FEMA REP-2, Rev.1, 7/87, Guidance on Offsite Emergency Radiation Measurement- i Systems, Phase-1 Airborne Release .
SORC Summary 07/10/39 RPCS Thyroid Dose Commitment Factor Paper (NRP-94-0557),11/22/94 l
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HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
6.1 Gaseous Effluent Release SITE AREA EMERGENCY - 6.1.3.d l
IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 100 l
mrem TEDE 4-Day Dose for 30 minutes or longer
! EAL Valid High Alarm received from M of the following Effluent RMS Channels:
- FRYS Noble Gas (Grid 1/3; 9RX680)
- NPV Noble Gas (Grid 1/3; 9RX590)
. SPV Noble Gas (Grid 1/3; 9RX580)
- IITV Noble Gas (Grid 1/3; 9RX518)
M Total Plant Vent release rate EXCEEDS 4.21E+07 Ci/sec Total Noble Gas i
M Dose Assessment results HQT available M
Release is ongoing for .>_30 minutes OPERATIONAL CONTITION - 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.
l l 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 EAL - 6.1.3.d Rev.00 Page 1 of 3
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. HCGS EAL/RALTechnical Basis
! be utilized if Dose Assessment is available. Dose Assessment will take in account actual meteorological conditions, plant vent Dows and plant vent effluent concentrations to provide a A more accurate assessment of a radiological release. If Dose Assessment is available than refer to EAL 6.1.3.a for classification.
The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPV, FRVS and HTV noble gas readings.
It is not intended that the release be averaged over 30 minutes but that the Release Rate exceed the EAL value for >30 minutes. -
Barrier Analysis N/A ESCALATION CRITERIA Emergency 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 performanc~e 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 perfo:med within 30 minutes.
The ODCM methodology calculates yearly values. To be consistent with the ODCM methodology the SAE classification trigger point of 100 mrem / hour needs to be converted to a yearly dose. This is done in the following manner; 365 days / year
- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day = 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year.
100 mrem / hour
- 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year = 8.76E+05 mrem / year.
ODCM Dose Rate Conversion Factor = 7.80E+03 mrem / year /pCi/m' ODCM X/Q = 2.67E-06 Sec/m'
- No allocation factor is used for SAE.
l l 8.76E+05 mrem /vear = 4.21E+07 pCi/Sec 3
2.67E-06 Sec/m
- 7.80E+03 mrem / year /pCi/m' EAL - 6.1.3.d Rev.00 Page 2 of 3 i
HCGS EAL/RALTechnical Basis ,
i 4 .21E+07 pCi/Sec is the SAE Total Noble Gas Release Rate. J O DEVIATION This EAL is based on default meteorological and isotopic mixture assumption as found in the ODCM. Depending on actual meteorological conditions and isotopic mixture, the Release Rate used as the threshold value in this EAL could produce TEDE value which could be within l Alert ranges or as high as the General Emergency threshold. This potential to overclassify or underclassify this event is not desirable. To preclude / limit this possibility, PSE&G has used 30 minutes instead of 15 as in ASI.1. This extra 15 minutes would allow personnel to obtain l dose assessment projections from a second onsite computer should the primary location )
computer fail. In Addition events that result in a release of this magnitude would required l degradation of multiple fission product barriers and should be classified per Section 3, Fission l Product Barriers.
l REFERENCES 1
NUMARC NESP-007, ASI.1, ASI.4 OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.
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. HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
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'd 6.1 Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.a IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mrem Total Effective Dose Equivalent (TEDE) or 5000 mrem Thyroid CDE Dose for the actual or projected duration of the release _
EAL Dose Assessment indicates EITHER one of the following at the MEA or beyond as calculated on the SSCL:
- TEDE 4-Day Dose of 11.0E+03 mrem
. Thyroid-CDE Dose of 15.0E+03 mrem based on Plant Vent effluent sample analysis and NQI on a default Noble Gas to Iodine Ratio OPERATIONAL CONDITION - 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. ;
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The Thyroid-CDE Dose or 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 N/A ESCALATION CRITERIA N/A i
O EAL - 6.1.4.a Rev.00 Page 1 of 2
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. HCGS EAURALTechnical Basis l
l- DISCUSSION -
No site allocation factor (.5) is used in this calculation due to the assumption that releases of j this magnitude will be from one site.
l DEVIATION NONE' ,
! REFERENCES -
i NUMARC NESP-007, AGl.3 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear !
Incidents l NUMARC Draft White Paper 7-25-94,9-10-94 l
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. HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
6.1 Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.b IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mrem Total Effective Dose Equivalent (TEDE) or 5000 mrem Thyroid CDE Dose for the actual or projected duration of the release _
EAL Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 1000 mrem /hr AN_P_
Release is expected to continue for j>__15 minutes p OPERATIONAL CONDITION - All Y BASIS An actual dose rate of 1000 mrem /hr indicates the EPA Protective Action Guide may be exceeded for the general public. 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 N/A ESCALATION CRITERIA N/A 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 l
. . _ _ _ _ . _ . . . _ _ . _ . . _ . _ . . _ . . _ _. .__._.m_
. HCGS EAURALTechnical Basis DEVIATION - .
i None I REFERENCES NUMARC NESP-007, AGl.4 .
EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear !
Incidents R l.
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' 6.0 Radiological Releases / Occurrences -
(G 6.1 Gaseous Efiluent Release l GENERAL EMERGENCY - 6.1.4.c 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 Analysis of field survey samples at the Protected Area Boundary indicates EITHER one of the following:
> 4.36E+ 03 CCPM
> 3.85E-06 Ci/cc I-131 OPERATIONAL CONDITION - All BASIS The Corrected Counts per Minute (CCPM) value is based on reading (s) obtained using a radiation count rate meter such as a RM-14 or E-140N with an HP260 probe attached. The Iodine-131 field survey sample concentration threshold is based on I-131 dose factors from EPA-400. The thresholds are based on a dose rate of 5000 mrem /hr Thyroid-CDE for I-131.
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 N/A ESCALATION CRITERIA N/A DISCUSSION i The release sample concentration calculations are as follows.
l n The sample concentration is calculated using the I-131 Dose Factor from EPA-400:
b EAL - 6.1.4.c Rev.00 Page 1 of 3
i 1
, HCGS EAURALTechnical Basis l Solving the following equation for pCi/cc: -
mrem /hr = (pCi/cc)(Dose Conversion Factor)
Then; 5000 mrem /hr I-131 Sample Concentration = (1.30E+09 mrem /pCi/cc/hr)=3.85E-06 Ci/cc l Where 1.30E+09 mrem / Ci/cc/hr is the Dose conversion factor from EPA-400, Table 5-4 and includes the EPA-400 breathing factor.
The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, and factors for using an RM-14 or E-140N with an HP260 probe.
l Solving the following equation for CCPM:
Ci/cc = ceru (Detector Emciency)(Collection Emciency)(Conversion Factor . DPM to pCiHVolume3. fi )(Converso Factor. cc to ft3)
Then; CCPM = (3.85E-%pCi/cc)(0.9,)(2.22E+06DPM/ Ci)(2.00E-03CCPM/DPM)(10ft')
- v/ (2.832E+04cc/ft ) = 4.36E+03 CCPM
! Where:
2.00E-03 = Detectc 'Wiciency- CCPMIDPM 0.9 (or 90%) = Coller .n Epiciency
! 2.22E+06 = Conversionfactor- DPM/pCi 10ft3= Volume i 2.832E+04 = 3 Conversionfactor- cc toft CCPM = Corrected Counts per Minute using an RM-14 or E-140N l with an HP260 probe.
- DEVIATION None ,
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, HCGS EAURALTechnical Basis REFERENCES -
NUMARCNESP-007, AGl.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actionsfor Nuclear incidents FEMA REP-2, Rev.1/ July 1987, Guidance on Ofsite Emergency Radiation Measurement Systems, Phase-1 Airborne Release SORC Summary 07/10/89 j RPCS Thyroid Dose Commitment Factorpaper NRP-94-OSS7,11-22-94 l
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. HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
O 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 High Alarm received from ANY one of the following Pint Effluent RMS Channels:
- FRVS Noble Gas (Grid 1/3; 9RX680)
. NPV Noble Gas (Grid 1/3; 9RX590)
. SPV Noble Gas (Grid 1/3; 9RX580)
. IITV Noble Gas (Grid 1/3; 9RX518)
AND U Total Plant Vent release rate EXCEEDS 4.21E+08 Ci/sce Total Noble Gas AND j Dose Assessment results N_QT available AND Release is ongoing for 2. 30 minutes I l
OPERATIONAL CONDITION - All BASIS ,
i 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 Wnt release rate was determined using default X/Q values (g from the ODCM which pmvides a less accurate method of evaluation release magnitude then EAL - 6.1.4.d Rev.00 Page 1 of 3
HCGS EAllRALTechnical Basis using dose assessment with real time meteorological data. For that reason, this EAL should not O be utilized if Dose Assessment is available. Dose Assessment will take in account actual meteorological conditions, plant vent flows and plant vent efnuent concentrations to provide a
- more accurate assessment of a radiological release. If Dose Assessment is available than refer
! to EAL 6.1.4.a for classification.
l The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPV, FRVS and HTV noble gas readings.
It is not intended that the release be averaged over 30 minutes but that the Release Rate exceed the EAL value for > 30 minutes. .
Barrier Analysis !
N/A i
ESCALATION CRITERIA NONE ,
DISCUSSION l
To obtain a site specine value to trigger the performance of dose assessment is not necessary, l
O. 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 30 minutes.
The ODCM methodology calculates yearly values. To be consistent with the ODCM methodology the GE classification trigger point of 1000 mrem / hour needs to be converted to a l yearly dose. This is done in the following manner; l
365 days / year
- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day = 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year.
1000 mrem / hour
- 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year = 8.76E+06 mrem / year.
3 ODCM Dose Rate Conversion Factor = 7.80E+03 mrem / year /pCi/m ODCM y./Q = 2.67E-06 Sec/m 3 l
No allocation factor is used for GE. l l
i 8.76E+06 mrem / year = 4.21E+08 pCi/Sec 3
2.67E-06 Sec/m'
- 7.80E+03 mrem / year /pCi/m ,
j 4.21E+08pCi/sec is the General Emergency Total Noble Gas Release Rate l EAL - 6.1.4.d l Rev. 00 I Page 2 of 3 1
. HCGS EAL/RALTechnical Basis DEVIATION This EAL is based on default meteorological and isotopic mixture assumptions as found in the ODCM. Depending on actual meteorological conditions and isotopic mixture, the Release Rate used as the threshold value in this EAL could produce TEDE values which could be within Alert or Site Area Emergency thresholds. This potential to underclassify this Event is not desirable. To preclude / limit this possibility, PSE&G has used 30 minutes instead of 15 as in AGl.1 This extra 15 minutes would allow personnel to obtain Dose Assessment projections from a second computer should the primary location computer fail. In addition events that result in a release of this magnitude would require degradation of multiple Fission Product Barriers and should be promptly classified.
REFERENCES NUMARC NESP-007, AGl.1, AGl.4 OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculatior. Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.
Technical Specification 3.11.2.1 O
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i 6.2 Liquid Effluent Release UNUSUAL EVENT - 6.2.1 l IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 2 Times the Radiological Technical Specifications for 60 minutes or longer !
~
EAL I
Valid Cooling Tower Blowdown Effluent Radiation Monitor High Alarm Condition AST 1
Sample analysis of liquid effluent indicates concentration in excess of l
2 times Technical Specification limits AND Release continues for .;>_60 minutes after the alarm occurs l
OPERATIONAL CONTITION - All '
BASIS A Valid Cooling Tower Blowdown Effluent Ra'diation Monitor High alarm condition corresponds to the Technical Specification Liquid Effluent Release Limit . Despite this limit being below the EAL threshold, exceeding this limit with a failure to terminate the discharge may be a precursor to an Unplanned Liquid Radiological Release in excess of 2 times Technical Specifications that continues for greater than 60 minutes. The threshold for this l EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude, that is not isolated in 60 minutes. The final integrated dose is very low and is not the primary concern. Valid is defined as the l Cooling Tower Blowdown Effluent Radiation Monitor High Alarm actuating specifically due l to a Liquid Release exceeding the Technical Specification limit, thus precluding unwarranted l event declaration as the result of spurious actuation. Unplanned is defined as any release for j which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit.
- p It is not intended that the release be averaged over 60 minutes, but exceed 2 times the
'Q Technical Specification limit for 60 minutes or longer. In addition, it is intended that the event EAL - 6.2.1 Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to an Alert when the Liquid Effluent Release exceeds :
200 times Technical Specification limits.
DISCUSSION The Cooling Tower Blowdown Effluent Radiation Monitor (9RX506) monitors radioactivity in the cooling tower blowdown before it is discharged into the Delaware River and warns personnel of an excessive amount of radioactivity (greater than Technical Specification limits) !
being released to the environment.
DEVIATION '
None
- REFERENCES NUMARC NESP-007, AUI.2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effluents Technical Specifications LCO 3.11.1.1
- HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response i
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L) 6.2 Liquid Efiluent Release j i
l ALERT - 6.2.2 IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Technical Specifications for 15 minutes or longer
~
EAL 1
Valid Cooling Tower Blowdown Ef0uent Radiation Monitor High Alarm Condition AND Sample analysis of liquid effluent indicates concentration in excess of 200 times Technical Specification limits AND l
g Release continues for .;>_15 minutes after the alarm occurs
()
OPERATIONAL CONDITION - All l BASIS A Valid Cooling Tower Blowdown Efnuent Radiation Monitor High alarm condition corresponds to the Technical Specification Liquid Ef0uent Release Limit . Despite this limit being well below the EAL threshold, exceeding this limit with a failure to terminate the discharge may be a precursor to an Unplanned Liquid Radiological Release in excess of 200 times Technical Specifications that continues for greater than 15 minutes. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control i implied by a radiological release of this magnitude, that is not isolated in 15 minutes. The l release duration was reduced from 60 minutes (UE) to 15 minutes in recognition of the increased severity of a release of this magnitude. Valid is defined as the Cooling Tower Blowdown Efauent Radiation Monitor High Alarm actuating snecifically due to a Liquid Release exceeding the Technical Specification limit, thus precluding unwarranted event declaration as the result of spurious actuation. Unplanned is defined as any release for which i 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 j Technical Specification limit for 15 minutes or longer. In addition, it is intended that the event EAL - 6.2.2 Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis be declared s soon as it is determined that the release will exceed 200 times the limit for 15 minutes or longer.
Barrier Analysis N/A ESCALATION CRITERIA N/A -
i DISCUSSION The Cooling Tower Blowdown Effluent Radiation Monitor (9RX506) monitors radioactivity in the cooling tower blowdown before it is discharged into the Delaware River and warns personnel of an excessive amount of radioactivity (greater than Technical Specification limits) i being released to the environment.
i DEVIATION l l
None !
REFERENCES NUMARC NESP-007, AA1.2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effluents HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response l
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I HCGS EAURALTechnical Basis g 6.0 Radiological Releases / Occurrences ~
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6.3 In-Plant Radiation Occurrences l
l UNUSUAL EVENT - 6.3.1.a l
IC Unplanned Increase in Plant Radiation
! EAL' I
l Unplanned increase in radiation levels inside the Protected Area .>.1000 times normal l as indicated by EITHER one of the following:
I Permanent or portable Area Radiation Monitors l General Area Radiological Survey )
i OPERATIONAL CONDITION - All l
BASIS l
O V An Unplanned increase in radiation levels within the Protected Area by a factor of 1000 times over normal represent a degradation in the control of radioactive material and a potential degradation in the level of safety of the plant. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are increasing in an uncontrolled manner. This condition specifically represents an uncontrolled l increase in radiation levels within the Protected Area. Planned evolutions which cause elevated radiation levels do not warrant classification under this EAL.
Barrier Analysis l
N/A ESCALATION CRITERIA Emergency Classification will escalate to an Alert (6.3.2.a) when radiation levels increase to a level that would impede access to areas required for the safe shutdown of the plant.
DISCUSSION i
Normal level is considered as the highest reading in the past 24-hours excluding current peak values. RM-11 computer trends, RMS strip charts, and/or SPDS can be used to confirm
' V,q these values.
EAL - 6.3.1.a Rev. 00 Page 1 of 2
. HCGS EAL/RALTechnical Basis l
Examples of a planned evolution that results in increased radiation levels within the Protected Area include, but are not limited to:
Radiography Lifting of the Reactor Vessel Moisture Separator / Dryer during Refuel Operations Performance of a TIP trace Relocation of radioactive materials, including radioactive waste DEVIATION -
NUMARC IC AU2 includes unexpected increases in Airborne concentration in addition to plant radiation. The corresponding Hope Creek IC does not address Airborne concentration, since an increase in Airborne concentration is not addressed in the example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked.
i REFERENCES NUMARC NESP-007, AU2.4 l
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EAL - 6.3.1.a l Rev. 00 Page 2 of 2
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. HCGS EAL/RALTechnical Basis l
6.0 Radiological Releases / Occurrences i
- O 6.3 In-Plant Radiation Occurrences UNUSUAL EVENT - 6.3.1.b IC Unplanned Increase in Plant Radiation EAL I
Uncontrolled water level decrease in the Reactor Cavity as indicated by l EITHER one of the following: ,
I
. Visual Observation
- Reactor Water Level Shutdown Range Indicator IBBLI-R605 OPERATIONAL CONDITION - 5 BASIS An Uncontrolled lowering of Reactor Cavity Level during Refueling (Operational Condition
- 5) represents a condition which can result in increased radiation levels, due to the loss of radiation shielding, if the Reactor Cavity level decrease can not be terminated. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Uncontrolled means that the level decrease can not be terminated.
I Determination of an uncontrolled level decrease is made through either Visual Observation or l indication in the Main Control Room. Visual Observation is the preferred method, whenever possible, however it is N_QIintended that an individual must be dispatched for classification purposes, if the existing radiation level increase trend prevents personnel from accessing the Refuel Floor, or if cameras are available to remotely verify the condition. In the event visual observation is not available by any means, then Main Control Room indication should be used.
l Barrier Analysis j N/A l
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Rev. 00 ,
Page 1 of 2 l
. HCGS EAIJRALTechnical Basis ESCALATION CRITERIA -
c Emergency Classification will escalate to an Alert as a result of uncovery of a fuel ssembly and/or indication of high radiation levels on the refueling floor.
DISCUSSION During Refueling operations, the RPV is flooded and RPV level indication is monitored on the 1 shutdown instrument range. Limitations on evolutions on with a potential for draining the :
RPV are imposed when refueling is in progress. Lowering of RPV level may result in tire loss )
of Shutdown Cooling if RPV level continues to lower unchecked. This may result in the loss of decay heat removal from the fuel contained in the RPV.
l Technical Specifications requires at least 22 feet 2 inches of water be maintained over the top t of the reactor pressure vessel flange while in Operating Condition 5 and either fuel assemblies !
are being handled or the fuel assemblies seated within the reactor vessel are irradiated. The Technical Specification minimum water level in the Reactor Vessel under these conditions is I based on the minimum water level required to remove 99% of the assumed 10% iodine gap activity that would be released from the rupture of an irradiated fuel assembly.
DEVIATION C 1) NUMARC states that this EAL will be applicable in all modes of operation. In other than Operational Condition 5, the RPV head will be fully tensioned, and lowering of vessel level would be classified by EALs in Section 3.0, Fission Product Barriers, or Section 8.1, Loss of Heat Removal Capability.
to plant radiation. The corresponding Hope Creek IC does not address Airborne l concentration, since an increase in Airborne concentration is not addressed in the '
example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked.
REFERENCES NUMARC NESP-0007, AU2.1 HC.OP-AB.ZZ.0142 (Q), Loss of Shutdown Cooling l HC.OP-AB.ZZ-0144 (Q), Loss of Fuel Pool Inventory / Cooling HC.OP-AB.ZZ-0101 (Q), Irradiated Fuel Damage HC.OP-AB.ZZ-126 (Q), Abnormal Release of Gaseous Radioactivity HCGS Technical Specifications Section 3/4 9.8 O EAL - 6.3.1.b Rev.00 Page 2 of 2
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. HCGS EAL/RALTechnical Basis
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6.0 Radiological Releases / Occurrences v
6.3 In-Plant Radiation Occurrences UNUSUAL EVENT - 6.3.1.c IC Unplanned Increase in Plant Radiation EAL -
Uncontrolled water level decrease in the Spent Fuel Pool as indicated by Valid Fuel Pool Low Level Alarm Condition AND Visual Observation l
OPERATIONAL CONDITION - All BASIS An Uncontrolled decrease in Spent Fuel Pool I2 vel represents a condition which can result in increased radiation levels, due to the loss of radiation shielding, if the Spent Fuel Pool level decrease can not be terminated. This event has a long lead time relative to potential for l radiological release outside the site boundary, thus the impact to public health and safety is l l very low. Uncontrolled means that the level decrease can not be terminated.
Determination of an uncontrolled level decrease is made through receipt of the Spent Fuel Pool low Level Alarm in the Main Control Room and Visual Observation.
Barrier Analysis !
N/A i ESCALATION CRITERIA Emergency Classification will escalate to an Alert as a result of uncovery ofirradiated fuel as 3
indicated by high radiation levels on the refueling.
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I EAL - 6.3.1.c Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis DISCUSSION -
O Normal Spent Fuel Pool level is at 40' of water in the pool. This level provides approximately 25' of water above the top fuel stored in pool, and 9' of water above fuel in transit. The low level alarm is set at 39' 9". This is above, but approaching the Technical Specification.
minimum required water level of 23 feet over the top of irradiated fuel assemblies seated in the spent fuel pool storage racks. The Technical Specification minimum water level in the Spent Fuel Pool is based on the minimum inventory and level required to remove 99% of the assumed 10% iodine gap activity that would be released from the rupture of an irradiated fuel assembly. -
To prevent accidental draining of the Spent Fuel Pool, no piping connections are made to the .
fuel pool below the normal water level. The spent fuel pool cooling water return lines are provided with vacuum breakers to prevent water from being siphoned out of the fuel pool should a break occur in one of these lines. The skimmer surge tanks receive the overflow from the spent fuel pool and serve as the suction source to the fuel pool cooling pumps.
Lowering of level in the skimmer surge tank will result in isolation of the pool filter demineralizers. This will result in the loss of the fuel pool cooling pumps. Subsequent heating of the water in tile spent fuel pool may occur depending on the heat load present.
DEVIATION O
O NUMARC IC AU2 includes unexpected increases in Airborne concentration in addition to plant radiation. The corresponding Hope Creek IC does not address Airborne concentration, since an increase in Airborne concentration is not addressed in the example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborr.e concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked.
REFERENCES NUMARC NESP-0007, AU2.2 HC.OP-AR.ZZ-0014(Q), Annunciator Response Procedures, Window D3-A5 (D3834) '
HC.OP-AB.ZZ-0144 (Q), Loss of Fuel Pool Inventory / Cooling HC.OP-AB.ZZ-0101 (Q), Irradiated Fuel Damage HC.OP-AB.ZZ-126 (Q), Abnormal Release of Gaseous Radioactivity HCGS Technical Specifications Section 3/4 9.9 HCGS UFSAR, Section 9.-2.2.2 O
EAL - 6.3.1.c Rev.00 1 Page 2 of 2
HCGS EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences
~
O 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 increase in radiation levels inside the Protected Area l 1000 times normal as indicated by EITHER one of the following:
- Permanent or portable Area Radiation Monitors
- General Area Radiological Survey AND Unplanned Dose Rates 12000 mrem /hr in ANX area of the plant which require
( ACCESS to maintain plant safety functions (EXCLUDING the Main Control Room and CAS)
OPERATIONAL CONDITION - All BASIS An Unplanned Dose Rate of 2000 mrem /hr or greater in ANY area of the plant which requires ACCESS to maintain plant safety functions, warrants declaration of an Alert, due to the impaired ability to operate the required plant equipment. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are increasing in an uncontrolled manner. The Dose Rate threshold of 2000 mrem /hr was chosen based upon NC.NA-AP.ZZ-0024, Radiation Protection Program Administrative Dose Limits and Extension criteria which requires Senior Radiation Protection Supervisor approval prior to exceeding 2000 mrem /yr. This value is low enough to ensure classification of an Alert before personnel access is severely hampered and high enough to allow any increase in normal radiation level, by a factor of 1000, to be classified as an Unusual Event per EAL 6.3.1.a. Radiation levels could be indicated by ARM or radiological survey.
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EAL - 6.3.2.a Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis Barrier Analysis -
O N/A 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 judgement must be used, based on existing plant conditions, to determine areas that contain systems that are required to be operated manually, or require local surveillances to assure reliable support of safe plant operation for the conditions that exist.
Areas having equipment that must be operated locally during an accident and areas along associated access routes that require HP coverage and continuous update of changing radiological conditions satisfy the definition of this condition.
Areas of the Plant which require access following an accident to maintain plant safety functions include but are not limited to: ,
Reactor Core Isolation Cooling (RCIC) system areas 1
(~_ Standby Liquid Control (SLC) system areas s N l Residual Heat Removal (RHR) system areas Diesel Generators and Adjacent Areas l Service Water System areas i Station Auxiliary Cooling (SACS) system areas )
Areas covered in the HC.OP-EO.ZZ-300's (300 series EOPs)
DEVIATION None REFERENCES NUMARC NESP-007, AA3.2 ;
NC.NA-AP.7.Z-024(Q) Radiation Protection Program i
O EAL - 6.3.2.a Rev.00 Page 2 of 2
HCG 5 EAL/RALTechnical Basis 6.0 Radiological Releases / Occurrences -
(J 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 Dose Rates .;>_.15 mrem /hr in EITHER one of the following:
- Main Control Room
+ Security Central Alarm Station (CAS)
OPERATIONAL CONDITION - All I
es BASIS I '
An Unplanned Dose Rate of greater than or equal to 15 mrem /hr represent a condition which would jeopardize continuous occupancy of the Control Room or Security CAS, and warrants declaration of an Alert. It is the impaired ability to operate the plant that results in the actual or potential substantial oegradation of the level of safety of the plant. In addition, unplanned increases in plant radiation levels represent a degradation in the control of radioactive materials and represent a degradation in the level of safety of the plant. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are increasing in an uncontrolled manner. Radiation levels can be determined by ARM or radiological survey.
Barrier Analysis N/A l ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant off-site doses.
f3 EAL - 6.3.2.b l Rev.00 I
Page 1 of 2
, HCGS EAURALTechnical Basis DISCUSSION - -
The Control Room and Security Central Alarm Station general area radiation level threshold is set at 15 mr/hr and was chosen because continuous occupancy is required. This is consistent with General Design Criteria 19, which addresses continuous occupancy of the Control Room for 30 days after an accident.
The Security Secondary Alarm Station (SAS) was excluded because it is fully redundant to the Security CAS. For a radiological event, SAS would be evacuated, with all Security functions performed by the CAS. -
Events which require Control Room evacuation will be classified per ECG Section 8.
DEVIATION :
None REFERENCES NUMARC NESP-007, AA3.1 10CFR50 NUREG/CR-4982 NEC Information Notice 08
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- a a - ..- a.. - r.-a.. . ._ a an. ..a a, w _ _
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. HCGS EAL/RALTechnical Basis ,
i 6.0 Radiological Releases / Occurrences -
i f)
V l 6.3 In-Plant Radiation Occurrences ALERT - 6.3.2.c i IC Major Damage to Irradiated Fuel EAL _
Major Damage to Irradiated Fuel has occurred AND !
i Valid High Alarm received from AN_Y one of the following RMS channels:
t Refuel Floor Exhaust Channel A (9RX627) ;
e Refuel Floor Exhaust Channel B (9RX628)
Refuel Floor Exhaust Channel C (9RX629) oO V OPERATIONAL CONDITION - All
]
BASIS Major Damage to an irradiated fuel bundle that result in a High Refuel Floor Exhaust 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 actual release of fission products from an irradiated Fuel Bundle, due to physical damage. Events that result in increased radiation levels due to shine, as a result of decreased shielding, but do not involve a release of fission products should not be classified under this EAL, but should be classified EAL 6.3.2.d, when those conditions exist.
Major Damage is defined as physical damage to an Irradiated Fuel Bundle that results from !
either dropping or physical contact with other components in the Fuel Pool, such that the I magnitude of the damage soecifically results in actuation of a Refuel Floor Exhaust High Radiation Alarm. Valid is defined as the High alarm occurring as a result of the damage to the irradiated fuel bundle which results in an actual uncontrolled rel; ease of fission products l from the cladding.
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! EAL - 6.3.2.c i Rev.00 Page 1 of 2
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- HCGS EAL/RALTechnical Basis ,
I i
Barrier Analysis -
.G N/A ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant offsite doses.
- . DISCUSSION The Refuel Floor Exhaust Rad Monitors are Process Monitors and are designed to detect a ;
release of Fission Products to the Reactor Building atmosphere. Hence, they are included as !
part of the EAL threshold, to confirm the magnitude of damage to an irradiated fuel bundle.
These monitors can also react as Area Radiation Monitors, in the event of increasing radiation levels due to decreased shielding, as would occur during a loss of Fuel Pool inventory event. It is important to distinguish between the cause for increased radiation levels when classifying an event under this EAL. ;
DEVIATION None ;
- O REFERENCES j U NUMARC NESP-007, AA2.1 l
l l HC.OP-SO.SM-0001(Q), Isolation Systems Operation l
HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation j l HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response, Att. 54,55,56 HCGS Technical Specifications,3.3.2 Table 3.3.2-2 HCGS-UFSAR, Section 11.5.2 NUREG/CR-4982 i NRC Information Notice 08 l
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. HCOS EAL/RALTechnical Basis !
6.0 Radiological Releases / Occurrences -
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.rvO) ' ,
6.3 In-Plant Radiation Occurrences ALERT - 6.3.2.d/6.3.2.e !
IC Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel ,
~
EAL i
I EITHER one of the following:
- Unplanned increase on AND' one of the following Area Rad Monitors or by general area rad survey indicates _> 2000 mrem /hr:
l
- Spent Fuel Storage Pool Area (9RX707)
- New Fuel Criticality Storage Channel A (9RX612)
- New Fuel Criticality Storage Channel B (9RX613) l n
- Visual observation of Irradiated Fuel uncovered ,
U 1 OPERATIONAL CONDITION - All BASIS i
An Unplanned Dose Rate of 2000 mrem /hr as indicated on any of the Refuel Floor Area )
Radiation Monitors warrants declaration of an Alert, as dose rates of this magnitude could be the result of a loss of shielding of irradiated Fuel Bundles or possible damage to an irradiated Fuel Bundle. Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate. The intent of this EAL is to classify those events that result in increased Dose Rates on the Refuel Floor.
Specifically, those events that result in increased radiation levels due to shine, as a result of decreased shielding, but do not involve a release of fission products should be classified under l this EAL. Those events that result in physical damage to an irradiated and are accompanied by increasing radiation levels should not be classified under this EAL, but should be classified L . EAL 6.3.2.c, when those coaditions exist.
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Unplanned is defined as those events or conditions which are not associated with a planned evolution, such as lifting of the Reactor Vessel Internals, that results in radiation levels are n increasing in an uncontrolled manner. The Dose Rate threshold of 2000 mrem /hr was chosen r EAL - 6.3.2.d Rev.00 Page 1 of 2
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l l , HCGS EAURALTechnical Basis i based upon NC.NA-AP.7Z-0024, Radiation Protection Program Administrative Dose Limits
' r^S and Extension criteria which requires Senior Radiation Protection Supervisor approval prior to I
C exceeding 2000 mrem /yr. This value is low enough to ensure classification of an Alert before personnel access is seveiely hampered and high enough to allow any increase in normal radiation level, by a factor of 1000, to be classified as an Unusual Event per EAL 6.3.1.a.
Radiation levels could be indicated by ARM or radiological survey.
Visual observation of irradiatea fuel uncovered will result in onsite dose levels changing significantly.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of ,
radioactive materials causes significant offsite doses.
DISCUSSION The Refuel Floor Area Radiation Monitors are designed to detect an increased radiation level p on the Refuel Floor. Hence, they are included as part of the EAL threshold, to determine the i magnitude of a loss of shielding to irradiated Fuel Bundles. Actual Damage to an irradiated fuel bundle will also cause an increase in these Area Radiation Monitors, however the Refuel Floor Exhaust Rad Monitors are specifically designed to detect the actual release of fission products to the atmosphere. It is important to distinguish between the cause for increased radiation levels when classifying an event under this EAL.
DEVIATION None REFERENCES NUMARC NESP-007, AA2.3, AA2.4 HCGS Technical Specifications, 3.3.7.1, Table 3.3.7.1-1 HC.RP-AR.SP-0001(Q), Radiation Monitoring System Alarm Response, Att. 41,42,77 NUREG-1229, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions For Nuclear l Incidents NRC Information Notice 08 (h
EAL - 6.3.2.d Rev.00 Page 2 of 2 I
, HCGS EAL/RALTechnical Basis 7.0 Electrical Power -
7.1 Loss of AC Power Capabilities UNUSUAL EVENT - 7.1.1 4
IC Loss of All Offsite Power to Vital Buses for greater than 15 minutes EAL _
Unplanned Imss of Power from Station Service Transformers 1 AX501 AND 1BX501 to 4
ALL 4.16 KV Vital Buses i AND 1
> 15 minutes have elapsed !
l OPERATIONAL CONDITION - All BASIS 1
/]
C An Unplanned Loss of Power from Station Service Transformers 1 AX501 AND 1BX501 (Offsite Power Sources) to the 4.16 KV Vital Buses for greaar than 15 minutes, reduces required plant redundancy and potentially degrades the level of safety by increasing plant vulnerability to a complete loss of all Vital AC power. Reliance on the EDGs to energize the ,
Vital Buses represents a significantly abnormal condition. The intent of the EAL is to classify I an Unusual Event when the EDGs are being used to energize their respective Vital Buses, due to a loss of the offsite power sources. In the case where one or more EDGs are unavailable or ;
fail to start for any reason, following the loss of the offsite power sources, an Unusual Event is warranted until onIV one Onsite or Offsite Power Source remains energized, such that the loss of this energized source would result in a complete loss of all 4.16 KV Vital Power. 15 minutes was chosen to exclude transient or momentary power losses and to allow restoration of available sources. Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities.
Although no fission product barriers are directly affected by the loss of the offsite power sources to the Vital 4.16 KV buses, the heat addition to the Primary Containment combined with heat removal capability dependent on Emergency Diesel Generator operation, warrants classification as an Unusual Event, since it is potential precursor to more serious conditions.
Barrier Analysis N/A EAL - 7.1.1 Rev.00 Page 1 of 3
, JCGS EAL/RALTechnical Basis ESCALATION CRITERIA -
q Emergency Classification will escalate to an Alert based on a Loss of Offsite Power to Vital 4.16 KV buses coincident with Onsite AC power being reduced to a single Vital 4.16 KV Bus.
(Operational Conditions 1,2, and 3); or having a Loss of all Offsite and Onsite AC power in Operational Conditions 4 or 5.
DISCUSSION Hope Creek normally has three physically separate , independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows:
500 KV Hope Creek - Salem Crosstie line.
The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3.
The New Freedine Line, referred to as the 5023 line, is a 42.9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County),
which feeds the 500 KV Switchyard Bus Section 5.
p V Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers 1 AX501 and 1BX501 normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either 1 AX501 or IBX501. Two of the four Vital Buses are normally provided power from 1 AX501 with alternate power from 1BX501; the other two are normally supplied power from 1BX501 with alternate power from 1 AX501. Less of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the altemate source, provided power is available.
Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EDG starts are initiated on degraded power conditions on both 1 AX501 and IBX501, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sus:ained loss of power).
DEVIATION None O
EAL - 7.1.1 Rev.00 Page 2 of 3
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. HCGS EAURALTechnical Basis .
! REFERENCES -
NUMARC NESP-0007, SU1 HC.OP-AB.ZZ.0135 (Q), Station Blackout // loss of Offsite Power // Diesel Generator Malfunction HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0102 (Q)-FC, Primary Containment Control i HCGS Technical Specifications 3/4.8, Electrical Power Systems l
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, HCGS EAIJRALTechnical Basis 7.0 Electrical Power -
7.1 Loss of AC Power Capabilities ALERT - 7.1.2.a IC AC power capability to Vital Buses reduced to a Single Power Source for greater than 15 minutes such that any additional single failure would result in a complete loss of all 4.16 KV Vital Buses _
EAL Loss of 4.16 ~KV Vital Bus Power Sources (Offsite and Onsite) which results in the availability of ONLY one 4.16 KV Vital Bus Power Source (Offsite or Onsite)
AND
> 15 minutes have elapsed f OPERATIONAL CONDITION - 1, 2, 3 BASIS A degradation of the six 4.16 KV Vital Bus Power Sources, which consirt of the Offsite power sources (I AX501 AND IBX501) and the Onsite power sources (4 EDGs), available to the 4.16 KV Vital Buses, such that a loss of any additional single energized source would result in a complete loss of all 4.16 KV Vital Power, represents a significant challenge to ;
plant safety and are classified under this EAL. These conditions could occur as a result of a l Loss of the Offsite power sources with concurrent failure of all but one EDG to supply power i to its Vital Bus, or due to a failure of all EDGs concurrent with the Offsite power sources i reduced to a single source (even though all 4.16 KV Vital Buses may still be energized). These conditions reduce redundancy and potentially degrade the level of safety by increasing plant vulnerability to a complete Loss of Vital AC power. The intent of this EAL is to classify an Alert in those conditions in which a loss of a single power source to the 4.16 KV Vital Buses would result in the loss of All 4.16 KV Vital power. Availability is defined as a power source that can be aligned to provide power to the bus within 15 minutes. This includes the power source, as well as, all required breakers needed to provide power.15 minutes was chosen to ,
exclude transient or momentary power losses.
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O EAL - 7.1.2.a Rev.00 4 Page 1 of 3
? i i i
. HCGS EA1JRALTechnical Basis 1 Barrier Analysis .
l
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! V N/A l FECALATION CRITERIA I l l
Emergency Classification will escalate to a Site Area Emergency based on a Loss of Power to I
, all 4.16 KV Vital Buses for > 15 minutes.
l DISCUSSION -
l Hope Creek normally has three physically separate , independent 500 KV transmission lines,
! connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network j
( (PJM). The three sources are as follows ,
500 KV Hope Creek - Salem Crosstie line.
The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Scraion 3. -
The New Freedom Line, referred to as the 5023 line, is a 42.9 mile tie to the New
]v Freedom Switching Station (located northeast of Hope Creek in Camden County),
which feeds the 500 KV Switchyard Bus Section 5.
l Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers 1 AX501 and IBX501 normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either I AX501 or 1BX501. Two of the four Vital Buses are normally provided power from 1 AX501 with alternate power from l
1BX501; the other two are normally supplied power from 1BX501 with alternate power from i
1 AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available.
Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its l associated Vital Bus. Additional automatic EDG starts are initiated on degraded power conditions on both 1 AX501 and IBX501, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained loss of power).
. DEVIATION None
(]N EAL - 7.1.2.a Rev.00 Page 2 of 3
, HCGS EAURALTechnical Basis REFERENCES -
l NUMARC NESP-0007, SA5 i HC.OP-AB.ZZ.0135 (Q), Station Blackout / Loss of Offsite Power / Diesel Generator l Malfunction l HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0102 (Q)-FC, Primary Containment Control HCGS Technical Specifications 3/4.8, Electrical Power Systems l
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. . HCGS EAL/RALTechnical Basis 1
7.0 Electrical Power -
A l
O 7.1 Loss of AC Power Capabilities l
l 4LERT - 7.1.2.b l
IC Loss of All Offsite Power and All Onsite AC Power to 4.16 KV Vital Buses during either Cold Shutdown or Refueling for greater than 15 minutes EAL l l
t l )
ALL 4.16 KV Vital Buses are deenergized AND i
l > 15 minutes have elapsed OPERATIONAL CONUITION - 4, 5, Defueled g BASIS
]
() A Loss of ALL 4.16 KV Vital Buses that occurs while the plant is in either Cold Shutdown or j
Refueling conditions, results in a compromise of plant systems. The intent of this EAL is to classify degraded AC power events that result in a Loss of Offsite power sources (I AX501 AND IBX501) to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs). With the plant in Cold Shutdown or Refueling, the reduced decay heat, and lower Reactor Coolant temperatures and pressures, increases the time available to restore one of the
- Vital Buses before Fission Product Barriers are threatened relative to classification of this condition in Operational Conditions 1,2, or 3. Thus this condition is classified as an Alert.
15 minutes was chosen to exclude transient or momentary power losses.
j Barrier Analysis l
N/A l
l ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on Radiological Release (EAL Section 6.0), or on the long term inability to remove Decay Heat (EAL Section !
8.0).
EAL - 7.1.2.b Rev.00 ,
Page 1 of 3 l
i
l HCGS EAURALTechnical Basis l
DISCUSSION -
- Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC l electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water.
Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged.
During a Imss of all AC power to the Vital Buses, all Class lE System Instruments remain powered from Class 1E Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers.
l Normally, Hope Creek has three physically separate , independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows:
500 KV Hope Creek - Salem Crosstie line.
L The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching l Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3.
The New Freedom Line, referred to as the 5023 line, is a 42.9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County),
which feeds the 500 KV Switchyard Bus Section 5.
l Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary l power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers l AX501 and IBX501 normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either I AX501 or IBX501. Two of the four Vital Buses are normally provided power from 1 AX501 with alternate power from IBX501; the other two are normally supplied power from IBX501 with alternate power from 1 AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available.
Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EDG starts are initiated on degraded power
(~~)T L
i EAL - 7.1.2.b
! Rev. 00' Page 2 of 3
- j. ,
HCGS EAL/RALTechnical Basis conditions on both 1 AX501 and IBX501, or under LOCA conditions (EDGs will not .
automatically provide power to the bus unless the bus has a sustained loss of power).
l DEVIATION i None REFERENCES i
NUMARC NESP-0007, sal l HC.OP-AB.ZZ.0135(Q), Station Blackout // Loss of Offsite Power // Diesel Generator Malfunction !
HC.OP.EO.ZZ-0100(Q)-FC, Reactor Scram l HC.OP.EO.ZZ-0102(Q)-FC, Primary Containment Control l HCGS Technical Specifications 3/4.8, Electrical Power Systems l
l l
l I
i l-l D
U
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EAL - 7.1.2.b Rev.00 Page 3 of 3 l
HCGS EAIJRALTechnical Basis
,7.0 Electrical Power -
7.1' Loss of AC Power Capabilities SITE AREA EMERGENCY - 7.1.3 IC Loss of All Offsite Power and All Onsite AC Power to All Vital AC Buses during either Power Operation, Startup or Hot Shutdown for greater than 15 minutes EAL.
ALL 4.16 KV Vital Buses are deenergized AND
> 15 minutes have elapsed OPERATIONAL CONDITION - 1,2,3 BASIS
- A Loss of ALL 4.16 KV Vital Buses that occurs while the plant is in either Power Operation, Startup or Hot Shutdown warrants declaration of a Site Area Emergency due to the compromise to all plant safety systems. The intent of this EAL is to classify degraded AC power events that result in a loss of Offsite power source (IAX501 AND IBX501) to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs). Declaration under this EAL should only occur for a loss of ALL 4.16 KV Vital Buses. Prolonged loss of Vital AC power may cause Core uncovery and the inability to remove heat from the containment. 15 minutes was chosen to exclude transient or momentary power losses. l Barrier Analysis Prolonged loss of AC power has the potential to cause a potential loss or loss of the Fission Product Barriers.
ESCALATION CRITERIA Emergency Classification will be escalate to a General Emergency if the power loss is extended beyond 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or on loss of Fission Product Barriers per EAL Section 3.0.
EAL - 7.1.3 Rev. 00 ;
Page l of 3 j i
" "' ~ , - - , , - - . . _ . . _ _ - _1
, HCGS EAL/RALTechnical Basis DISCUSSION -
O Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable.
Thus, the ability to remove decay heat and control containment parameters is severely challenged.
l During a less of all AC power to the Vital Buses, all Class IE System Instruments remain
! powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC l power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the .
batteries. Battery power is limited depending on the discharge rate and predischarge condition l' l
l of the battery. The ability to restore power to AC buses may eventually be threatened as l battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit {
breakers. )
Normally, Hope Creek has three physically separate , independent 500 KV transmission lines, :
connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows:
500 KV Hope Creek - Salem Crosstic line.
O The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3.
The New Freedom Line, referred to as the 5023 line, is a 42.9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County),
which feeds the 500 KV Switchyard Bus Section 5.
Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical ,
loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary l power systems, via Station Service Transformers which supply Vital and Non-Vital Station Imads. Station Service Transformers l AX501 and IBX501 normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either I AX501 or IBX501. Two of the four Vital Buses are normally provided power from 1 AX501 with alternate power from IBX501; the other two are normally supplied power from 1BX501 with alternate power from 1 AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source.
Additionally, each 4.16 KV Vital !3us has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EDG starts are initiated on degraded power
[
EAL - 7.1.3 Rev.00 Page 2 of 3
HCGS EAL/RALTechnical Basis
~ conditions on both 1 AX501 and IBX501, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained lost of power).
Under a Loss of Vital AC Power condition, operation and control of plant systems is guided i
by the Station Blackout / Loss of Offsite Power / Diesel Generator Malfunction Abnormal Operating Procedure. Successful coping maintains the following key parameters within given acceptable limits:
l 1. Reactor water level > (TAF) l 2. Suppression pool level low enough to prevent HPCI and/or RCIC steam exhaust 4ine flooding
! 3. Reactor pressure high enough to maintain HPCI and RCIC operable l 4. Containment pressure < design limit
- 5. Torus temperature < design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)
- 6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The l parameters listed above can be maintained as long as battery power remains available. Battery power is limited depending on the discharge rate and predischarge condition of the battery.
l The HCGS IPE assum:s that the batteries will be available for four hours, even though the design battery depletion time is six hours. Additionally, the loss of ventilation to the HPCI
! and RCIC turbine areas may result in a system isolation due to elevated temperatures.
Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the j diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC power.
DEVIATION l None REFERENCES NUMARC NESP-0007, SSI HC.OP-AB.ZZ.0135 (Q), Station Blackout / Loss of Offsite Power / Diesel Generator Malfunction HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0102 (Q)-FC, Primary Containment Control i HCGS Technical Specifications Section 3/4.8, Electrical Power Systems
/~N U
EAL - 7.1.3 Rev.00 l Page 3 of 3 i
l l
I l
l HCGS EAL/RALTechnical Basis g 7.0 Electrical Power -
(_)
7.1 Loss of AC Power Capabilities GENERAL EMERGENCY - 7.1.4.a j i
i IC Prolonged Loss of All Offsite and Onsite AC Power to All Vital AC Buses i EAL _
ALL 4.16 KV Vital Buses are deenergized l
AND l l \
l Restoration of Power to at least one 4.16 KV Vital Bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is HQIlikely l
OPERATIONAL CONDITION - 1,2, and 3
! 1 BASIS A Loss of ALL 4.16 KV Vital Buses for a prolonged period of time (> 4 Hours) represents a compromise to all plant safety systems. The intent of this EAL is to classify degraded AC power events that result in a Loss of offsite power source (1 AX501 AND 1BX501) to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs) for greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Prolonged loss of Vital AC power may cause Core uncovery and the inability to remove heat I from the containment. 4 Hours is based on the assumptions of the Station Blackout Coping l
Studies for Hope Creek. Beyond the four hour window, Reactor injection capability may no l longer be available, and degradation in core cooling will commence. However; a General l Emergency should be declared before 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if it can be determined that the power loss ;
cannot be recovered within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or if potential loss or loss of fission product bamers is ;
l imminent. !
l I i Barrier Analysis i
Although not directly related to Fission Product Barriers, these events will eventually i result in the loss of all three barriers if power cannot be restored. In addition, the
( extent of the loss of power will result in degraded monitoring capability. It is therefore important is such events to closely monitor the Fission Product Barriers and use judgement related to the IMMINENT Loss or Potential Loss of barriers as directed in p EAL Section 3.0
- k l EAL - 7.1.4.a R';v. 00 Page 1 of 3 l
l
! HCGS EAL/RALTechnical Basis i
I
[ ,- ESCALATION CRITERIA -
N/A l DISCUSSION 10 CFR 50.2 defines a station blackout (SBO) as complete loss of AC power to Vital AND Non-Vital buses. less of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service l
Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant l
systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may i
be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged.
I During a Loss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC <
l power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as !
l battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers. i Under a Loss of Vital AC Power condition, operation and control of plant systems is guided by the Station Blackout // Loss of Offsite Power // Diesel Generator Malfunction Abnormal Operating Procedure. Successful coping maintains the following key parameters within given acceptable limits:
- 1. Reactor water level > (TAF) i i 2. Suppression pool level low enough to prevent HPCI and/or RCIC steam exhaust line flooding
- 4. Containment pressure < design limit
- 5. Torus temperature < design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)
- 6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The parameters listed above can be maintained as long as battery power remains available. Battery power is limited dependint on the discharge rate and predischarge condition of the battery.
l The HCGS IPE assumes (based on the Coping Study) that the batteries will be available for four hours, even though the design battery depletion time is six hours. Additionally, the loss of ventilation to the HPCI and RCIC turbine areas may result in a system isolation due to l elevated temperatures.
O N.)
EAL - 7.1.4.a Rev.00 Page 2 of 3
1 HCGS EAL/RALTechnical Basis Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the O diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC power.
The likelihood of 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 protect ive 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?
l
- 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 evacaate the 10 mile EPZ. Taking into consideration the above factors, declaring a General Emergency leaves sufficient time for the offsite authorities to implement Protective Actions well before a radioactive release would occur while providing sufficient time for on-site and off-site mitigation activities to restore AC power.
l DEVIATION l
None REFERENCES NUMARC NESP-0007, SGI l
l Station Blackout Coping Studies HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP.EO.ZZ-0104 (Q)-FC, Radioactive Release Control HC.OP-AB.ZZ.0135 (Q), Station Blan out / Loss of Offsite Power / Diesel Generator Malfunction HCGS Technical Specifications Section 3/4.8, Electrical Power Systems HCGS Individual Plant Evaluation, Section 3.1.1.4.6, 3.1.2.1.6 1
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EAL - 7.1.4.a Rev. 00 Page 3 of 3
, HCGS EAURALTechnical Basis q 7.0 Electrical Power -
LJ 7.1 Loss of AC Power Capabilities GENERAL EMERGENCY - 7.1.4.b IC Prolonged Loss of All Offsite and Onsite AC Power to All Vital AC Buses EAL _
ALL 4.16 KV Vital Buses are deenergized AST Loss of any 2 Fission Product Barriers has occurred or is Imminent OPERATIONAL CONTITION - 1,2, and 3 BASIS p
Loss of ALL 4.16 KV Vital Buses may result in Safety System Losses and Fission Product Barrier degradation.. Prolonged loss of Vital AC power may cause Core uncovery and the inability to remove heat from the containment. Reactor injection capability may no longer be available, and degradation in core cooling will commence. Indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with emphasis on EC Judgement as it relates to imminent loss of Fission Product Barrier and because abilities to monitor the barriers is degraded. Imminent is defined as expected to occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Barrier Analysis Although not directly related to Fission Product Barriers, these events will eventually result in the loss of all three barriers if power cannot be restored. In addition, the extent of the loss of power will result in degraded' monitoring capability. It is therefore important is such events to closely monitor the Fission Product Barriers and use judgement related to the IMMINENT Loss or Potenti:ll Loss of barriers as directed in EAL Section 3.0 ESCALATION CRITERIA N/A EAL - 7.1.4.b Rev.00 Page 1 of 3 l
1 J
l HCGS EAL/RALTechnical Basis p\ DISCUSSION U 10 CFR 50.2 defines a station blackout (SBO) as complete loss of AC power to Vital AND Non-Vital buses. Imss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged.
During a IAss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers.
Under a Imss of Vital AC Power condition, operation and control of plant systems is guided by the Stat'.on Blackout //less of Offsite Power // Diesel Generator Malfunction Abnormal g Operating Procedure. Successful coping maintains the following key parameters within given Q acceptablelimits:
- 1. Reactor water level > (TAF)
- 3. Reactor pressure high enough to maintain HPCI and RCIC operable Containment pressure < design limit j 4.
- 5. Torus temperature < design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)
- 6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The parameters listed above can be maintained as long as battery power remains available. Battery power is limited depending on the discharge rate and predischarge condition of the battery.
Additionally, the loss of ventilation to the HPCI and RCIC turbine areas may result in a system isolation due to elevated temperatures.
Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC l power.
The likelihood ofloss of the second Barrier should be based on a realistic appraisal of the
)
situation since a delay in an upgrade decision based on only a chance of mitigating the event EAL - 7.1.4.b j Rev.00 Page 2 of 3
I HCGS EAL/RALTechnical Basis could result in a loss of valuable time in preparing and implementing public protective actions. l O 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, and the loss r
may be mitigated, it is necessary to give the Emergency Coordinator a reasonable idea of how quickly he may need to declare a General Emergency based on these conditions. ,
i It is estimated that several hours are required to fully evacuate the 10 mile EPZ. Taking into consideration the above factors, declaring a General Emergency leaves sufficient time for the l
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 f
REFERENCES NUMARC NESP-0007, SGI HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram .
[
HC.OP.EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP.EO.ZZ-0104 (Q)-FC, Radioactive Release Control i HC.OP-AB.ZZ.0135 (Q), Station Blackout / Loss of Offsite Power / Diesel Generator l Malfunction HCGS Technical Specifications Section 3/4.8, Electrical Power Syste.ns HCGS Individual Plant Evaluation, Section 3.1.1.4.6, 3.1.2.1.6 I
I
- l i
h
!O EAL - 7.1.4.b i l
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, HCGS EAL/RALTechnical Basis 7.0 Electrical Power -
l 7.2 Loss of DC Power Capabilities UNUSUAL EVENT - 7.2.1 !
IC Unplanned Loss of All Vital 125 VDC Power during either Cold Shutdown or Refueling Mode for greater than 15 minutes EAL Unplanned degraded voltage condition for ALL Vital 125 VDC Buses, such that voltage is < 108 VDC l AND ,
l l > 15 rninutes have elapsed i
1 i
OPERATIONAL CONDITION - 4, 5. and Defueled ;
/g~T 9 ASIS An Unplanned degraded voltage condition (< 108 VDC) for ALL Vital 125 VDC Buses for greater than 15 minutes with the unit in Operational Condition 4 or 5 compromises the ability to monitor and control plant functions. The minimum required voltage value is based on the minimum voltage required for Vital 125 VDC bus operability following a battery discharg test per Technical Specification 4.8.2.1.b. Although continued equipment operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed. Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities.15 l minutes was chosen to exclude transient or momentary power losses.
Barrier Analysis i
N/A ESCALATION CRITERIA Emergency Classification will escalate if the condition effects the inability to maintain cold shutdown, based on Loss of Decay Heat Removal Capability EAL section 8.1.
. O EAL - 7.2.1 Rev.00 Page 1 of 2
HCGS EAL/RALTechnical Basis a
DISCUSSION U Vital 125 VDC 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 such as 4.16 KV Breaker Controls, HPCI, RCIC, CS, and RHR pump controls required to maintain safe plant conditions may not operate, and core uncovery with subsequent reactor coolant system and primary containment failure might occur.
In operating condition 4 or 5, a minimum of two of the four DC power channels are required '
by Technical Specifications, including either channel A (10D410) or channel B (10D420).
The loss of one of the required two 125 VDC distribution systems would require that core alterations be suspended, that handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel be stopped.
The design limits of the IE battery banks are as follows:
125 VDC Vital Power:
CHANNEL Switchgear Battery CAPACITY A 10D410 1 AD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B 10D420 IBD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C 10D430 1CD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D 10D440 1DD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> l DEVIATION None REFERENCES NUMARC NESP-0007, SU7 HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-AB.ZZ.0147 (Q), DC System Grounds HC.OP-AB.ZZ.0150 (Q),125VDC System Malfunction HC.OP-AB.ZZ.0151 (Q), + or - 24 Volt DC Malfunction HC.OP- AB.ZZ-0135 (Q), Station Blackout // Loss of Offsite Power // Diesel Generator Malfunction HCGS Technical Specifications Section 3.8.2.2; 3.8.3.2 LCR 93-12, HCGS Technical Specifications Section 4.8.2.1 Revision Request l
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!O EAL - 7.2.1 f
Rev. 00 Page 2 of 2
l HCGS EAIJRALTechnical Basis 7.0 Electrical Power t
7.2 Loss of DC Power Capabilities SITE AREA EMERGENCY - 7.2.3 1 IC Unplanned loss af All Vital 125 VDC Power during either Power Operations, Startup or Hot Shaidown for greater than 15 minutes ,
EAL 1
Unplanned degraded voltage condition for ALL Vital 125 VDC Buses, such that voltage is < 108 VDC AND
> 15 minutes have elapsed OPERATIONAL CONDITION - 1,2, 3
.O BASIS An Unplanned degraded voltage condition (< 108 VDC) for ALL Vital 125 VDC Buses for greater than 15 minutes with the unit in Operational Condition 1,2 or 3 compromises the ability to monitor and control plant functions. The minimum required voltage value is based on the minimum voltage required for Vital 125 VDC bus operability following a battery discharge test per Technical Specification 4.8.2.1.b. Although continued equipment operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed.
Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities.15 minutes was chosen to exclude transient or momentary power losses.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate based on other EALs indicating Radiological Release (EAL Section 6.0) or loss of Fission Product Barriers (EAL Section 3.0).
5 EAL - 7.2.3 Rev. 00 Page 1 of 3
, HCGS EAURALTechnical Basis l l
DISCUSSION O Vital 125 VDC 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 such as 4.16 KV Breaker Controls, HPCI, RCIC, CS, and RHR pump j controls required to maintain safe plant conditions may not operate, and core uncovery with subsequent reactor coolant system and primary containment failure might occur.
less of ADS may create a loss oflow pressure ECCS availability due to the potential inability !
to depressurize the reactor. In addition, loss of these buses will eventually lead to MSIV l closure and reactor scram due to the loss of the Primary Containment Instrument Gas (PCIG). l Subsequent to MSIV closure, much of the equipment noted above will be required for plant j stabilization and shutdown. I f
A sustained loss of 125 VDC power will threaten the ability to remove heat from the reactor !
core and from the containment. SRVs will remain operable in the relief mode and the heat [
addition to the containment could result in a loss of the primary containment as a fission j product release barrier. ;
i HPCI and RCIC also require 250 VDC vital power for system operability. Loss of vital 250 i VDC power will only render the associated system inoperable; it does not affect the operability l of the systems listed / discussed above. Loss of all Vital IE 125 VDC power will also render j these systems inoperable for automatic initiation, and from the control room due to loss of control power. The loss Vital IE 250 VDC system requires that HPCI and/or RCIC bc l declared inoperable and the respective Technical Specification LCO be entered. Loss of these j sources is therefore not included in this EAL. !
l The design limits of the IE battery banks are as follows:
125 VDC Vital Power:
Channel Switchgear Battery CAPACITY A 10D410 1 AD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B 10D420 IBD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C 10D430 ICD 411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D 10D440 1DD411 1800 AH at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> In operating conditions 1,2, or 3, the loss of any single channel 125 VDC power source would l require the channel to be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the unit placed in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
O EAL - 7.2.3 Rev.00 Page 2 of 3 l
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. HCGS EAL/RALTechnical Basis DEVIATION -
None ,
REFERENCES l
. NUMARC NESP-0007, SS3 ;
HC.OP.EO.Z7.-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0202 (Q)-FC, Emergency Depressurization HC.OP-AB.ZZ.0147 (Q), DC System Grounds HC.OP-AB.ZZ.0149 (Q),250VDC System Malfunction HC.OP-AB.ZZ.0150 (Q),125VDC System Malfunction l HC.OP-AB.ZZ.0151 (Q), + or - 24 Volt DC Malfunction ;
HC.OP- AB.ZZ-0135 (Q), Station Blackout //1.oss of Offsite Power // Diesel Generator l Malfunction HCGS Technical Specifications Section 3/4.8.2.1, 3/4.8.3.1 LCR 93-12, HCGS Technical Specifications Section 4.8.2.1 Revision Request l
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HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability ALERT - 8.1.2 IC Inability to Maintain the Plant in Cold Shutdown EAL _
Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions AND EITHER one of the following occur:
- RCS Temperature has increased to > 200 F (Excluding a momentary increase > 200 F with heat removal function restored)
Q V
- An UNCONTROLLED temperature increase is RAPIDLY approaching 200 F (with HQ heat removal functions restored)
OPERATIONAL CONDITION - 4,5 BASIS A loss of decay heat removal capabilities necessary to maintain Cold Shutdown conditions could potentially lead to core damage if corrective actions are not implemented. Declaration of an Alert is warranted when ALL Technical Specification required systems are not available to provide Decay Heat Removal functions and can not be restored to prevent boiling in the core. The specification of a temperature INCREASE, rather than specific equipment failures, recognizes the potential for long heatup times providing adequate time for restoration of some form of alternate cooling. The statement " Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions" is intended to represent a complete loss of functions available, or an inadequate ability, to provide core cooling during the Cold Shutdown and Refueling Modes, including alternate decay heat removal methods. This EAL allows for actions taken in accordance with OP-AB.ZZ-0142, Ioss of Shutdown Cooling Abnormal Operating Procedure to reestablish RHR in the Shutdown Cooling Mode or provide for an alternate methods of decay heat removal, with the intent of maintaining RCS temperature below 200'F.
EAL - 8.1.2 Rev.00 )
Page 1 of 3 l l
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. HCGS EAURALTechnical Basis
! For loss of an in-service decay heat removal system with other decay heat removal methods l available, actions taken to provide'for restoration of a decay heat removal function may
! require time to implement. If the event results in RCS temperature momentarily (not to exceed 15 minutes) rising above 200*F with heat removal capability restored, Emergency Coordinatorjudgement will be required to determine whether heat removal systems are ;
adequate to prevent boiling in the core and restoration of RCS temperature control. l 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 l lost. Unplanned is defined as a condition that is not due to scheduled operations or l
maintenance activities, in which an RHR system is intentionally removed from service.
Barrier Analysis
)
N/A ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on inability to maintain RPV Water level above the Top of the Active Fuel, or increased Radiological Releases.
i
'V (3 DISCUSSION l i
The Residual Heat Removal (RHR) system provides the normal method for decay heat removal operating in the Shutdown Cooling Mode. With RHR unavailable for shutdown cooling operation, (including the loss of SACS and/or service water which supply cooling water to the RHR heat exchangers), alternate decay heat removal system can be aligned to control decay
]
heat. An unavailability of these systems, can result in a gradual increase in reactor coolant l temperature to the values specified in this EAL. The rate of increase in coolant temperature i would be dependent on the amount of decay heat present. The threshold for this EAL is the RCS temperature transition value between Operational Condition 4 and Operational Condition 3.
Procedural guidance is provided to establish an alternate method of decay heat removal. These alternate methods include: aligning Reactor Water Cleanup system (RWCU), with maximum RACS aligned to the Non-regenerative heat exchanger; aligning condensate transfer via the
80"; maximizing fuel pool cooling if the RPV head is removed and the reactor cavity flooded; using the "C" RHR pump crosstied to the "A" RHR loop.
l l If these alternate means are unavailable, or ineffective, decay heat removal must be
! accomplished by feed-and -bleed using ECCS systems and discharging steam to the
/ Suppression Pool via the SRVs to the Suppression Pool.
EAL - 8.1.2 Rev. 00 Page 2 of 3 l
- . - _ . _ ~ . . . . - - - . - . . . - - - - . .- .
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. HCGS EAURALTechnical Basis DEVIATION -
None REFERENCES NUMARC NESP-0007, SA3 NUMARC Questions and Answers, June 1993, "Sptem Malfunction Question #6b" HC.OP-AB.ZZ-0142 (Q), Loss of Shutdown Coolinh :
l HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control -
- HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NEDO-2121, Supplement A to BWR Owners Group ,
Emergency Procedure Guidelines, Revision 4 l HCGS Technical Specifications Sections 3/4.3, 3/4.4.9, 3/4.7.1, 3/4.7.2 i
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EAL - 8.1.2 Rev.00 ,
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. HCGS EAIJRALTechnical Basis
,- 8.0 System Malfunctions -
8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY - 8.1.3.a IC Loss of Reactor Water Level that has or will Uncover Fuel in the Reactor Vessel EAL Reactor Water Level REACHES -161" (Top of Active Fuel)
OPERATIONAL CONDITION - 4,5 BASIS Reactor Water Level reaching -161" indicates a loss of core submergence. Without core submergence, the integrity of the fuel clad barrier can no longer be assured, even with the reduced decay heat levels in Cold Shutdown and Refuel. This event is classified based on
(')
U reaching the Reactor Water level threshold (instead of being able to restore and maintain above the threshold) due to the potentially severe consequences of a loss of core submergence. Since the design of the normal and emergency makeup systems would preclude this condition, an extreme challenge to their ability to provide core cooling by submergence has occurred.
Additionally, ECCS availability and Containment Integrity requirements may be relaxed under these Operational Conditions, thus classification at the Site Area Emergency level is warranted.
Barrier Analysis Fuel Clad Barrier has been potentially lost RCS Barrier has been lost.
ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on abnormal radiological releases.
O(~h EAL - 8.1.3.a Rev. 00 Page 1 of 2
. HCGS EAL/RALTechnical Basis DISCUSSION -
O Core Submergence ensures adequate core cooling. When RPV level decreases to below TAF the ability to effectively remove decay heat can no longer be guaranteed, and the fuel cladding fission product barrier can no longer be considered intact. Sustained partial or total core uncovery can result in clad damage and a significant release of fission products to the reactor coolant. Sustained core uncovery can also result in a breach of the reactor vessel, or an unisolated intersystem LOCA with the RHR System.
DEVIATION -
None REFERENCES ;
NUMARC NESP-0007, SS5 l
HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control '
HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control O
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. HCGS EAL/RALTechnical Basis 8.0 System Malfunctions -
U 8.1 Loss ofIIeat Removal Capability SITE AREA EMERGENCY- 8.I.3.b IC Complete Loss of Functions Needed to Achieve Cold Shutdown Conditions l l
EAL ,, j l
Loss of Main Condenser capabilities, as evidenced by an inability to remove -
Decay Heat from the Reactor AND Loss of Torus capabilities as evidenced by EITHER one of the following:
. Entry into an Unsafe region of ANY of the following curves: 1'
. Heat Capacity Temperature Limit (HCTL) Curve
. Heat Capacity Level Limit (HCLL) Curve j
. Insufficient SRV capacity to reduce RPV pressure l
OPERATIONAL CONDITION - 1, 2, 3 i BASIS A Complete Loss of decay heat removal systems required to ACHIEVE Cold Shutdown conditions from a Hot Shutdown condition, represents a significant challenge to the plant due to the failure of multiple systems designed for the protection of the public. Hence, declaration of a Site Area Emergency is warranted. This EAL specifically includes a degradation of those plant systems required to ACHIEVE a Cold Shutdown condition. It does N_QIinclude an inability to MAINTAIN a Cold Shutdown condition. The inability to MAINTAIN Cold Shutdown Conditions is specifically addressed by EAL 8.1.2. Hence, a Loss of RHR Shutdown Cooling is not included in this EAL. This EAL includes a Loss of Service Water and/or SACS capabilities, based on the effect a loss of these systems has on the ability to maintain Torus capabilities with the Safe Region of the referenced EOP curves. Loss is defined as the systems being unavailable to perform their intended design function. Hence, in the case where the Main Condenser became isolated from the Reactor due to an MSIV EAL - 8.1.3.b Rev.00 Page 1 of 3 l
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. HCGS EAURALTechnical Basis Isolation, but the MSIV could be reopened by procedure, then a Loss of the Main Condenser capabilities has not occurred.
Barrier Analysis l.
N/A l
! ESCALATION CRITERIA i
l Emergency Classification will escalate to a General Emergency based on loss of Fission-Product Barriers or Radiological Releases.
I DISCUSSION j l l l A loss of both the normal heat sink for the reactor, and an impending severe degradation of
- alternate heat removal capability to the Torus. Loss of the heat sink for the reactor when in a i l Hot Shutdown condition will limit the ability to maintain that condition, or to cooldown the l reactor if required. l The Main Condenser can be lost for a variety or reasons; loss of Circulating Water, loss of the I turbine Control and/or Bypass Valve functions, main steam line isolation, etc. With the Main l Condenser not available, heat must be removed from the RCS by the SRVs and absorbed in the Suppression Pool. Loss of the pressure control ability of the SRVs as indicated by the inability to reduce RPV pressure represents a loss of control of a major RCS parameter; which could result in RPV overpressure conditions, or the inability to cooldown if cold shutdown is required. l The Heat Capacity Temperature Limit curve is defined as the highest torus temperature at which initiation of RPV depressurization will not result in exceeding either the suppression pool design temperature or the primary containment pressure limit before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent. The Heat Capacity Level Limit is defined as the higher of either the elevation of the containment downcomer opening or the lowest torus level at which initiation of RPV depressurization will not result in exceeding the Heat Capacity Temperature Limit. Violation of either curve would require an immediate emergency depressurization, thus ensuring that all the heat immediately present in the reactor has been transferred to the containment while maintaining the i
containment within design limits. This represents a serious potential threat to the Primary Containment.
DEVIATION The NUMARC IC associated with EAL SS4 suggests that the IC should include a Complete Loss of Function needed to achieve or maintain Hot Shutdown. The NUMARC basis includes I both reactivity control and decay heat removal. At Hope Creek, as with all other BWRs, the
\
i operator action of placing the Reactor Mode Switch in the Shutdown position that results in EAL - 8.1.3.b Rev.00 Page 2 of 3
. HCGS EAURALTechnical Basis Control Rod inserting into the core such that the Reactor will remain shutdown under all l- conditions without boron, places the Reactor in a Hot Shutdown condition. No additional actions are required to maintain the Reactor in this condition. Systems are required and additional operator actions are required to achieve Cold Shutdown conditions. Based on this, l Hope Creek has modified the NUMARC IC for SS4 to apply specifically to a total loss of decay heat removal, since reactivity control concerns are addressed under the ATWS Section.
l REFERENCES -
NUMARC NESP-0007, SS4 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NEDO-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 3/4.1.3, 3/4.1.5 I
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. HCGS EAL/RALTechnical Basis c 8.0 System Malfunctions i 0 8.2 Loss of Assessment Capability UNUSUAL EVENT - 8.2.I.a IC Unplanned Loss of All Onsite or Offsite Communications Capabilities i l
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)
OPERATIONAL CONDITION - All O BASIS V '
l An Unplanned loss of communication ability significantly degrades the operating crews ability to perform tasks necessary for plant operations and/or the ability to communicate with ;
offsite authorities, warrants dee.lara'. ion of an Unusual Event. The loss of off-site 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 N/A
- ESCALATION CRITERIA l
l None 1
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l DISCUSSION None O
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( EAL - 8.2.1.a l l
Rev.00 Page 1 of 2
. HCGS EAURALTechnical Basis DEVIATION -
O None REFERENCES l 1
NUMARC NESP-0007, SU6 l
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. HCGS EAL/RAI Technical Basis 8.0 System Malfunctions -
(~)
\,3 l 8.2 Loss of Assessment Capability UNUSUAL EVENT - 8.2.I.b IC Unplanned 1.oss 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)
. Essex (Centrex) Phone System NAWAS s.
- EMRAD e FTS 2000 1
OPERATIONAL CONDITION - All l-BASIS l
l An Unplanned loss of communication ability significantly degrades the operating crews l 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 off-site 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 N/A l
ESCALATION CRITERIA None DISCUSSION i
l O None EAL - 8.2.1.b Rev.00 Page 1 of 2
1
. HCGS EAL/RALTechnical Basis
- DEVIATION -
lO f
- None j REFERENCES l NUMARC NESP-0007, SU6 i
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. HCGS EAL/RALTechnical Basis
,_ 8,0 System Malfunctions -
O S.2 Loss of Assessment Capability UNUSUAL EVENT - 8.2.1.c 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 Main Control Room Overhead Annunciators AND EmIER one of the following:
- 15 minutes have elapsed since the loss of OHAs e A significant transient is in progress O)
( OPERATIONAL CONDITION - 1, 2, 3 BASIS l l
A Unplanned IAss of > 75% of all Main Control Room Overhead Annunciators without a plant transient in Operational Co_nditions 1,2 or 3 for greater than 15 minutes warrants a heightened awareness by Control Room Operators. Qualification of > 75% is left to the l 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. CRIDS is available to provide compensatory indication.15 minutes is used as a threshold to exclude transient or momentary power losses. l The 15 minutes clock starts when the annunciators have been lost, or are determined to have i 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 had previously been lost for at least 15 minutes but the annunciators were available at the time of discovery, classification is not required under this EAL but a review of the "After The Fact" RAL must be completed. Unplanned loss of annunciators excludes scheduled maintenance and testing activities.
EAL - 8.2.1.c Rev.00 Page 1 of 2
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. HCGS EAURALTechnical Basis Significant transients include response to automatic or manually initiated actions such as:
O
- Reactor Scram
. Load Rejection > 25% Power l
- ECCS Injection l
- Thermal Power oscillations of 10% ,
l Barrier Analysis i
N/A ESCALATION CRITERIA .
Emergency classification will be escalate to an Alert if a transient is in progress or if CRIDS and SPDS becomes unavailable and 15 minutes have elapsed since the loss of OHAs.
DISCUSSION Without Control Room annunciators, there may be difficulty initially recognizing changing plant conditions, as well as, monitoring conditions associated with normal plant operations.
SNSS judgement of the severity of the loss should also be based on the need to initiate increased or continuous plant equipment monitoring. Also, specific annunciator loss should be judged against those needed for by the operating staff for operation in abnormal and !
emergency operating procedures. .
Most alarm conditions for the annunciator system have CRIDS digital alarm points as well. By monitoring the CRIDS screens, most alarm conditions can be observed and responded to independent of the overhead annunciators. ;
)
This EAL is not required in modes 4 or 5 due to the limited number of safety systems required for operation.
DEVIATION A section for declaring an UE has been added if a transient is in progress when the loss of annunicators occurs as requested by the NJ-BNE. Two independent events occurring at the same time warrants a expeditious notification and not waiting 15 minutes for the Unusual l Event declaration. j REFERENCES NUMARC NESP-0007, SU3 HC.OP. AB.ZZ-0143 (Q), Ixss of Overhead Annunciators / Loss of CRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram EAL - 8.2.1.c Rev.00 Page 2 of 2
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. HCGS EAL/RALTechnical Basis l g3 8.0 System Malfunctions -
V 8.2 Loss of Assessment Capability ALERT - 8.2.2.a IC Unplanned Loss of Control Room Annunciators and a Significant Transient is in Progress or Compensatory Indicators are Unavailable EAL Unplanned Loss of > 75% of Main Control Room Overhead Annunciators AST A significant transient is in progress l AND 15 minutes have elapsed since the loss of OHAs
(
C) 1 OPERATIONAL CONTITION - 1, 2, 3 BASIS i i
An Unplanned Loss of > 75% of Main Control Room Overhead Annunciators with a significant transient in progress significantly hampers operator response. Qualification 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. Significant transients include response to automatic or manually initiated actions such as: 1
- Reactor Scram
. Load Rejection > 25% Power l
[
- ECCS Injection
. Thermal Power oscillations of 10%
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 4
O V
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 i
EAL - 8.2.2.a Rev. 00 Page 1 of 2
HCGS EAURALTechnical Basis
[ .
. minutes with the annunciators available at the time of discovery, classification is not required O .under this EAL but a review of the "After The Fact" RAL must be completed. Unplanned loss of annunciators excludes scheduled maintenance and testing activities.
Barrier Analysis N/A ESCALATION CRITERIA.
Emergency Classification will escalate to a Site Area Emergency based on alternate indications are not available with a loss of control room annunciators and a significant plant transient in progress and both a failure of CRIDS and SPDS.
- DISCUSSION Without Control Room annunciators,-it may be difficult to monitor conditions associated with
(. normal plant operations. During transient event such as those listed in the EAL, the difficulty I becomes more acute.
Ioss of control room annunciators significantly reduces the ability of the operations staff to monitor and evaluate plant conditions. SNSS judgement of the severity of the loss should also
, .t
. be based on the need to initiate increased or continuous piant equipment monitoring. Most l- alarm conditions for the annunciator system have CRIDS digital alarm points as well. By
" monitoring the CRIDS screens, most alarm conditions can be observed and responded to,
- independent'of the overhead annunciators. The safety parameter display system (SPDS) also provides information and indication related to selected plant parameters during a plant transient.
This EAL is not required in modes 4 or 5 due to the limited number of safety systems required for operation.
DEVIATION None REFERENCES 1
NUMARC NESP-0007, SA4 i
~HC.OP.AB.ZZ-0143 (Q), Imss of Overhead Annunciators / Loss of CR!DS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram lO EAL - 8.2.2.a j Rev.00 Page 2 of 2
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. HCGS EAURALTechnical Basis 1
8.0 System Malfunctions t
O G
8.2 Loss of Assessment Capability ALERT - 8.2.2.b l
IC Unplanned Imss of Control Room Annunciators and a Significant Transient is in Progress or Compensatory Indicators are Unavailable EAL Unplanned Loss of > 75% of Main Control Room Overhead Annunciators l AND l
l BOTH of the following:
- CRIDS
- SPDS ,
are NOT AVAILABLE i
iq AND lkJ l 15 minutes have elapsed since the loss of OHAs l OPERATIONAL CONDITION - 1, 2, 3 BASIS An Unplanned Loss of > 75% of Main Control Room Overhead Annunciators with loss of backup control room monitoring significantly hampers operator response. Qualification of >
75% is left to the discretion of the Senior Nuclear Shift Supervisor (SNSS), and is considered j approximately 75%. It is not intended that a detailed count be performed, but that a rough approximation be used to determine the severity of the loss.
15 minutes is used as a threshold to exclude transient or momentary power losses. The 15 minutes clock starts when the annunciators have been lost, or are determined to have been lost.
If upon time of discovery it is determined that the annunciators have been lost for at least 15 minutes prior to discovery, classification must be made under this EAL regardless of time required for restoration. If it is determined that the annunciators were lost for at least 15 minutes with the annunciators available at the time of discovery, classification is not required
"""*" "^' '"' '*"i"*""~^"*""*^'""**""*"'
(D' i EAL - 8.2.2.b Rev.00 i
Page 1 of 2 1
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,- HCGS EAL/RALTechnical Basis i
Unplanned loss of annunciators excludes scheduled maintenance and testing activities. The .
fifteen minutes also allows for attempting to restore the CRIDS computer.
-Barrier Analysis l N/A i
ESCALATION CRITERIA 1
Emergency Classification will escalate to a Site Area Emergency based on alternate indications !
are not available with a loss of control room annunciators and a significant plant transient in l progress and both a failure of CRIDS and SPDS.
DISCUSSION i i l The Control Ronm Integrated Display System (CRIDS) is not essential for the safe shutdown j
- or operation of sae plant. However, with the loss of control room annunciators the loss of i l CRIDS significantly reduces the ability of the operations staff to monitor and evaluate plant !
conditions. SNSS judgement of the severity of the loss should also be based on the need to !
- initiate increased or continuous plant equipment monitoring. Most alarm conditions for the j l
annunciator system have CRIDS digital alarm points as well. By monitoring the CRIDS l : screens, most alarm conditions can be observed and responded to, independent of the overhead j j annunciators.
, The safety parameter display system (SPDS) also provides information and indication related !
to selected plant parameters during a plant transient. Ioss of this assessment tool may hamper l operstors attempt to comply with directions provided in EOPs or may limit the recognition of l
- signi5 cant parameter values called out in the EOPs. It is not included in the threshold for this !
EAL because of the limited scope of the parameters it monitors. 'l l
This EAL is not required in modes 4 or 5 due to the limited number of safety systems required - i for operation, j I
i DEVIATION ;
None !
t REFERENCES l NUMARC NESP-0007, SA4
' HC.OP. AB.ZZ-0143 (Q), Loss of Overhead Annunciators / Loss of CRIDS
lO f EAL - 8.2.2.b Rev.00 Page 2 of 2
HCGS EAL/RALTechnical Basis 8.0 System Malfunctions -
\
8.2 Loss of Assessment Capability I
SITE AREA EMERGENCY - 8.2.3 IC Inability to Monitor a Significant Transient in Progress EAL _
Loss of > 75% of Main Control Room Overhead Annunciators M
A significant transient is in progress M
l BOTH of the following:
- CRIDS 1 g-g
- SPDS l y/ are NOT AVAILABLE M
Main Control Room Indications are HQI AVAILABLE to monitor M of the following:
- RCS Status
- Reactivity Control e ECCS
- Containment Parameters OPERATIONAL CONDITION - 1, 2, 3 BASIS An Loss of > 75% of Main Control Room Overhead Annunciators with loss of backup l control room monitoring, AND while a transient is in progress represents a major loss of l ability to properly respond to a transient condition. Quantification of > 75% is left to the l 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. Backup monitoring from CRIDS compounds the EAL - 8.2.3 l Rev.00 l
Page 1 of 2 I
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HCGS EAIJRALTechnical Basis
, ability to monitor the progress of the transient. In addition, a Loss of Main Control Room indications for one of the systems listed in the EAL must also occur. Significant transients '
include response to automatic or manually initiated actions such as: ;
e Reactor Scram
. Load Rejection > 25% Power e ECCS Injection
- Thermal Power oscillations of 10% 1 Barrier Analysis l
N/A ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on either the Loss of Fission Product Barriers; increased plant radiation levels or releases; or EC judgement.
DISCUSSION l Without Control Room annunciators, it may be difficult to monitor conditions associated with normal plant operations. During transient event 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.
The safety parameter display system (SPDS) also provides information and indication related to selected plant parameters during a plant transient. Loss of this assessment tool may hamper operators a. tempt to comply with directions provided in EOPs or may limit the recognition of significan'. parameter values called out in the EOPs. It is not included in the threshold for this EAL because of the limited scope of the parameters it monitors.
This EAL is not required in modes 4 or 5 due to the limited number of safety systems required for operation.
DEVIATION None REFERENCES NUMARC NESP-0007, SS6 HC.OP.AB.ZZ-0143 (Q), Loss of Overhead Annunciators / Loss of CRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram O
U EAL - 8.2.3 Rev.00 Paae 2 of 2
HCGS EAURALTechnical Basis i
8.0 System Malfunctions .
l 8.3 Loss of Control Room Habitability ALERT - 8.3.2 IC Main Control Room Evacuation has been Initiated EAL
! Main Control Room Evacuation har been initiated OPERATIONAL CONDITION - All l BASIS Main Control Room evacuation represents a serious plant situation since the degree of plant control at the Remote Shutdown Panel is not as complete as from the Main Control Room.
The intent of this EAL is to declare an Alert when the determination to evacuate the Main m Control Room has been made based on environmental / personnel safety concerns, and the
- k physical process of evacuating the Control Room per OP-AB.ZZ-0130, Control Room l Evacuation has commenced.
l l Barrier Analysis
{
- N/A l
ESCALATION CRITERIA I Emergency Classification will escalate to a Site Area Emergency if control cannot be I
established within 15 minutes.
I DISCUSSION Control Room evacuation requires establishment of plant control from outside the control room
( Remote Shutdown Panels (RSP)). Support from the Technical Support Center (TSC) and/or other Emergency Operations Facility (EOF) is necessary.
Establishing remote system control will bypass many protective trips and interlocks. In addition, most of the instrumentation and assessment tools available in the Main Control Room
. will not be available. Operator actions upon deciding that the control room should be EAL - 8.3.2 Rev.00 Page 1 c,f 2
1
. HCGS EAURALTechnical Basis evacuated include scramming the reactor and closing the MSIVs. With these actions taken all inventory and pressure control cari be accomplished at the RSP.
A fire in any one of the following fire zones has the potential to render redundant safe shutdown controls and instrumentation in the Main Control Room inoperable.
Fire Zone Description i 5202 Cable Spreading Room l 5302 Control Equipment Room -
5403 Control Equipment Room Mezzanine 5510 Main Control Room 5605 Class IE Panel Room 5620 1E Panel Room HVAC l 5704 Diesel Area HVAC DEVIATION
! None l REFERENCES t
l NUMARC NESP-0007, HA5 HC.OP-AB.ZZ-0130 (Q), Control Room Evacuation l
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! l O EAL - 8.3.2 Rev.00 Page 2 of 2 l
, ., HCGS EAIJRALTechnical Basis 8.0 System Malfunctions -
L O 8.3 Loss of Control Room Habitability 1'
SITE AREA EMERGENCY - 8.3.3 IC Main Control Room Evacuation has been Initiated and Plant Control
- cannot be established EAL Main Control Room Evacuation has been initiated l- AND l
L Control of the plant CANNOT be established from outside the Main Control Room within 15 minutes OPERATIONAL CONDITION - All BASIS Failure to transfer and establish control of safety systems needed to maintain the Reactor in a 1 safe shutdown condition and remove decay heat, could result in damage to the fission product barriers, and the ability to determine plant status may be lost. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the control room, arrive at the Remote Shutdown Panel, and reestablish plant control to preclude core uncovery and/or core damage. The term " control of the plant" will require SNSS assessment to determine whether sufficient control has been established to maintain adequate core -
cooling.
Barrier Analysis l
N/A ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on loss of fission
{ product barriers, abnormal radiological releases, or Emergency Director judgement.
lO EAL - 8.3.3 Rev. 00 Page 1 of 2
i l .
HCGS EAURALTechnical Basis l l l DISCUSSION -
lO Most of the monitoring capability of the Remote Shutdown Panel is not enabled until control is l
transferred. During this transitional period the function of monitoring and/or controlling {
! parameters necessary for plant safety may not be occurring and may result in a threat to plant I
safety. If the transitional period is prolonged, damage to plant systems and safety barriers may occur and worsen without actions being taken to mitigate the consequences.
Control Room evacuation requires establishment of plant control from outside the control room l i
( Remote Shutdown Panels (RSP)). Support from the Technical Support Center (TSC) and/or l other Emergency Operations Facility (EOF) is necessary.
l A fire in any one of the following fire zones has the potential to re-rider redundant safe
- shutdown controls and instrumentation in the Main Control Room inoperable.
Fire Zone Description l 5202 Cable Spreading Room l 5302 Control Equipment Room 5403 Control Equipment Room Mezzanine l
l 5510 Main Control Room 5605 Class IE Panel Room
, 5620 IE Panel Room HVAC l 5704 Diesel Area HVAC l
DEVIATION None REFERENCES NUMARC NESP-0007, HS2 HC.OP-AB.ZZ-0130 (Q), Control Room Evacuation I
i
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!O EAL - 8.3.3 I Rev. 00 Page 2 of 2
. HCGS EAL/RALTechnical Easis 8.0 System Malfunctions -
8.4 Technical Specifications UNUSUAL EVENT - 3.4.1 IC ' Inability to Reach Required Operational Condition within Technical Specification Limits EAL' Plant is NQI brought to the REOUIRED Operational Condition within the Technical Specification required time limit OPERATIONAL CONDITION - 1,2, 3 BASIS Failure to place the unit in an Operational Condition in compliance with the Technical Specification LCO ACTION Statement warrants declaration of an Unusual Event due to the O plant being outside the defined Technical Specification safety envelope. Classification under this EAL is specific to an INABILITY OR FAILURE to comply with the mode change requirements of those Technical Specification LCOs that require the plant placed in a more conservative Operational Condition. Classification should be made under this EAL for a failure to comply with ANY Technical Specification required change in Operational Condition FROM the Operational Conditions in which this EAL applies (Operational Conditions 1,2 and 3). An Unusual Event is declared when the plant FAILS TO COMPLY WITH THE OPERATIONAL CONDITION change stated in the ACTION Statement of an LCO, and NOT as the result of a required ACTION.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate based upon system malfunctions or other conditions
! covered in various other EAL sections.
4
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Rev.00 l
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. HCGS EAL/RALTechnical Basis ,
DISCUSSION -
O A shutdown required by the Technical Specifications requires a one hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when actions are completed within the allowable action statement time in the 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 STATEMENT time period from the POINT OF RECOGNITION and is not related to how long a plant condition may have existed.
_ i DEVIATION None REFERENCES NUMARC NESP-0007, SU2 HCGS Technical Specifications ;
10CFR50.72 NUMARC Questions and Answers, June 1993, " General Question #7" NUMARC Questions and Answers, June 1993, " General Question #8" O I i
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EAL - 8.4.1 Rev.00 Page 2 of 2
.- HCGS EAIJRALTechnical Basis 9.0 Hazards -Internal / External O 9.1 Security Threats UNUSUAL EVENT - 9.1.1 IC ' Confirmed Security Event Which Indicates a Potential Degradation in the Level of l'
Safety of the Plant -
EAL l
l-Confirmed security threat directed toward the station as evidenced by AHX
- 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 e 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 l
- Hostage / Extortion situation that threatens normal plant operations, resulting in SCP-8 l implementation
- Destructive device discovered within the Protected Area, resulting in SCP-10 ;
implementation j
-i OPERATIONAL CONDITION - All l BASIS !
L 1
Security events classified under this EAL represent a potential degradation in the level of ,
safety of the plant. The EAL threshold is satisfied if the event is identified as being directed -
toward the station The intent of this EAL is to classify security events which threaten the Protected Area, but have not been determined to threaten Plant Vital Areas. A confirmed security threat exists if physical evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The SNSS/EC should declare an Unusual Event upon consulting with Security to determine the validity of the entry conditions. 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 l-
~ concerning the specific event to aid in classification.
EAL - 9.1.1 Rev.00 Page 1 of 2 i
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~- - . . . _ . ~ , .'
. HCGS EAIJRALTechnical Basis Barrier Analysis -
N/A ESCALATION CRITERIA Emergency Classification will escalate to an Alert based upon an actual hostile intrusion or malicious acts within the Protected Area.
DISCUSSION -
Security events which do not represent a potential degradation in the level of safety of the plant are reported under RAL Section 11.0, One Hour Non-Emergency Safeguards Event (10 CFR 73.71 or 10 CFR 50.72), and will not result in an Unusual Event declaration.
The following is an index of Security Contingency Procedures referenced by this event:
- SCP-5 " Security Threat"
- SCP-7 " Internal Disturbance"
- SCP-8 " Hostage Situation"
- SCP-10 " Discovery of Destructive Devices or Evidence of Malicious Acts"
- SCP-11 " Civil Disturbance" DEVIATION None REFERENCES NUMARC NESP-007, HU4.1, HU4.2 Safeguards Contingency Plan O
EAL - 9.1.1 Rev.00 l
Page 2 of 2
. HCGS EAURALTechnical Basis 9.0 Hazards -Internal / External -
G l 9.1 Security Threats j 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 i e Hostage held on-site in a non-vital area, resulting in SCP-8 implementation OPERATIONAL CONDITION - All BASIS O
'd Security events classified under this EAL represent an escalated threat to the level of safety of the plant. The event is confirmed if physical evidence supporting the hostile intrusion or assault exists. The intent of this EAL is to classify security events which represent an actual
- intrusion into the Protected Area. The SNSS/EC should declare an Alert upon consulting ;
l with the Security to determine the validity of the entry conditions. Security Contingency l 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 N/A i
ESCALATION CRITERIA -
Emergency Classification will be escalate to a Site Area Emergency based upon a hostile intrusion or act in Plant Vital Areas.
O EAL - 9.1.2 f Rev.00 Page 1 of 2 l
l
HCGS EAL/RALTechnical Basis DISCUSSION -
O The following is an index of Security Contingency Procedures referenced by this event:
- SCP-6 " Discovery ofIntruders or Attack"
- SCP-8 " Hostage Situation" DEVIATION None -
REFERENCES NUMARC NESP-007, HA4.1, HA4.2 Safeguards Contingency Plan l
O I
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i iO EAL - 9.1.2 Rev.00 Page 2 of 2
. HCGS EAURALTechnical Basis 9.0 IIazards -Internal / External I
O 9.1 Security Threats SITE AREA EMERGENCY - 9.1.3 IC Security Event in a Plant Vital Area EAL _
Confirmed hostile intrusion or malicious acts in Plant Vital Areas as evidenced by :
[
- Discovery of an intruder (s), armed and violent, within a Plant Vital Area, resulting in l SCP-6 implementation i e Malicious acts or destructive device discovered in a Plant Vital Area resulting in SCP-10 implementation OPERATIONAL CONDITION - All l BASIS O Security events classified under this EAL represent an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or assault has progressed from the Protected Area to a Plant Vital Area. These areas contain vital equipment which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such failure, destruction or release are also considered vital. Security Contingency Procedure (SCP)
- numbers are referenced following each EAL threshold. Since some SCP numbers appear in i more than one EAL, the on-duty PSE&G Security Supervisor will provide information concerning the specific event to aid in classification.
Barrier Analysis N/A l ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based upon the actual loss of
, physical control of the Main Control Room or Remote Shutdown Panel.
i l
O EAL - 9.1.3 Rev.00 Page 1 of 2
--.a.Lo-4_-- - -4 m -e 4 6e 4 4 4m- % . _4--e.. s -- -_4. -- -.k& 4--+f-h44-3--.----d- _M.,*m,-4-._-. J-4-> -% - , .,
. HCGS EAURALTechnical Basis DISCUSSION -
(
Plant Vital Areas are within the Protected Area and are generally controlled by card key .
readers. A hostile intrusion into a Plant Vital Area could represent a situation that threatens the safety of plant personnel and the general public.
The following is an index of the Security Contingency Procedure referenced by this event:
~
SCP-6 Discovery ofIntruders or Attack" SCP-10 " Discovery of Destructive Device or Evidence of Malicious Acts" -
DEVIATION ,
None REFERENCES NUMARC NESP-007, HSI.1, HSI.2 Safeguards Contingency Plan i
1 O !
1 1
O EAL - 9.1.3 l Rev.00 Page 2 of 2
i .
HCGS EAL/RALTechnical Basis 9.0 Hazards -Internal / External -
l 9.1 Security Threats GENERAL EMERGENCY - 9.1.4 ,
l IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown l
EAL Security event resulting in the actual loss of physical control of EITHER one of the following:
- Main Control Room
- Remote Shutdown Panel l
l l OPERATIONAL CONDITION - All BASIS p
C/ Security events classified under this EAL represent conditions under which a hostile force has taken physical control of areas required to reach and maintain cold shutdown. Both the Main Control Room and Remote Shutdown Panel are included, since control of either could hamper l l the operating crew's ability to perform a safe plant shutdown. Actual loss of physical control l is defined as the condition where licensed Control Room operators can no longer take required action to operate the plant, including unauthorized transfer of plant control from the Main Control Room.
Barrier Analysis N/A l
ESCALATION CRITERIA N/A DISCUSSION Security threats which meet the threshold for declaration of a General Emergency are an actual loss of physical control of the Main Control Room or the Remote Shutdown Panel. This situation places the plant in a potentially unstable condition with high potential of multiple
' e)
( barrier failures.
, EAL - 9.1.4 Rev. 00 Page 1 of 2
_ . . _ _ _ . _ . . ~ . _ - --
e
. HCGS EAL/RALTechnical Basis f
DEVIATION -
None ,
REFERENCES 4
NUMARC NESP-007, HGl.1, HGl.2 Safeguards Contingency Plan
)
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i EAL - 9.1.4 Rev.00 Page 2 of 2
. HCGS EAURALTechnical Basis 9.0 IIazards -Internal / External O 9.2 Fire l
UNUSUAL EVENT - 9.2.1 l IC Fire within the Protected Area Boundary Not Extinguished within 15 minutes of Detection
~
EAL Valid Fire Alarm is received in the Main Control Room QR Report of a fire from personnel at the scene AST
- Fire is within ANY one of the following Plant Structures (EXCLUDING small fires that
! have HQ potential to affect Safety Systems or Protected Area Permanent Plant l Structures)
!
- Reactor Building q e Turbine Building f )
- Control / Aux Building
- Service Water Intake Structure i
e Service / Rad Waste Building
. Low Level Radwaste Interim Storage Facility AND l Fire is NOI extinguished within 15 minutes of EITHER one of the following:
- Receipt of a Valid Fire Alarm l
- Report of a fire from the scene OPERATIONAL CONDITION - All BASIS Fires classified under this EAL include those of a magnitude and extent that may be a potential precursor to damage to Safety Systems, and hence have safety significance. This EAL includes Plant Vital Structures and also structures and areas that are contiguous to plant vital 1
structures, 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 Main Control Room is (o) considered to be Valid when the alarm is substantiated by the receipt of related independent EAL - 9.2.1 Rev.00 Page 1 of 3 I
l
i l
. HCGS EAL/RALTechnical Basis
, alarms (fire, temperature, deluge, etc.) in the Main Control Room or by visual confirmation if only a single detector is alarming. This EAL EXCLUDES such items as fires in Structures other then those listed in the EAL, waste-basket fires, and other small fires of no safety significance based on the judgement of the SNSS that N_Q potential to affect a Safety System i exists. Emergency Coordinatorjudgement 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 i Fire Alarm or report of a fire from personnel at the scene.15 minutes was determined to be a I reasonable time limit for small fires to be extinguished. A Safety System is defined as any system or compor ent included within the Technical Specification, i Fire is defined as combusuun 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 j smoke and heat are observed. l Barrier Analysis l !
l N/A l l
ESCALATION CRITERIA i l
l Emergency Classification will escalate to an Alert if the fire damages more than one plant i safety system or damages any plant vital structures.
l 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 non-vital structures include:
Circulating Water Structure l Station Service Transformer and Switchyard Area Hope Creek Admin. Building Onsite Warehouses Onsite Trailers Main and Aux Guardhouse Nuclear Services Building Auxiliary Boiler House DEVIATION None i EAL - 9.2.1
! Rev. 00 Page 2 of 3
- HCGS EAURALTechnical Basis REFERENCES .
O NUMARC NESP-0007, HU2 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-0001(Z) ,
NUMARC Questions and Answers, June 1993, " Hazards Question #7" ;
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EAL - 9.2.1
[
Rev.00 Page 3 of 3
. HCGS EAL/RALTechnical Basis l
9.0 IIazards -Internal / External -
i O 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 M one of the following Plant Vital Structures:
- Reactor Building
!
- Control / Aux Building e Service Water Intake Structure i
e Service / Rad Waste Building M
The Fire is of a magnitude that it SPECIFICALLY results in Damage to M one of the l
following:
- TWO OR MORE subsystems of a Safety System e MORE THAN ONE Safety System
. Any Plant Vital Structure which renders the structure incapable of performing its ;
Design Function i l M l Damaged Safety System (s) or Plant Vital Structure is required for the present Operational .
Condition OPERATIONAL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the fire and the effects on safety systems required for the present Operating Condition. Specific system degradation is addressed in the
! System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term " Damage" is defined as evidence that the fire has caused component g malfunction (pump trip, breaker trip, etc.) or a report of visible scorching, blistering or other Q
deformation that may have resulted in the equipment / structure being INOPERABLE or
( EAL - 9.2.2 i
Rev.00 Page 1 of 3
i
. HCGS EAL/RALTechnical Basis 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 Operational Condition.
For example, a fire that has been confirmed to be localized to a single piece of equipment, like a 4.16 KV Breaker, with no potential to spread to adjacent equipment, does not warrant classification as an Alert. In thc > vent, however, that the fire has spread or is believed to be spreading to other 4.16 KV Breakers for component (s) required for the present operating condition, then an Alert is warranted.
Fire is defined as combustion characterized by the generation heat and smoke. Sources of l 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.
i l
Barrier Analysis l 1
-N/A i
ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, loss of i fission product barriers, or abnormal radiological releases. The EC may use Emergency l
Coordinator Discretion and escalate the classification to SAE based on the nature of the fire.
DISCUSSION No lengthy and timely 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 l fire. In short, if the fire is big enough that it has damaged more than one safety system, or more than one subsystem of a safety system, then the fire is big enough tojustify an Alert declaration. Damage to Plant Vital Structures must be to the extent that EC judgement 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.
- DEVIATION I
i None EAL - 9.2.2
. Rev.00 Page 2 of 3 I
HCGS EAL/RALTechnical Basis REFERENCF3 -
{ i l NUMARC NESP-0007, HA2 l
HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-0001(Z) 1 I
HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System NUMARC Questions and Answers, June 1993, " Hazards Question #7"
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EAL - 9.2.2
' ' Rev.00 Page 3 of 3
. HCGS EAL/RALTechnical Basis l
!p 9.0 IIazards -Internal / External -
V 9.3 Explosion UNUSUAL EVENT - 9.3.1 l IC Natural and Destructive Phenomena Affecting the Protected Area r
EAL Confirmed Explosion within the Protected Area AST l Report of visible damage to Plant equipment or Protected Area Permanent Plant Structures l
l OPERATIONAL CONDITION - All BASIS Occurrence of these event within the Protected Area, that cause visible damage to plant equipment or Protected Area Permanent Plant Structures warrant declaration as an Unusual l 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 l sufficient for declaration. A confirmed explosion is defined as visual evidence that a rapid, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to damage or potentially damage permanent plant structures, systems or components.
Barrier Analysis N/A l
i ESCALATION CRITERIA l Emergency Classification will escalate to Alert if the explosion damages more than one safety systems or damages any plant vital structure.
f EAL - 9.3.1 Rev. 00 Page 1 of 2
(
. HCGS EAL/RALTechnical Basis DISCUSSION -
O 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-0007, HUI.5 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-0001(Z) l O 1
i 4 l h.
(V 1
l EAL - 9.3.1 1 Rev. 00 Page 2 of 2 l
y-, ,, , - - - . -
. HCGS EAIJRALTechnical Basis -
9.0 Hazards -Internal / External !
l l
O 9.3 Explosion l
ALERT - 9.3.2 IC Explosion Affecting the Operability of Plant Safety Systems Required to Establish or i Maintain Safe Shutdown i
- t
~
EAL l 5
l l i i
Confirmed Explosion within M one of the following Plant Vital Structures: }
l e Reactor Building .
l e Control / Aux Building l
- Service Water Intake Structure j e Service / Rad Waste Building ;
M i The Explosion is of a magnitude that it SPECIFICALLY results in Damage to M one of :
the following: i i e TWO OR MORE subsystems of a Safety System e MORE THAN ONE Safety System j e Any Plant Vital Structure which renders the structure incapable of performing its - l Design Function !
L u
M :
! i l Damaged Safety System (s) or Plant Vital Structure is required for the present Operational Condition :
L. ,
1 OPERATIONAL CONDITION - All !
BASIS .
The primary concern in this EAL is the magnitude of the explosion and the effects on safety systems recaired for the present Operational Condition. Specific system degradation is -
addressed '.n 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 roused' 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 j ~ d function. A Safety System is defined as any system required to maintain safe operation or to EAL - 9.3.2 Rev.00 Page 1 of 2
. HCGS EAL/RALTechnical Basis l
l i establish or maintain cold shutdown . In those cases where it is believed that the explosion I may have caused damage to Safety Systems, then an Alert declaration is warranted, since the I
full extent of the damage may not be known. For Plant Vital Structure damage, classification l_ is required under this EAL if the structure houses or otherwise supports safety systems l required for the present Operating Condition.
! A confirmed explosion is defined as visual evidence that a rapid, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to damage or potentially damage permanent plant structures, systems or components.
Barrier Analysis
- \
N/A !
l ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use Emergency l Coordinator Discretion and escalate the classification to SAE based on the nature of the ;
explosion. ;
l DISCUSSION l A b No lengthy and timely assessment of damage is required prior to classification. In this EAL, ,
l no attempt is made to quantify the magnitude of the damage to any safety system but instead an attempt is made to identify any damage in order to quantify the magnitude and extent of the explosion. In short, if the explosion is big enough that it has damaged more than one safety ;
- system, or more than one subsystem of a safety system, then the explosion is big enough to l justify an Alert declaration. Damage to Plant Vital Structures must be to the extent that EC l judgement 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.
l DEVIATION L
i None REFERENCES NUMARC NESP-0007, HA2 )
HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-000l(Z) j EAL - 9.3.2
- Rev.00 l Page 2 of 2
HCGS EAL/RALTechnical Basis i
9.0 IIazards -Internal / External .
9.4 Toxic Gases l UNUSUAL EVENT - 9.4.1.a h
IC Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the ,
1 Plant EAL l
l Notification by Local, County, or State Officials for the potential need to EVACUATE j l- non-essential personnel due to an Offsite Toxic Gas release l AND
- 1 l SNSS deems evacuation of non-essential personnel is required l
4 OPERATIONAL CONDITION - All
/~~N V BASIS Notification by Local, County, or State Officials for the potential need to EVACUATE non-l 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 subs'ance 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 N/A I
! ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the Toxic Gas enters either a Plant Vital f 1 Area or an area contiguous to a Plant Vital Area.
EAL - 9.4.1.a l
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. HCGS EAIJRALTechnical Basis DISCUSSION -
O None i
DEVIATION None REFERENCES l NUMARC NESP-0007, HU3.1 and HU3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air !
Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System O l i
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. HCGS EAL/RALTechnical Basis l 9.0 Hazards -Internal / External -
l' ii l 9.4 Toxic Gases l ALERT - 9.4.2.a i
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 _
P EAL Uncontrolled Toxic Gas release within ANY one of the following Plant Structures
- Reactor Building l
- Turbine Building I
- Control / Aux Building
- Service Water Intake Structure
- Service / Rad Waste Building l
l AND O Toxic Gas concentrations result in ANY one of the following: i e An IDLH atmosphere i e Plant personnel report severe adverse health reaction;, including burning eyes, l l nose, throat, or dizziness l
- The Lower Toxicity Limit being EXCEEDED l
i AND
! Plant personnel are unable to perform actions necessary to complete a Safe Shutdown L of the plant without appropriate personnel protection equipment l
OPERATIONAL CONDITION - All BASIS i
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 p 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 EAL - 9.4.2.a Rev.00 Page 1 of 3
I l . HCGS EAL/RALTechnical Basis possible with appropriate personnel protection equipment, since this equipment restricts normal
[_ vision and mobility.
Barrier Analysis l
N/A L
ESCALATION CRITERIA
- Emergency Classification will escalate to an Alert if the Flammable Gas enter either a Plant l Vital Area or an area contiguous to a Plant Vital Area.
l DISCUSSION l
I This EAL should not be construed to include confined spaces that must be ventilated prior to l entry or situations involving Site Protection personnel who are using respiratory equipment j during the performance of their duties unless it also affects personnel not involved with Site j Protection activates. These areas include the Drywell (when inerted) and ALL Confined Spaces. In addition, those situations that require personnel to wear respiratory protection equipment as the result of airborne contamination as required by Radiation Protection l personnel do not meet the intent of this EAL.
i An offsite event (such as a tanker accident or a barge accident) may place the Protected Area l
C within the evacuation area. The evacuation is determined from the DOT Evacuation Tables i
for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous ,
l Materials. !
DEVIATION l None l REFERENCES l
NUMARC NESP-0007, HU3.1 and HU3.2
- HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply l HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System ]
l O -
[
EAL - 9.4.1.b Rev.00 Page 2 of 2
_.l l
. HCGS EAURALTechnical Basis 1 i
g 9.0 Hazards -Internal / External -
V 9.4 Toxic Gases f
UNUSUAL EVENT - 9.4.1.c l l
IC Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the :
Plant i EAL i
Uncontrolled Flanunable Gas release within the Protected Area that RESULTS in Flammable Gas concentrations EXCEEDING 25% of the LEL !
l AND 1
Routine Plant Operations are IMPEDED based on EITHER one of the following: :
I
- Access restrictions caused by the uncontrolled release j i
- Personnel injuries have occurred as a result of the release v 1 l
OPERATIONAL CONDITION - All l BASIS l An uncontrolled Flammable Gas release within the Protected Area, in high enough l concentrations, will adversely affect the health and safety of plant personnel, along with the L safe operation of the plant. This EAL specifically addresses those conditions where a l l Flammable Gas concentration EXCEEDING 25% of the LEL exists anywhere within the l 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 / confined to a single compartment or area. For ,
example, an uncontrolled release of hydrogen into the Turbine Building in concentration l exceeding 25% of the LEL (Lower Explosive Limit) warrants declaration of an Unusual l Event. In comparison, a controlled release of Hysogen during Generator purging or I
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
- mcludes those conditions where access is only possible with appropriate personnel protection I
d equipment, since this equipment restricts normal vision and mobility.
l EAL - 9.4.1.c Rev.00 Page 1 of 2
l
. HCGS EAL/RALTechnical Basis !
I
, . ' Barrier Analysis -
N/A j l
ESCALATION CRITERIA i I
1 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, i
l DISCUSSION -
l l For Hydrogen Gas, the explosive limit is 4 %. Hence, a threshold of 25 % of the LEL equates j to 1% Hydrogen. This EAL should not be construed to include those controlled evolutions that ;
l may discharge a Flammable Gas within the Protected Area, but present no danger to plant :
[ safety, since the evolution is planned and controlled.
i :
i 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 i for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous l l Materials. I l
DEVIATION ,
' O U None l l
REFERENCES i I NUMARC NESP-0007, HU3.1 and HU3.2 ,
HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air l i
Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System l
l
, O l V
f EAL - 9.4. l.c Rev. 00 Page 2 of 2 I-
1
, HCGS EAIJRALTechnical Basis
(
9.0 IIazards -Internal / External q
]
v 9.4 Toxic 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 Structures
- Reactor Building :
- Turbine Building l
- Control / Aux Building f
. Service Water Intake Structure )
e Service / Rad Waste Building l
,o Agg
(,) 4 Toxic Gas concentrations result in ANY one of the following:
- An IDLH atmosphere l e Plant personnel report severe adverse health reactions, including burning eyes, e nose, throat, or dizziness l
- The Lower Toxicity Limit being EXCEEDED AST Plant personnel are unable to perform actions necessary to complete a Safe Shutdown of the plant without appropriate personnel protection equipment OPERATIONAL CONDITION - All BASIS An uncontrolled Toxic Gas release entering any of the plant structures listed in the EAL, that threatens the ability of plant personnel to perform actions required for safe shutdown of the I
(
'u) plant, warrants declaration of an Alert. The EAL threshold includes those conditions that present a significant challe. ige to plant personnel. This EAL specifically addresses only those EAL - 9.4.2.a Rev.00 Page 1 of 3
, HCGS EAL/RALTechnical Basis 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 l
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 i releases that can not be isolated / confined to a single compartment or area, or are not as the j result of a designed plant safety feature. For example, an uncontrolled release of chlorine / ammonia into the Turbine Building that directly effects plant personnel, warrants declaration of an Alert. A Cardox discharge inside any area that contains this safety feature (i.e. Diesel Bays) does not warrant Alert declaration, unless personnel injuries have occurred l as a direct result of the discharge. -
Barrier Analysis l N/A j ESCALATION CRITERIA Emergency Classification will escalated based on further damage to plant safety systems, loss l of fission product barriers, or abnormal radiological releases. The EC may use Emergency j Coordinator Discretion and escalate the classification to SAE based on the nature of the toxic gas release. ,
i DISCUSSION Access is considered impeded if the Toxic Gas concentrations are life threatening, i.e. require l the use of personnel protective equipment. Use of protective equipment also limits the j mobility and vision. The cause or magnitude of the gas concentration is not the major concern ;
i _ in this EAL, but rather that access required to an area that may be impeded. An IDLH l atmosphere is any atmosphere that is determined to be Immediately Dangerous to Life and Health.
l This EAL should not be construed to include confined spaces that must be ventilated prior to l entry or situations involving Site Protection personnel who are using respiratory equipment i during the performance of their d0 ties unless it also affects personnel not involved with Site Protection activities. These areas include the Drywell (when inerted) and . ALL Confined Spaces. In addition, those situations that require personnel to wear respiratory protection equipment as the result of airborne contamination as required by Radiation Protection ]
personnel do not meet the intent of this EAL. l 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 i Materials.
'O EAL - 9.4.2.a Rev. 00 Page 2 of 3
1 HCGS EAURALTechnical Basis ,
I DEVIATION .
O None REFERENCES NUMARC NESP-0007, HA3.1 and HA3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply.
HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System l l i i
i
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O .
l I
1 1
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i EAL - 9.4.2.a Rev. 00 Page 3 of 3 I
. HCGS EAL/RALTechnical Basis .
9.0 IIazards -Internal / External -
9.4 Toxic 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 M one of the following Plant Structures
- Reactor Building l
- Turbine Building
- Control / Aux Building
- Service Water Intake Structure j e Service / Rad Waste Building S
(%J3 Flammable Gas concentrations EXCEED 50% of the LEL M i Plant personnel are unable to perform actions necessary to complete a Safe Shutdown of the plant without appropriate personnel protection equipment OPERATIONAL CONDITION - All l
BASIS An uncontrolled Flammable Gas release entering any of the plant structures listed in the EAL, that threatens the ability of plant personnel to perform actions required for safe shutdown of the plant, warrants declaration of an Alert. The EAL threshold includes those conditions that present a significant challenge to plant personnel. This EAL specifically addresses only those plant structures that either contain safe shutdown equipment or are contiguous to those areas. Release classified under this EAL include those that originate both onsite and offsite. A Flammable Gas is considered to be any substance that is capable of p being easily ignited or burning quickly. Uncontrolled Flammable Gas releases are considered
\_/ to be those releases that can not be isolated / confined to a single compartment or area, or are EAL - 9.4.2.b Rev.00 Page 1 of 2
F .
1 i
! , HCGS EAL/RALTechnical Basis i l
[ not as the result of a designed plant safety feature. For example, an uncontrolled release of hydrogen into the Turbine Building in concentration exceeding 50% of the LEL (Lower .
Explosive Limit) warrants declaration of an Alert. In comparison, a controlled release of l Hydrogen during Generator purging does not warrant event declaration, as this evolution is .
controlled. !
l Barrier Analysis N/A ESCALATION CRITERIA l
i Emergency Classification will escalated based on subsequent damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use l Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the flammable gas release.
l
! i 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 I may discharge a Flammable Gas within the Protected Area, but present no danger to plant i 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 1 for Selected Hazardous Materials in the DOT Emergency Respon.,e Guide for Hazardous Materials.
DEVIATION None REFERENCES NUMARC NESP-0007, HA3.1 and HA3.2 i HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air !
Supply I l
HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System t r I
EAL - 9.4.2.b Rev.00 Page 2 of 2
1
, HCGS EAURALTechnical Basis
! , 9.0 Ilazards -Internal / External -
b 9.5 Seismic Event l UNUSUAL EVENT - 9.5.1 a / 9.5.1.b l
IC Natural and Destructive Phenomena Affecting the Protected Area EAL
< i l
l -
EITHER one of the following conditions:
- Seismic Event felt by personnel within the Protected Area e Valid Actuation of the Seismic Trigger (> 0.01g) has occurred as verified by the l SMA-3 Event Indicator (flag) being WIIITE on Panel 10-C-673 in the Upper 1
! Relay Room l
OPERATIONAL CONDITION - All
(~
l (./ BASIS The condition that the Seismic Event has been felt by personnel within the Protected Area, or a Valid Actuation of the Seismic Trigger indicates that a Seismic Event of a magnitude greater than 0.0lg has occurred. This threshold warrants declaration of an Unusual Event. Valid is defined as the Seismic Trigger actuation being the direct result of a Seismic Event Classification should be based on a Valid actuation of the Seismic Trigger as verified in the Upper Relay Room. Additional information can be obtained by contacting the National Earthquake Center in Denver, Colorado at (303) 273-8500. However, it is important to realize that it will take the Earthquake Center approximately 30 minutes to provide the requested ;
l information. The time required to obtain this additional information should not result in a ;
delay of event classification for a valid actuation.
Barrier Analysis N/A l
l ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the a subsequent seismic event occurred in p excess of the Operating Basis Earthquake level (0.lg).
l V EAL - 9.5.1 Rev. 00 Page 1 of 2
- - - - - _ . - - . . .- - -. . _. - = . _ - . . _ - _ _ .
. HCGS EAL/RALTechnical Basis l
DISCUSSION
- O An earthquake of a magnitude equivalent to 0.0lg is not expected to affect the capability of plant safety functions. This threshold value is well below the Operating Basis Earthquake level of 0.lg.
An approximate relationship between acceleration and magnitude is as follows:
An Acceleration of: is approx. equal to a Richter Scale Magnitude of:
0.0lg 4.0 -
l 0.02g 4.5 0.lg 5.5 0.2g 6.5 i l DEVIATION l
l None REFERENCES ,
l NUMARC NESP-0007, HUl.1 HC.OP AB.ZZ-0139 (Q), Acts of Nature O HCGS Technical Specification Section 3/4.3.7.2, Seismic Monitoring Instrumentation HC.OP-SO.SG-0001 (Z), Seismic Instrumentation System Operation HC.OP-AR.ZZ-0011 (Q), Overhead Annunciator Window Box C6 l
i l
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- O I EAL - 9.5.1 Rev.00 Page 2 of 2
- , HCGS EAL/RALTechnical Basis l 9.0 IIazards -Internal /Ex1ernal -
1 9.5 Seismic Event l
l ALERT - 9.5.2 IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL l
l Seismic Event felt by personnel within the Protected Area l
E Valid Actuation of the Seismic Trigger (> 0.01g) has occurred as verified by the SMA-3 Event Indicator (flag) being Wi1ITE on Panel 10-C-673 in the Upper l Relay Room AST i Valid Actuation of the Seismic Switch (> 0.1g) has occurred as verified by EITHER~
lh us one of the following:
j e Valid Actuation of Main Control Room Overhead Annunciator C6-C4 l . AMBER Alarm light on the Seismic Switch Power Supply Drawer is lit on Panel 10-C-673 in the Upper Relay Room l
l OPERATIONAL CONDITION - All BASIS
! A Valid Actuation of the Seismic Switch indicates that a Seismic Event of a magnitude greater l than 0.lg (Operating Basis Earthquake) has cccurred. The Salem SNSS must be informed of this information immediately. At this level, plant safety systems are designed to remain functional and within design stress and deformation limits. Thus, an earthquake of this l 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.
L This threshold warrants declaration of an Alert. Valid is defined as the Seismic Switch actuation being the direct result of a Seismic Event. The condition that the Seismic Event has been felt by personnel within the Protected Area or along with Seismic Trigger actuation
'O provides further confirmation that an event has occurred. Classification should be based or' V ;
EAL - 9.5.2 l
Rev.00 "
l Page 1 of 2 l
1
]
L 1
HCGS EAIJRALTechnical Basis Valid actuation of the Seismic Switch as verified in the Upper Relay Room. Additional
~
(
A information can be obtained by contacting the National Earthquake Center in Denver, Colorado at (303) 273-8500. However, it is important to realize that it will take the Earthquake Center approximately 30 minutes to provide the requested information. The time required to obtain this additional information should not result in a delay of event classification for a valid actuation.
Barrier Analysis N/A -
ESCALATION CRITERIA Emergency Classification will escalate if the seismic event caused additional 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 event.
DISCUSSION l Seismic Event annunciation on panel 10C673 would alert operators to this event and the active l seismic monitoring instrumentation would begin to monitor the event. This threshold value p associated with this EAL is well below the Design Basis Earthquake of 0.2g that is the G1 maximum seismic event that is expected to occur based on local geological and seismological factors.
An approximate relationship between acceleration and magnitude is as follows:
Acceleration: Richter Scale Magnitude (approximate):
l 0.0lg 4.0 0.02g 4.5 0.lg 5.5 0.2g 6.5 l
l DEVIATION !
I j None !
REFERENCES NUMARC NESP-0007, HA1.1 l HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4.3.7.2, Seismic Monitoring Instrumentation HC.OP-SO.SG-0001 (Z), Seismic Instrumentation System Operation
(. HC.OP-AR.ZZ-00ll (Q), Overhead Annunciator Window Box C6 EAL - 9.5.2 Rev.00 ,
Page 2 of 2 l.
i l,
. HCGS EAURALTechnical Basis i q 9.0 IIazards -Internal / External -
O 9.6 Iligh Winds
- l UNUSUAL EVENT - 9.6.I.a IC Natural and Destructive Phenomena Affecting the Protected Area i
EAL -
i
! Report of a Tornado TOUCHING DOWN within the Protected Area i
l OPERATIONAL CONDITION - All l l
i BASIS l
A tomado touching down within the Protected Area is an observed event with the potential to cause damage to structures containing systems or functions necessary for safe shutdown of the plant. As such, a tornado represents a potential degradation in the level of safety of the plant.
!. O Verification of the tornado should be by direct observation and report by plant personnel.
lO l l Barrier Analysis l
l N/A I ESCALATION CRITERIA
- Emergency Classification will escalate to an Alert if the tornado causes damage to Plant Vital l l Structures or affects the operability of Technical Specification required equipment i i
i DISCUSSION l 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 O
V None l EAL - 9.6.1.a l Rev.00 Page 1 of 2
i
. HCGS EAURALTechnical Basis l REFERENCES -
)
NUMARC NESP-0007, HUI.2 and HUI.7 HC.OP-AB.ZZ-0139 (Q), Acts of Nature l HCGS Technical Specification Section 3/4, 3.7.3, Meteorological Monitoring Instrumentation HCGS UFSAR Sections 2.3,3.3.1 l
1 f
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- EAL - 9.6.1.a Rev.00 Page 2 of 2 l
. HCGS EAURALTechnical Basis 9.0 IIazards -Internal / External -
l O 9.6 High Winds UNUSUAL EVENT - 9.6.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Sustained wind speeds > 75 MPH for 15 minutes, from at ANY elevation of the Met Tower OPERATIONAL CONDITION - All l
BASIS Sustained wind speeds in excess of 75 MPH are of sufficient velocity to have the potential to cause damage to Plant Vital Areas. These conditions are indicative of unstable weather
, conditions and represent a potential degradation in the level of safety of the plant. The
) windspeed threshold is well below the structure design basis of 108 mph, and is set at the value used to characterize Hurricane force winds. Sustained wind speed means winds in excess of the threshold value for greater than 15 minutes.
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the high winds cause damage to Plant Vital Structures or affects the operability of Technical Specification required equipment. q DISCUSSION Verification of sustained wind speed will be by observation of meteorological tower data.
The Wind Speed indication from the Met Tower instrumentation is full scale at 100 mph.
l O EAL - 9.6.1.b l
{
Rev.00 I Page 1 of 2
, HCGS EAL/RALTechnical Basis
! The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions:
l Wilmington (302) 573-6142 Mount Holly (609) 261-6604 ,
Mount Holly (609) 261-6602 l l DEVIATION None -
l REFERENCES l
NUMARC NESP-0007, HU1.2 and HU1.7 HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4,3.7.3, Meteorological Monitoring Instrumentation HCGS UFSAR Sections 2.3,3.3.1 l
l ;
I O
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l I l !
i
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l lO I EAL - 9.6.1.b Rev. 00 Page 2 of 2
1
. HCGS EAL/RALTechnical Basi: i
,s 9.0 IIazards -Internal / External -
I 9.6 High Winds ;
1 ALERT - 9.6.2 j IC Natural and Destructive Phenomena Affecting the Plant Vital Area l
~
l EITHER one of the following:
l
- Report of a Tornado TOUCHING DOWN within the Protected Area l
- Sustained wind speeds > 75 MPH for 15 minutes, from at ANY elevation of the Met Tower AhP i
l l i l The Wind Speed is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
- TWO OR MORE subsystems of a Safety System ;
e MORE THAN ONE Safety System !
- Rendering ANY of the following structures incapable of performing its Design l Function: )
e Reactor Building e Control / Aux Building
!
- Service Water Intake Structure
- Service /Radwaste Building l AND Damaged Safety System (s) or Plant Vital Structure is required for the present Operating Condition ,
OPERATIONAL COhTITION - All
- . BASIS l
l The primary concern in this EAL is the magnitude of the high winds and the effects on safety 4
functions required for the present Operating Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not c required prior to classification. The term " Damage" is defined as evidence that the high winds
- ( has caused component malfunction (pump trip, breaker trip, etc.) or a report of visible j EAL - 9.6.2 Rev.00 Page 1 of 3
l HCGS EAL/RALTechnical Basis l l
scorching, blistering or other 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 j damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the i damage may not be known. For Plant Vital Structure damage, classification is required under l this EAL if the structure houses or otherwise supports safety systems required for the present i Operating Condition. !
It is not intended that a lengthy engineering analysis be performed to determine if damage has affected structural design but EC judgement must determine whether to exclude minor exterior i damage which does not affect the structural design capability.. Sustained wind speed means )
winds in excess of the threshold value for greater than 15 minutes. t l
Barrier Analysis N/A ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use Emergency 1 Coordinator Discretion and escalate the classification to SAE based on the nature of the winds.
DISCUSSION The windspeed threshold is well below the structure design basis of 108 mph, and is set at the value used to characterize Hurricane force winds. The Wind Speed indication from the Met Tower instrumentation is full scale at 100 mph.
The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions:
Wilmington (302) 573-6142 Mount Holly (609) 261-6604 Mount Holly (609) 261-6602 DEVIATION None O EAL - 9.6.2 Rev.00 Page 2 of 3
HCGS EAURALTechnical Basis REFERENCES -
NUMARC NESP-0007, HA1.2 arid HA1.3 ;
HC.OP-AB.ZZ-0139 (Q), Acts of Nature ;
HCGS Technical Specification Section 3/4, 3.7.3, Meteorological Monitoring Instrumentation j HCGS UFSAR Sections 2.3,3.3.1
{
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l EAL - 9.6.2 Rev.00
' Page 3 o' 3 i
HCGS EAURALTechnical Basis 9.0 IIazards -Internal / External -
O 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 l
Visual Observation of Uncontrolled Flooding that confirms M one of the following:
- Reactor Building Floor Levels above the Maximum Normal Floor Level (> 1")
referenced in EOP 103, Secondary Containment Control e Receipt of a SSWS Pump Room Flooded Alarm
. Greater than 2" of water in M area that contains a Safety System (s), not included above
, OPERATIONAL CONDITION - All BASIS Uncontrolled flooding in the areas listed in the EAL represents the potential to directly impact i continued safe operation of the plant. This EAL specifically addresses those areas of the plant where uncontrolled flooding presents a challenge Safety System (s). Visual Observation of the flooding should occur prior to classificatior. to validate any alarm conditions. Uncontrolled flooding is defined as event or condition that does not result from a controlled evolution.
Events classified under this EAL, for example, include the effects of flooding from system malfunctions, component failures, or repair activity failures (such as a failed freeze seal).
Those events that result in the flooding of an area as the direct result of a planned evolution, such as system draining in preparation for an equipment outage, do not warrant event l classification, unless the draining can not be successfully terminated. Safety System is defined as any system or component included in the Tech lical Specification.
l l Barrier Analysis I N/A EAL - 9.7.1 Rev.00 Page 1 of 2
- , HCGS EAL/RALTechnical Basis l
l ESCALATION CRITERIA '
v Emergency Classification will escalate to an Alert if the flooding results in damage to equipment required for the present Operational Condition.
I DISCUSSION For the purpose of implementing this EAL, levels in the Reactor Building that would require classification under this EAL are defined as the Maximum Normal Floor Level in the EOPs.
Exceeding this level in any of the Reactor Building areas would require running all available sump pumps. Iflevel in these areas cannot be lowered to below the 1" level, then systems
! discharging into this area are to be isolated, except for systems required to:
- Ensure adequate core cooling
- Shutdown the reactor
- Protect primary containment integrity
- Suppress a fire DEVIATION None l
REFERENCES l i s s NUMARC NESP-0007, HU1.7 l HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HCGS Technical Specifications Section 3/4 7-3, Flood Protection l
l i
i o EAL - 9.7.1 Rev.00 Page 2 of 2 l
l l
l
, HCGS EAL/RALTechnical Basis l
9.0 IIazards -Internal / External -
,.)
(
9.7 Flooding ALERT - 9.7.2 IC Internal Flooding Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown
~~
EAL '
Visual Observation of Flooding within AliY one of the following Plant Vital Structures:
- Reactor Building
- Control / Aux Building
- Service Water Intake Structure e Service / Rad Waste Building AND
! The Flooding is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
l j (q ;
- TWO OR MORE subsystems of a Safety System
- e MORE THAN ONE Safety System l
- Any of the above listed Plant Vital Structures which renders the structure incapable of performing its Design Function AND l Damaged Safety System (s) or Plant Vital Structure is required for the present Operational Condition l
OPER ATIONAL CONDITION - All l
BASIS The primary concern in this EAL is the magnitude of the internal flooding and the effects on safety systems required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not 3
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 p 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
()
EAL - 9.7.2 Rev 00 Page 1 of 2
HCGS EAL/RALTechnical Basis
! operation or to establish or maintain cold shutaown. In those cases where it is believed that
!OO the internal flooding may have cau' sed damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the structure houses or otherwise supports safety systems required for the present Operational Condition.
l Bstrier Analysis l
N/A ESCALATION CRITERIA Emergency Classification will escalate based on damage to plant systems, loss of fission l product barriers, or abnormal radiological releases. The EC may use Emergency coordinator l Discretion and escalate the classification to SAE based on the nature of the flooding.
! DISCUSSION i
Degraded system performance or observation of potential for damage that could degrade l system performance is used as the indicator that the safety system operability was actually l affected. A report of damage should not be interpreted as mandating a lengthy and timely assessment prior to justification; there is no inference in this EAL that the actual magnitude of
,/N damage be qualified or quantified.
l DEVIATION None l REFERENCES l
NUMARC NESP-0007, HA1.7 l HCGS Technical Specifications l
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EAL - 9.7.2 Rev.00 Page 2 of 2
HCGS EAL/RALTechnical Basis 9.0 IIazards -Internal / External .
l O 9.8 Turbine Failure / Vehicle - Missile Impact l
l UNUSUAL EVENT - 9.8.1.a IC Natural and Destructive Phenomena Affecting Certain Structures Within the Protected l
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 l
l OPERATIONAL CONDITION - 1,2,3 '
,s BASIS L,
Main Turbine failure of sufficient magnitude to cause damage to the turbine casing or generator seals increases the potential for leakage of combustible / explosive gases and of combustible liquids to the Turbine Building, warrants declaration of an Unusual Event. The i presence of H2gas in sufficient quantities may present a flammable / explosive hazard. Oil may also be present which may contribute to the flammability hazard.
l !
l Barrier Analysis i 1
l
! N/A 4 i i' i
ESCALATION CRITERIA l
l Emergency Classification will escalate to an Alert based upon damage done by missiles
!. generated by the failure or by any subsequent fire.
l DISCUSSION i
~
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 4 flammable liquids to the Turbine Building. Potential rupture of the condenser and condenser O, j V
EAL - 9.8.1.a Rev. 00 Page 1 of 2
HCGS EAURALTechnical Basis tubes may cause flooding in the lower levels of the Turbine Building. This damage should be readily observable.
l 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 l
, Modes 1,2,3 are the only Operational Conditions where Main Steam pressure is high enough to allow for Main Turbine operation. _
i REFERENCES l 4
NUMARC NESP-007, HUl.6 l l
I l
l 0
{
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- O EAL - 9.8.1.a l
l_ Rev. 00 Page 2 of 2 1 - - - .
l l ,
HCGS EAL/RALTechnical Basis 9.0 IIazards -Internal / External -
l O l\)
9.8 Turbine Failure / Vehicle - Missile Impa'.i UNUSUAL EVENT - 9.8.I.b IC Natural and Destructive Phenomena Affecting Certain Structures Within the Protected !
Area l i ~
l EAL l
Vehicle Crash / Missile Impact with or within ANY one of the following Plant
! Structures:
l l
- Reactor Building
- Turbine Building
- Control / Aux Building
- Service Water Intake Structure i e Service /Radwaste Building OPERATIONAL CONDITION - All l BASIS l l A Vehicle Crash / Missile Impact with or within a listed Plant Structure represents a potential ,
l challenge to plant safety. Events classified under this EAL include those of a magnitude and !
l 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 l l includes flying objects from both offsite and onsite, rotating equipment or turbine failure i causmg turbine casing penetration.
l Barrier Analysis l
None ESCALATION CRITERIA 1
Emergency Classification will escalate to Alert if the vehicle crash or missile impact causes damage to Plant Vital Structures.
G EAL - 9.8.1.b Rev.00 Page 1 of 2
~. _ ._. ___.___.._ _.___.. _ _ _ -._ __
HCGS EAIJRALTechnical Basis '
DISCUSSION .
Any security aspects of this event should be considered under ECG Section 9.1, Security ,
Events, i
DEVIATION I i
None l
. REFERENCES NUMARC NESP-0007, HUl.4 NUMARC Questions and Answers, June 1993, " Hazards Question #6" h
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!O EAL - 9.8.1.b Rev. 00 Page 2 of 2
i ,
HCGS EAL/RALTechnical Basis
! 9.0 IIazards -Internal / External- -
O 9.8 Turbine Failure / Vehicle - Missile Impact
- ALERT - 9.8.2 l IC Natural and Destructive Phenomena Affecting Certain Structures Within the Plant Vital l Area 1 -
EAL Vehicle Crash / Missile Impact with or within ANY one of the following Plant Vital Structures:
- Reactor Building
- Control / Aux Building
- Service Water Intake Structure l
- Service / Rad Waste Building l
AND The Vehicle Crash / Missile Impact is of a magnitude that it SPECIFICALLY results
- in Damage to AfD' one of the following
e TWO OR MORE subsystems of a Safety System e MORE THAN ONE Safety System ,
l e Any of the above Plant Vital Structures which renders the structure incapable of l performing its Design Function I
AND Damaged Safety System (s) or Plant Vital Structure is required for the present Operational Condition OPERATIONAL CONDITION - All l BASIS The primary concern in this EAL is the magnitude of the vehicle crashes / missile impact and the effects on safety systems required for the present Operational Condition. Specific system i degradation is addressed 5 the System Malfunction EALs. A detailed assessment of system damage is not reqmred prior to classification. The term " Damage" is defined as evidence that the vehicle crashes / missile impact has caused component malfunction (pump trip, breaker d EAL - 9.8.2 Rev.00 Page 1 of 2
HCGS EAL/RALTechnical Basis trip, etc.) that may have resulted in the equipment / structure being INOPERABLE or otherwise incapable of performing it's design function.
A Safety System is defined as any system required to maintain safe operation or to establish or maintain cold shutdown. In those cases where it is believed that the vehicle crashes / 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, l classification is required under this EAL if the structure houses or otherwise supports safety systems required for the present Operational Condition.
Barrier Analysis i i
N/A l l
ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, fission product barriers, or abnormal radiation releases in other EAL sections. The EC may use Emergency Coordinator Discretion and escalate the classification based on the nature of the i damage.
DISCUSSION This EAL is intended to address the threat to safety related equipment imposed by vehicle of missile impacts. No attempt should be made to assess the magnitude of damage to Safety Systems or Plant Vital Structures prior to classification. The evidence af damage is sufficient for declaration.
DEVIATION None l REFERENCES NUMARC NESP-0007, HA1.5 and HA1.6 NUMARC Questions and Answers, June 1993, " Hazards Question #6" l
lO
! EAL - 9.8.2 l Rev.00 Page 2 of 2 i