ML18128A088
| ML18128A088 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs, Nine Mile Point, FitzPatrick, 07201036  |
| Issue date: | 04/30/2018 |
| From: | Exelon Generation Co |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18128A077 | List: |
| References | |
| [[::JAF-18-0040|JAF-18-0040]], NMP1L3209 EP-AA-1014, Addendum 3, Rev 0 | |
| Download: ML18128A088 (310) | |
Text
{{#Wiki_filter:ATTACHMENT 7 Radiological Emergency Plan Addendum Revision EP-AA-1014, Addendum 3, Revision 0, "James A. FitzPatrick Nuclear Power Plant Emergency Action Levels"
FitzPatrick Annex Exelon Nuclear
- ~* Exelon Generation.,,
EP-AA-1014, Addendum 3 Revision 0 JAMES A. FITZPATRICK NUCLEAR POWER PLANT EMERGENCY ACTION LEVELS
- Apri 2018 Page 1 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- SECTION 1.0 TABLE OF CONTENTS PAGE PURPOSE ..................................................................................................4 2.0 PROCEDURE ............................................................................................4
3.0 REFERENCES
...........................................................................................4 4.0 DEFINITIONS .............................................................................................5 5.0 ATTACHMENTS .........................................................................................9 ATTACHMENT 1 - EMERGENCY CLASSIFICATION AND DECLARATION PROCESS .................................................................. 10 ATTACHMENT 2 - EAL CLASSIFICATION MATRIX (POSTED ATTACHMENT) ................................................................... 11 ATTACHMENT 3 - EAL BASES ......................................................................... 12 CATEGORY A ABNORMAL RAD RELEASE/RAD EFFLUENT ........................ 12 CATEGORY C COLD SHUTDOWN/REFUE~ING SYSTEM MALFUNCTION .. 52
- CATEGORY E ISFSl .........................................................................................113 CATEGORY H HAZARDS ................................................................................ 116 CATEGORY S SYSTEM MALFUNCTION ....................................................... 178 CATEGORY F FISSION PRODUCT BARRIER DEGRADATION .................... 236 ATTACHMENT 4 - FISSION PRODUCT BARRIER LOSS I POTENTIAL LOSS MATRIX AND BASES ......................................... 244 ATTACHMENT 5 -ADDITIONAL GUIDANCE FOR CLASSIFICATION .......... 308
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- 1.0 PURPOSE To assess emergency conditions and events, CLASSIFY the emergency classification in accoridnace with the approved Emergency.
Action Level scheme and DECLARE the emergency. 2.0 PROCEDURE 2.1 Classify and declare emergencies in accordance with Attachment 1, Emergency Classification and Declaration Process.
3.0 REFERENCES
3.1 Developmental 3.1.1 NEI 99-01 Revision 5, Methodology for Development of Emergency Action Levels, Final Draft, February 2008 (ADAMS Accession Number ML080450149) 3.1.2 EP-1 EOP Entry and Use 3.1.3 10CFR50Appendix E.IV.D 3.1.4 NSIR/DPR-ISG-01 Interim Staff Guidance On Emergency Planning for Nuclear Power Plants
- 3.2 Implementing 3.2.1 3.2.2 3.2.3 IAP-2 Classification of Emergency Conditions EAL Comparison Matrix EAL Classification Matrix 3.3 Commitments 3.3.1 NRC Safety Evaluation for Technical Specifications Amendment No. 278, ELIMINATION OF REQUIREMENTS FOR POST ACCIDENT SAMPLING. The Licensee has committed to establish and maintain the capability for classifying fuel damage events at the Alert level threshold (typically this is 300 µCi/ml dose equivalent iodine). This capability will be described in plant procedures .
- Apri 2018 Page 3 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- 4.0 DEFINITIONS Bomb Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.
Civil Disturbance A group of people violently protesting station operations or activities at the site. Classification Categorization of plant conditions or events into the appropriate emergency classification level. ConfinementBounda~ The barrier(s) between areas containing radioactive substances and the environment. Containment Closure Is the action taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions? As applied to JAFNPP, Containment Closure is established when either the Primary
- Containment or Secondary Containment is Operable per Sections 3.6.1.1 or 3.6.4.1 of Technical Specifications.
Declaration Announcement in the Control Room or EOF that an EAL has been met and an emergency classification level has been entered. Explosion Is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized/energized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components. Extortion Is an attempt to cause an action at the station by threat of force. Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed .
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I i FitzPatrick Annex Exelon Nuclear
- Flooding The presence of water outside the confines of a system or component, and in quantities that can adversely affect plant systems or components. This can include standing or spraying water.
Hostage Person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward JAFNPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on JAFNPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area) .
- Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.
Imminent Mitigation actions have been ineffective, additional actions are not expected to be successful, and trended information indicates that the event or condition will occur. Where imminent timeframes are specified, they shall apply. Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force. Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage .
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- Normal Plant Operations Activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from Normal Plant Operations. .
Projectile An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area which normally encompasses all controlled areas within the security protected area fence as depicted in FSAR Figure 2.1-4.
- Restore and Maintain The definition of restore and maintain is consistent with that used in the EOPs.
Sabotage Deliberate damage, misalignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment found tampered 'with or damaged due to malicious mischief may not meet the definition of Sabotage until this determination is made by security supervision. Security Condition Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A Security Condition does not involve a Hostile Action. Significant Transient AN UNPLANNED EVENT INVOLVING ANY OF THE FOLLOWING: D AUTOMATIC/MANUAL RUNBACK > 25% THERMAL POWER D ELECTRICAL LOAD REJECTION > 25% FULL ELECTRICAL LOAD D REACTOR SCRAM D ECCS INJECTION D Thermal power oscillations > 10% (peak to peak)
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- Strike Action Work stoppage within the Protected Area by a body of workers to enforce compliance with demands made on JAFNPP. The strike action must threaten to interrupt Normal Plant Operations.
Transitory Event An event in which an EAL declaration criteria may be met momentarily during the normal expected response of the plant as part of the designed plant response, or result from appropriate Operator actions, but those criteria are not met at the time of classification. Unisolable A breach or leak that cannot be promptly isolated.
- For the purposes of this definition, prompt generally means that the leak can be isolated within the 15 minute classification clock time. Isolation attempts are to be from the Control Room, credit is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach.
Example: Hanging jumpers per OP-18 with one RPS bus de-energized to reset the scram and isolate an SDIV leak. This action would be acceptable because the jumpers are in the Control Room and they can be hung within the 15 minute classification clock time. If the action were to take longer than 15 minutes, due to complications priorities, etc., the classification would be made at the 15 minute clock time and actions to isolate the breach would continue. Unplanned A parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or
- redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.
Visible Damage Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included .
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- Vital Area Any plant area which contains vital equipment. Any area which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation.
5.0 ATTACHMENTS 5.1 Attachment 1, Classification and Declaration Process 5.2 Attachment 2, EAL Classification Matrix (posted attachment) 5.3 Attachment 3, EAL Bases 5.4 Attachment 4, Fission Product Barrier Loss/ Potential Loss Matrix and Basis 5.5 Attachment 5, Additional Guidance for Classification
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- ATTACHMENT 1 Page 1 of 1 EMERGENCY CLASSIFICATION AND DECLARATION PROCESS Start NOTES Cold a.- Hot side of l Attachment 2.
- An emergency condition shall be assessed, classified, and declared within 15 minutes Ensure correct EAL chart in use. JAJ>-2 Attachments 3, 4 and 5 of the availability of l assist classification prncess. indlcations that an EAL has been met or exceeded.
Assess all EAL Initiating Conditions for applicability to plant conditions
- Entry into an emergency classification is not expected and events. for planned outages of systems or equipment in which compensatory Using IAP-2 Attachment 2, underline measures have been taken.
any part of each EAL met by current conditions. Has an o current EAL been met or conditions meet exceeded, BUT the EAL cleared prior to the
, OR~N
"-- exceed an emergency No EAL? declaration?
Yes Yes Return to START. Verify classification No AND ' Did the circle each EAL met. condition clear as a result of operator action or normal plant response? Declare the emergency classification ) as follows: 15 min notlflc;ition dock llegins upon declaration. Crew UPDATE: I am declaring a Yes {emergency classification level} at {time} based on EAL {EAL Number}, + {provide potential escalation paths}. NO CLASSIFICATION REQUIRED
- Notify NRC of transitory event in accordance with 10CFRS0.72.
- Direct EP to notify Oswego County Initiate IAP-1. and NYS.
Continuously monitor and evaluate for changing conditions .
- Apri 2018 . Page 9 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Exelon Nuclear*
- ATTACHMENT 2 Page 1 of 1 EAL CLASSIFICATION MATRIX (POSTED ATTACHMENT) 1.Side 1: Hot conditions 2.Side 2: Cold conditions 3.Figure IAP-2.1 - located in the Electronic Document Management System (EDMS) as its own document and can be found in 'Procedures* as 'IAP-2 Figure IAP-2.1 .
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- ATTACHMENT 3 - EAL BASES Category A - Abnormal Rad Release/ Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission* product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for ~mergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of Primary Containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of Primary Containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:
- 1. Offsite Rad Conditions Direct indication of effluent radiation monitoring systems provides a rapid
- assessment mechanism to determine releases in excess of classifiable limits .
Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.
- 2. Onsite Rad Conditions Sustained general area radiation levels in excess of those indicating loss of control of radioactive materials or those levels which may preclude access to vital plant areas also warrant emergency classification.
- 3. CR/CAS Radiation Sustained general area radiation levels in excess of 15 mR/hr may preclude access to areas requiring continuous occupancy also warrant emergency classification .
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- Category:
ATTACHMENT3-EALBASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the radiological effluent ODCM limits for 60 minutes or longer EAL: AU1.1 Unusual Event Any valid gaseous monitor reading > Table A-1 column "UE' for
;:: 60 min. (Note 2)
Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Table A Effluent Monitor Classification Thresholds Low Range Monitors
- Ill 0
Q) Ill cu C) Monitor STACK RX BLDG EXH REFUEL FLR EXH TURB BLDG EXH GE SEE HI RANGE NIA NIA SEE HI RANGE SAE SEE HI RANGE NIA NIA SEE HI RANGE Alert see Hi range
~9.9E5 cpm
~9.9E5 cpm
~9.9E5 cpm u
E
~5E5 cps
~2E4 cpm
~2E4 cpm
~5E4 cpm RADW BLDG EXH SEEHI RANGE SEE HI RANGE ~9.9E5 cpm ~2E4 cpm SWEFF NIA NIA M0,000 cps ~400 cps RADW EFF NIA ~200 x hi-hi trip or off- ~2 X hi-hi trip NIA "C scale hi
*sO'"
- J High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert u E
STACK ~11,600 mRlhr ~1160 mRlhr ~116 mRlhr NIA TURB BLDG EXH ~12 mRlhr * ~1.2 mRlhr
- NIA NIA RADW BLDG EXH ~33 mRlhr * ~3.3 mRlhr
- NIA N/A
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- Mode Applicability:
All ATTACHMENT 3 - EAL BASES NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments. for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The .occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation* in these features and/or controls. . The high alarm multiples are specified in AU1 .1 and AA 1.1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying
- these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate:
This EAL addresses radioactivity releases,. that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the EAL. This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The column 'UE' gaseous release values in Table A-1 represent two times the alarm setpoint of the specified monitors (ref. 1). The setpoints are established to ensure the ODCM release limits are not exceeded. (ref. 2) Instrumentation that may be used to assess this EAL is listed below: D Turbine Bldg. Exhaust Radiation Monitor: 17RM-431 and 17RM-432 D Reactor Bldg. Vent Radiation Monitors: 17RM-452A and 17RM-452B D Refuel Floor Vent Duct Radiation Monitors: 17RM-456A and 17RM-456B D RadWaste Bldg. Vent Exhaust Radiation Monitors: 17RM-458A and 17RM-458B D Stack Gas Radiation Monitors: 17RM-50A and 17RM-50B JAFNPP Basis Reference(s):
- 1. OP-31 Process Radiation Monitoring Systems
- 2. DVP-01.02 Offsite Dose Calculation Manual
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the radiological effluent ODCM limits for 60 minutes or longer. EAL: AU1 .2 Unusual Event Any valid liquid monitor reading> Table A-1 column 'UE" for
~ 60 min. (Note 2)
Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Table A Effluent Monitor Classification Thresholds Low Range Monitors u Monitor GE SAE Alert STACK SEE HI RANGE SEE HI RANGE see Hi range ;;:5E5 cps (I)
- , RX BLDG EXH N/A N/A ;;:9.9E5 cpm ;;:2E4 cpm 0
Cl) (I) REFUEL FLR EXH N/A N/A ;;:9.9E5 cpm ;;:2E4 cpm nl Cl TURB BLDG EXH SEE HI RANGE SEE HI RANGE ;;:9.9E5 cpm ;;:5E4 cpm
- RADW BLDG EXH SWEFF SEE HI RANGE N/A SEE HI RANGE N/A
;;:9.9E5 cpm
;;:40,000 cps
;;:2E4 cpm
;;:400 cps
;;:200 x hi-hi trip or off- ;;:2 X hi-hi trip "ti RADW NIA N/A scale hi
*s CT EFF
- J High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert UE STACK ;;:11,600 mR/hr ;;:1160 mR/hr ;;:116mR/hr N/A TURB BLDG ;;:12 mR/hr* ;;:1.2 mR/hr
- N/A N/A EXH RADW ;;:33 mR/hr * ;;:3_3 mR/hr
- N/A ' N/A BLDG EXH
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- Mode Applicability:
All ATTACHMENT 3 - EAL BASES NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.
- This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.
Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls. The high alarm multiples are specified in AU1 .2 and AA1 .2 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate. This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the EAL. This EAL is intended for sites that have established effluent monitoring on non-routine release pathways for which a discharge permit would not normally be prepared .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Instrumentation that may be used to assess this EAL is listed below: D RadWaste Effluent Radiation Monitor: 17RM-350 D Service Water Radiation Monitor: 17RM-351 RBCLC process monitors are not included in this EAL. These monitors detect radiation in the closed cooling water loop. Service Water monitors would detect any leak into Service Water through the heat exchangers. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from this system. JAFNPP Basis Reference(s):
- 1. OP-31 Process Radiation Monitoring Systems
- 2. DVP-01.02 Offsite Dose Calculation Manual
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- ATTACHMENT3-EALBASES Category: A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than two times the radiological effluent ODCM limits for 60 minutes or longer EAL:
AU1 .3 Unusual Event Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates> 2 x ODCM limits for~ 60 min. (Note 2) Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if 81") ongoing release is detected and the release start time is unknown. Mode Applicability:
- All NEI 99-01 Basis:
The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as .it is determined that the condition will likely exceed the applicable time. This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls. The ODCM multiples are specified in AU1 .3 and AA1 .3 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate .
- Apri 2018 Page 18 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- ATTACHMENT3-EALBASES This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc ..
JAFNPP Basis: No Additional. JAFNPP Basis Reference(s):
- 1. OP-31 Process Radiation Monitoring Systems
- 2. DVP-01.02 Offsite Dose Calculation Manual
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the radiological effluent ODCM limits for 15 minutes or longer EAL: AA1 .1 Alert Any valid gaseous monitor reading> Table A-1 column 'Alert* for
;:: 15 min. (Note 2)
Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Table A Effluent Monitor Classification Thresholds Low Range Monitors Monitor GE SAE Alert u
- UI
- I 0
Cl) UI C'CI (!) STACK RX BLDG EXH REFUEL FLR EXH TURB BLDG EXH RADW BLDG EXH SEE HI RANGE N/A N/A SEE HI RANGE SEE HI RANGE SEE HI RANGE N/A N/A SEE HI RANGE SEE HI RANGE see Hi range
~9.9E5 cpm
~9.9E5 cpm
~9.9E5 cpm
~9.9E5 cpm
~5E5 cps
.:2E4 cpm
.:2E4 cpm
~5E4 cpm
~2E4 cpm SWEFF N/A N/A .:40,000 cps ~400 cps
~200 x hi-hi trip or ~2 X hi-hi trip "C RADW N/A NIA off-scale hi
":i EFF C'
- J High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert UE STACK ~11,600 mR/hr e::1160 mR/hr ~116 mR/hr N/A TURB BLDG ~12 mR/hr * ~1.2 mR/hr
- N/A N/A EXH RADW ~33 mR/hr * ~3.3 mR/hr
- N/A N/A BLDG EXH . '
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- Mode Applicability:
All ATTACHMENT3-EALBASES NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as* it is determined that the condition will likely exceed th.e applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls. The high alarm setpoint multiples are specified in AU 1.1 and AA 1.1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The values for the gaseous effluent radiation monitors are based upon not exceeding 10 mR/hr at the site boundary as a result of the release. The values are derived from JAF-CALC-MULTl-01162. The 10.0 mR/hr value is based on a release rate not exceeding 500 mrem per year, as provided in the ODCM, prorated over 8766 hours, multiplied by 200, and rounded (500/8766 x 200 = 11.4). Since the calculated monitor readings for the Reactor, Turbine and RadWaste Building normal range monitors are in excess of the instruments upper range (1 E6) but at the very bottom of the corresponding high range instrument, an indication of 9.9E5 cpm on the normal range has been conservatively utilized. Instrumentation that may be used to assess this EAL is listed below: D Turbine Bldg. Exhaust Radiation Monitor: 17RM-431 and 17RM-432 D Reactor Bldg. Vent Radiation Monitors: 17R M-452A and 17RM-452B D Refuel Floor Vent Duct Radiation Monitors: 17RM-456A and 17RM-456B D RadWaste Bldg. Vent Exhaust Radiation Monitors: 17RM-458A and 17RM-458B D Stack Gas Radiation Monitors: 17RM-50A and 17RM-50B JAFNPP Basis Reference(s):
- 1. OP-31 Process Radiation Monitoring Systems
- 2. DVP-01.02 Offsite Dose Calculation Manual
- 3. JAF-CALC-MULTl-01162
- Apri 2018 Page 22 of 309 EP-AA-1014 Addendum 3 (Revision 0)
I __ __J
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- ATTACHMENT3-EALBASES Category: A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the radiological effluent ODCM limits for 15 minutes or longer EAL:
AA1 .2 Alert Any valid liquid monitor reading> Table A-1 column "Alert' for
~ 15 min. (Note 2)
Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Table A Effluent Monitor Classification Thresholds Low Range Monitors Monitor GE SAE Alert UE STACK SEE HI RANGE SEE HI RANGE see Hi range ~5E5 cps 1/1 RX BLDG EXH ~9.9E5 cpm ~2E4 cpm
- I N/A N/A 0
*CII 1/1 REFUEL FLR EXH N/A N/A ~9.9E5 cpm ~2E4 cpm ns
(!) TURB BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9E5 cpm ~5E4 cpm RADW BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9E5 cpm ~2E4 cpm SWEFF N/A NIA ~40,000 cps ~400 cps
~200 x hi-hi trip or off- ~2 X hi-hi trip "C ":i RADW N/A N/A scale hi er EFF
- J High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert u I=
STAC ~11,600 mR/hr ~1160 mR/hr ~116 mR/hr N/A K
~12 mR/hr * ~1.2 mR/hr
- N/A N/
TURB BLDG
~33 mR/hr * ~3.3 mR/hr
- N/A A EXH RADW N/
BLDG EXH A
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- Mode Applicability:
All ATTACHMENT 3 - EAL BASES NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls. The high alarm setpoint multiples are specified in AU1 .2 and AA1 .2 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate . This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.
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- JAFNPP Basis:
ATTACHMENT3-EALBASES Instrumentation that may be used to assess this EAL is listed below: D RadWaste Effluent Radiation Monitor: 1?RM-350 D Service Water Radiation Monitor: 1?RM-351 RBCLC process monitors are not included in this EAL. These monitors detect radiation in the closed cooling water loop. Service Water monitors would detect any leak into Service Water through the heat exchangers. Therefore, the Service Water radiation monitor adequately detects offsite radioactivity releases from this system. This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100. JAFNPP Basis Reference(s):
- 1. OP-31 Process Radiation Monitoring Systems
- 2. DVP-01.02 Offsite Dose Calculation Manual
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- Category:
ATTACHMENT3-EALBASES A - Abnormal Rad Release I Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the radiological effluent ODCM limits for 15 minutes or longer EAL AA1 .3 Alert Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits for;:: 15 min. (Note 2) Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown. Mode Applicability: All NEI 99-01 Basis: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or con_trols. The ODCM multiples are specified in AU 1.3 and AA 1.3 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an off-site dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate .
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- ATTACHMENT 3 - EAL BASES This EAL addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm
- drains, heat exchanger leakage in river water systems, etc.
JAFNPP Basis: Confirmed sample analyses in excess of two hundred times the site Offsite Dose Calculation Manual (ODCM) limits that continue for 15 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by raising the magnitude of the release by a factor of 100 over the Unusual Event level (i.e., 200 times ODCM). Prorating the 500 mRem/yr basis of the 10 CFR 20 non-occupational MPG limits for both time (8766 hr/yr) and the 200 multiplier, the associated Site Boundary dose rate would be approximately 10 mRem/hr. Two samples are not required but may be a method of assessing this EAL. If sample analysis indicates the threshold is met and nothing is done within 15 minutes to affect a release reduction, the Emergency Director can conclude that the EAL threshold is met without second sample results. The required release duration was reduced to 15 minutes in recognition of the raised severity. JAFNPP Basis Reference(s):
- 1. DVP-01.02 Offsite Dose Calculation Manual
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem thyroid COE for the actual or projected duration of the release EAL: AS 1 .1 Site Area Emergency Any valid radiation monitor reading greater than Table A-1 column
'SAE' for ~ 15 min. (Note 1)
Note 1: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results. See EAL AS1 .2. Table A Effluent Monitor Classification Thresholds Low Range Monitors Monitor GE SAE Alert UE STACK SEE HI RANGE SEE HI RANGE see Hi range ~SES cps Ill
- , RX BLDG EXH N/A N/A ~9.9ES cpm ~2E4 cpm 0
CII Ill REFUEL FLR EXH N/A N/A ~9.9ES cpm ~2E4 cpm ca (!) TURB BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9ES cpm ~SE4 cpm
,__ RADW BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9ES cpm ~2E4 cpm SWEFF N/A N/A ~40,000 cps ~400 cps
~200 x hi-hi trip or off- ~2 X hi-hi trip
,:s RADW N/A N/A scale hi
*sC" EFF
- J High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert UE STACK ~11,600 mR/hr ~1160 mR/hr ~116 mR/hr N/A TURB BLDG ~12 mR/hr * ~1.2 mR/hr
- N/A N/A EXH RADW ~33 mR/hr * ~3.3 mR/hr
- N/A N/A BLDG EXH
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- Mode Applicability:
All ATTACHMENT 3 - EAL BASES NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Since dose assessment is based on actual meteorology,* whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted,. or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. JAFNPP Basis: The values specified in this EAL were derived from JAF-CALC-MULTl-01162.
- Because of the proximity of the calculated values to the monitor bottom range, the Turbine Building and RadWaste Building values also specify that the associated normal range monitors indicate upscale to preclude declaration based upon signal noise.
If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor readings listed in Table A-1. For the purposes of this EAL, the Site Boundary for JAFNPP is defined in Figure 4.1-1 of the JAFNPP Technical Specifications (ref. 4) .
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- JAFNPP Basis Reference(s):
- 1. JAF-CALC-MUL Tl-01162 ATTACHMENT 3 - EAL BASES
- 2. OP-31 Process Radiation Monitoring Systems
- 3. DVP-01 .02 Offsite Dose Calculation Manual
- 4. JAFNPP Technical Specifications Section section 4.1.1, Figure 4.1-1 Figure A-1 JAF Site Boundary (Ref. 4) l<C CN- RIO
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- Category:
ATTACHMENT3-EALBASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem thyroid COE for the actual or projected duration of the re~~e
- EAL:
AS1 .2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mRem TEDE or 500 mRem thyroid COE at or beyond the site boundary Mode Applicability: All NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES For the purposes of this EAL, the Site Boundary for JAFNPP is defined in Figure
- 4. 1-1 of the JAFNPP Technical Specifications (ref. 2).
JAFNPP Basis Reference(s):
- 1. DVP-01.02 Offsite Dose Calculation Manual
- 2. JAFNPP Technical Specifications Section section 4. 1. 1, Figure 4. 1-1
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- ATTACHMENT 3 - EAL BASES Figure A-1 JAFNPP Site Boundary (Ref. 2) l< C: ON - m I
_... _ I
- I I
I I
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- Subcategory:
ATTACHMENT 3 - EAL BASES 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 100 mRem TEDE or 500 mRem thyroid COE for the actual or projected duration of the release EAL: AS1 .3 Site Area Emergency Field survey indicates closed window dose rate > 100 mRem/hr that is expected to continue for~ 1 hr at or beyond the site boundary OR Analyses of field survey samples indicate thyroid COE of> 500 mRem for 1 hr of inhalation at or beyond the site boundary Mode Applicability: All NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs) . Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. JAFNPP Basis: For the purposes of this EAL , the Site Boundary for JAFNPP is defined in Figure 4.1-1 of the JAFNPP Technical Specifications (ref. 2) . JAFNPP Basis Reference(s):
- 1. DVP-01 .02 Offsite Dose Calculation Manual
- 2. JAFNPP Technical Specifications Section section 4.1.1, Figure 4.1-1
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- ATTACHMENT 3 - EAL BASES Figure A-1 JAFNPP Site Boundary (Ref. 2)
UKE 00 RIO
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mRem TEDE or 5,000 mRem thyroid COE for the actual or projected duration of the release using actual meteorology EAL: AG1 .1 General Emergency Any va lid radiation monitor reading greater than Table A-1 column
'GE' for 2::: 15 min. (Note 1)
Note 1: The Emergency Director should not wait until the applicable time has elapsed , but should declare the event as soon as it is determined that the condition will likely exceed the applicable time . If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. Do not delay declaration awaiting dose assessment results . See EAL AG1 .2. Table A Effluent Monitor Classification Thresholds Low Range Monitors Monitor GE SAE Alert u STACK SEE HI RANGE SEE HI RANGE see Hi range ~5E5 cps RX BLDG EXH N/A N/A ~9.9E5 cpm ~2E4 cpm II)
- I REFUEL FLR EXH N/A N/A ~9.9E5 cpm ~2E4 cpm 0
Cl) U) TURB BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9E5 cpm ~5E4 cpm nl (!) RADW BLDG EXH SEE HI RANGE SEE HI RANGE ~9.9E5 cpm ~2E4 cpm SWEFF N/A N/A ~40,000 cps ~400 cps
~200 x hi-hi trip or off- ~2 X hi-hi trip 'C RADW N/A N/A scale hi *:; EFF C"
- i High Range Monitors
- With the corresponding low range monitor upscale Monitor GE SAE Alert u i=
STACK ~11 ,600 mR/hr ~1160 mR/hr ~116 mR/hr N/A TURB BLDG ~12 mR/hr * ~1.2 mR/hr
- N/A NIA EXH RADW ~33 mR/hr * ~3 .3 mR/hr
- N/A N/A BLDG EXH
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- Mode Applicability:
All ATTACHMENT 3 - EAL BASES NEI 99-01 Basis: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted , or may indicate that a higher classification is warranted . For this reason , emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information . If the results of these dose assessments are available when the classification is made (e .g., initiated at a lower classification level), the dose assessment results override the monitor reading EAL. JAFNPP Basis:
- The values specified in this EAL were derived from JAF-CALC-MULTl-01162 .
Because of the proximity of the calculated values to the monitor bottom range , the Turbine Building and RadWaste Building values also specify that the corresponding normal range monitors indicate upscale to preclude declaration based upon signal noise . For the purposes of this EAL, the Site Boundary for JAFNPP is defined in Figure 4.1-1 of the JAFNPP Technical Specifications (ref. 2). JAFNPP Basis Reference(s):
- 1. DVP-01.02 Offsite Dose Calculation Manual
- 2. JAFNPP Technical Specifications Section section 4.1.1 , Figure 4.1-1
- 3. JAF-CALC-MUL Tl-01162 4 OP-31 Process Radiation Monitoring Systems
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- ATTACHMENT 3 - EAL BASES Figure A-1 JAF Site Boundary (Ref. 2) 1<c ~ mo
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mRem TEDE or 5,000 mRem thyroid COE for the actual or projected duration of the release using actual meteorology EAL: AG1 .2 General Emergency Dose assessment using actual meteorology indicates doses
> 1,000 mRem TEDE or 5,000 mRem thyroid COE at or beyond the site boundary Mode Applicability:
All NEI 99-01 Basis:
- This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.
Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted , or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information . If the results of these dose assessments are available when the classification is made (e.g. , initiated at a lower classification level}, the dose assessment results override the monitor reading EAL.
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES For the purposes of this EAL, the Site Boundary for JAFNPP is defined in Figure 4.1-1 of the JAFNPP Technical Specifications (ref. 2) . JAFNPP Basis Reference(s):
- 1. DVP-01 .02 Offsite Dose Calculation Manual
- 2. JAFNPP Technical Specifications Section section 4.1.1, Figure 4.1-1 Figure A-1 JAFNPP Site Boundary (Ref. 2)
I< C Ol'*Cii. RIO
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release I Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mRem TEDE or 5,000 mRem thyroid COE for the actual or projected duration of the release using actual meteorology EAL: AG1 .3 General Emergency Field survey results indicate closed window dose rates
> 1,000 mRem/hr expected to continue for~ 1 hr at or beyond the site boundary OR Analyses of field survey samples indicate thyroid COE > 5,000 mRem for 1 hr of inhalation at or beyond the site boundary Mode Applicability:
All
- NEI 99-01 Basis:
This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs) . Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage . JAFNPP Basis: For the purposes of this EAL, the Site Boundary for JAFNPP is defined in Figure 4.1-1 of the JAFNPP Technical Specifications (ref. 2) . JAFNPP Basis Reference(s):
- 1. DVP-01 .02 Offsite Dose Calculation Manual
- 2. JAFNPP Technical Specifications Section section 4.1.1, Figure 4.1-1
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- ATTACHMENT 3 - EAL BASES Figure A-1 JAFNPP Site Boundary (Ref. 2) u t<E atn,mo
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Onsite Rad Conditions Initiating Condition: Unplanned rise in plant radiation levels EAL: AU2.1 Unusual Event Unplanned low water level alarm indicating uncontrolled water level decrease in the refueling cavity, spent fuel pool or fuel transfer canal with all irradiated fuel assemblies remaining covered by water AND Valid area radiation monitor reading increases:
- 18RIA-051-12 Spent Fuel Pool (EPIC A-1229)
- 18RIA-051-14 New Fuel Vault (EPIC A-1231)
- 1BRIA-052-30 Refuel Floor West (EPIC A-1247)
- All NEI 99-01 Basis:
This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel.
- These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.
The refueling pathway is the combination of refueling cavity, canal and spent fuel pool. While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered .
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- ATTACHMENT 3- EAL BASES For example, a refueling bridge ARM reading may increase due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Also, a monitor could in fact be properly responding to a known event involving transfer or relocation of a source, stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head. Generally, increased radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss.
For refueling events where the water level drops below the RPV flange classification would be via CU2.1. This event escalates to an Alert per AA2.1 if irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier Table for events in operating modes 1-3. JAFNPP Basis: The spent fuel pool low water level alarm setpoint is actuated by 19LS-60. Water level restoration instructions are performed in accordance with AOP-53, AOP-68 and OP-30. When the spent fuel pool and reactor cavity are connected, there could exist the possibility of uncovering irradiated fuel. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool.
- The following EPIC points supply the information for the Refuel Floor area radiation monitors used in the EAL:
- 18RIA-051-12 Spent Fuel Pool (EPIC A-1229)
- 18RIA-051-14 New Fuel Vault (EPIC A-1231)
- 18RIA-052-30 Refuel Floor West (EPIC A-1247)
JAFNPP Basis Reference(s):
- 1. AOP-53 Loss of Spent Fuel Pool, Reactor Cavity or Equipment Storage Pit Water Level
- 2. OP-32 Area Radiation Monitoring
- 3. OP-30 Fuel Pool Cooling and Cleanup System
- 4. ARP 09-3-1-9 FUEL POOL COOL & CLN UP TROUBLE
- 5. AOP-68 Spent Fuel Pool Trouble
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- Category:
ATTACHMENT 3 - EAL BASES A - Radioactivity Release / Area Radiation Subcategory: 2 - Onsite Rad Conditions Initiating Condition: Unplanned rise in plant radiation levels EAL: AU2.2 Unusual Event Unplanned valid area radiation monitor readings or survey results rise by a factor of 1000 over normal* levels.
- Normal levels can be considered as the highest reading in the past 24 hours excluding the current peak value Mode Applicability:
All NEI 99-01 Basis: This EAL addresses increased radiation levels as a result of events that have resulted
- in unplanned increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.
This EAL addresses increases in plant radiation levels that represent a loss of control of radioactive material resulting in a potential degradation in the level of safety of the plant. This EAL excludes radiation level increases that result from planned activities such as use of radiographic sources and movement of radioactive waste materials. A specific list of ARMs is not required as it would restrict the applicability of the Threshold. The intent is to identify loss of control of radioactive material in any monitored area .
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- JAFNPP Basis:
ATTACHMENT3-EALBASES It is recognized that some plant area radiation monitors may not be able to detect or display a reading that is 1,000 times NORMAL LEVELS. The intent of this IC is to rely on currently installed plant monitors and not to require design changes/backfits. In cases where an installed area radiation monitor cannot detect or display values at or above 1,000 X NORMAL LEVELS value, then survey instrument results may be used.' JAFNPP Basis Reference(s): None
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release I Rad Effluent Subcategory: 2 - Onsite Rad Conditions Initiating Condition: Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the RPV EAL: AA2.1 Alert Damage to irradiated fuel or loss of water level (uncovering irradiated fuel outside the RPV) that causes EITHER: Valid sustained Refuel Floor Exhaust Radiation Monitors 17RIS-456A or B 2: 10,000cpm (HI alarm 09-75-1-15) OR Valid sustained refuel floor rad monitor > Maximum Safe Operating Value on any of the following radiation monitors: 18RIA-051-12 18RIA-051-14 1BRIA-052-30 Mode Applicability: Spent Fuel Pool (EPIC A-1229) New Fuel Vault(EPIC A-1231) Refuel Floor West (EPIC A-1247) 1000 mR/hr 1000 mR/hr
.200,000 mR/hr All NEI 99-01 Basis:
This EAL addresses increases in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant. This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage .
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- ATTACHMENT 3 - EAL BASES Increased ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Increased background at the ventilation monitor due to water level decrease may mask increased ventilation exhaust airborne activity and needs to be considered.
While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. Escalation of this emergency classification level, if appropriate, would be based on AS1 .2 or AG1 .2. JAFNPP Basis: Area radiation levels on the refuel floor at or above the Maximum Safe Operating value are indicative of radiation fields which may limit personnel access. Access to the refuel floor (which can be by video camera) is required in order to visually observe water level in the spent fuel pool. Without access to the refuel floor, it would not be possible to determine the applicability of EAL AA2.1. The following EPIC points supply the information for the Maximum Safe __ Operating Values used in the EAL: ,
- 18RIA-051-12 Spent Fuel Pool (EPIC A-1229) 1000 mR/hr 18RIA-051-14 New Fuel Vault (EPIC A-1231) 18RIA-052-30 Refuel Floor West (EPIC A-1247)
JAFNPP Basis Reference(s): 1000 mR/hr 200,000 mR/hr
- 1. EOP-5 Secondary Containment Control
- 2. OP-32 Area Radiation Monitoring
- 3. JAFNPP EPG/SAG
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- Category:
ATTACHMENT3-EALBASES A - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Onsite Rad Conditions Initiating Condition: Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the RPV EAL: AA2.2 Alert A water level drop in the reactor refueling cavity, spent fuel pool or fuel transfer canal that will result in irradiated fuel becoming uncovered Mode Applicability: All NEI 99-01 Basis: This EAL addresses increases in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or
- substantial potential degradation in the level of safety of the plant.
These events escalate from AU2.1 in that fuel activity has been released, or is anticipated due to fuel heatup. This EAL applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage. In light of Reactor Cavity Seal failure incidents at two different PWRs and loss of water in the Spent Fuel Pit/Fuel Transfer Canal at a BWR, explicit coverage of these types is appropriate given their potential for increased doses to plant staff. JAFNPP Basis: No Additional JAFNPP Basis Reference(s): None
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- Category:
ATTACHMENT 3 - EAL BASES A - Abnormal Rad Release / Rad Effluent Subcategory: 3 - CR/CAS Radiation Initiating Condition: Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions EAL: AA3.1 Alert Dose rates > 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions: Control Room OR Central Alarm Station (CAS) OR Secondary Alarm Station (SAS) Mode Applicability: All
- NEI 99-01 Basis:
This EAL addresses increased radiation levels that: impact continued operation in areas requiring continuous occupancy to maintain safe operation or to perform a safe shutdown. The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL. The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other EAL may be involved. The value of 15mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section 111.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert. Areas requiring continuous occupancy include the Control Room, Secondary Alarm Station (SAS) and the Central Alarm Station (CAS). JAFNPP Basis: No additional JAFNPP Basis Reference(s): None
- Apri 2018 Page 50 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- ATTACHMENT 3 - EAL BASES Category C - Cold Shutdown I Refueling System Malfunction EAL GROUP: Cold Conditions (RCS temperatures 212°F); EALs in this category are applicable only in one or more cold operating modes.
Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentatiori necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, Containment Closure, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refuel, DEF - Defueled).
- The events of this category pertain to the following subcategories:
- 1. Loss of AC Power Loss of emergency plant electrical power can compromise plant safety
- system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4160 V emergency buses.
- 2. RPV Level RPV water level is a measure of inventory available to ensure adequate core cooling and, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolant that covers the reactor core. The RPV and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail.
- 3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions .
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- 4. Communications ATTACHMENT 3 - EAL BASES Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 5. Inadvertent Criticality Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.
- 6. Loss of DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital 125-Volt DC power sources .
- Apri 2018 Page 52 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to emergency buses EAL: CU1.1 Unusual Event AC power capability to emergency buses 10500 and 10600 reduced to a single power source (Table C-4) for 2: 15 min. (Note 3) such that any additional single failure would result in loss of all AC power to emergency buses. Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Table C-4 AC Power Sources Offsite
. Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- Station Service Transformer T-4 (While backfeeding from Main Transformer)
Onsite
- EDGA
- EDGB
- EDGC
- EDGD Mode Applicability:
4 - Cold Shutdown, 5 - Refuel
- Apri 2018 Page 53 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES The condition indicated by this EAL is the degradation of the off-site and on-site AC power systems such that any additional single failure would result in a loss of all ac power to emergency busses. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its emergency bus. The subsequent loss of this single power source would escalate the event to an Alert in accordance with CA 1.1. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. JAFNPP Basis: A diagram of the JAFNPP electrical distribution system is given in Figure C-1 (ref. 1). The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loads are supplied by buses 10100, 10200, 10300, 10400,and 10700. The 115 KV system provides two independent sources of off-site power to the in-house distribution system, via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two
- independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlled through the position of Oil Circuit Breakers (OCB) 10012 and 10022. A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2)
- Apri 2018 Page 54 of 309 EP-AA-1014 Addendum 3 (Revision 0)
I _J
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700.
The normal AC power (with the Main Generator on-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 10300; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively. (ref. 4) The emergency AC power source provides power to auxiliaries required for safe shutdown of the plant in the event neither the normal nor the reserve power sources are available. It consists of two independent on-site AC generating power sources, which supply the emergency buses 10500 and 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant safety. (ref. 5) An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115 kv system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established. (ref. 6) If emergency bus AC power is reduced to a single source for greater than 15 minutes, an Unusual Event is declared under this EAL. This cold condition EAL is equivalent to the hot condition loss of AC power EAL SA 1.1 .
- Apri 2018 Page 55 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- JAFNPP Basis Reference(s):
ATTACHMENT3-EALBASES
- 1. Drawing 71-002 AC Distribution
- 2. OP-44 115 KV System
- 3. OP-45 345 KV System
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer
- Apri 2018 Page 56 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3 - EALBASES Figure C-1: JAFNPP Main Single Line Diagram
--~--#4~---------L--------~---
uN,, ' oco I) ~~;~ ~ 11 5 KV 1 ~~~~ ~I ( I
#3 oao LtUHl s..
I . IOCIOt IOOt J
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~C R.IIJ.
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- Apri 2018 Page 57 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - Loss of AC.Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL: CA1 .1 Alert Loss of all offsite AND all onsite AC power (Table C-4) to emergency Buses 10500 AND 10600 for 2: 15 min. (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Table C-4 AC Power Sourc,es Offsite
- Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- Station Service Transformer T-4 (While backfeeding from Main Transformer)
Onsite
- EDGA
- EOG B
- EDGC
- EDGD Mode Applicability:
4 - Cold Shutdown, 5 - Refuel, DEF - Defueled
- Apri 2018 Page 58 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- NEI 99-01 Basis:
ATTACHMENT3-EALBASES Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, containment heat removal, spent fuel heat removal and the ultimate heat sink. The event can be classified as an Alert when in cold shutdown, refueling, or defueled mode because of the significantly reduced decay heat and lower temperature and pressure, increasing the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels / Radiological Effluent EALs. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. JAFNPP Basis: A basic diagram of the JAFNPP electrical distribution is given in Figure C-1 (ref. 1). The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loc;1ds are supplied by buses
- 10100, 10200, 10300, 10400,and 10700 .
The 115 KV system provides two independent sources of off-site power to the in-house distribution system, via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlled through the position of Oil Circuit Breakers (OCB) 10012 and 10022. A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2)
- Apri 2018 Page 59 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115.KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700.
The normal AC power (with the Main Generator on-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 10300; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively. (ref. 4) The emergency AC power source provides .power to auxiliaries required for safe shutdown of the plant in the event neither the normal nor the reserve power sources are available. It consists of two independent on-site AC generating power sources, which supply the emergency buses 10500 and 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity 'to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant
- safety. (ref. 5)
An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding -of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115 kv system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established. (ref. 6) This EAL is the cold condition equivalent of the hot condition loss of all AC power EAL SS1 .1 .
- Apri 2018 Page 60 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- JAFNPP Basis Reference(s):
ATTACHMENT 3 - EAL BASES
- 1. Drawing 71-002 AC Distribution
- 2. OP-44 115 KV System
- 3. OP-45 345 KV System
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer
- Apri 2018 Page 61 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT 3 - EAL BAS ES Figure C-1: JAFNPP Main Single Line Diagram (ref. 1)
#3
=~=--#4~---------L--------~---
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- Apri 2018 Page 62 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT3-EALBASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: RCS leakage EAL: CU2.1 Unusual Event RPV level cannot be restored and maintained> 177 in. for~ 15 min. (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Mode Applicability: 4 - Cold Shutdown NEI 99-01 Basis:
- This EAL is considered to be a potential degradation of the level of safety of the plant. The inability to maintain or restore level is indicative of loss of RCS inventory.
Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close per design should be considered applicable to this EAL if the relief valve cannot be isolated. Prolonged loss of RCS inventory may result in escalation to the Alert emergency classification level via either CA2.1 or CA3.1 .
- Apri 2018 Page 63 of 309 EP-AA-1014 Addendum 3 (Revision O)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES The difference between CU2.1 and CU2.2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In Cold Shutdown the RPV will normally be intact and RPV level is typically controlled below the elevation of the RPV flange and above the low-end of the normal control band. In the Refuel mode the RPV is not intact and any planned evolutions to lower RPV level below .
the elevation of the RPV flange must be carefully controlled. JAFNPP Basis: The condition of this EAL may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. When RPV level drops to 177 in. (low level scram setpoint), level is well below the normal control band and automatic RPS and PCIS actuations are required (ref. 1). Figure C-3 illustrates the elevations of the RPV level instrument ranges (ref. 2). JAFNPP Basis Reference(s):
- 1. EOP-2 RPV Control
- 2. Drawing S02-069
- Apri 2018 Page 64 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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FitzPatrick Annex Exelon Nuclear
- Category:
- Subcategory: 2 - RPV Level ATTACHMENT 3 - EAL BASES C - Cold Shutdown I Refueling System Malfunction Initiating Condition: RCS Leakage EAL:
CU2.2 Unusual Event Unplanned RPV level drop for~ 15 min. (Note 3) below EITHER: D 370 in. (RPV flange) D RPV level band when the RPV level band is established below the RPV flange Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable ime. Mode Applicability:
- 5 - Refuel NEl-99-01 Basis:
This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant. . Refueling evolutions that decrease RPV water level below the RPV flange are carefully planned and procedurally controlled. An unplanned event that results in water level decreasing below the RPV flange, or below the planned RPV water level for the given evolution (if the planned RPV water level is already below the RPV flange), warrants declaration of an Unusual Event due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame then it may indicate a more serious condition exists. Continued loss of RCS inventory will result in escalation to the Alert emergency classification level via either CA2.1 or CA3.1 .
- Apri 2018 Page 66 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES The difference between CU2.1 and CU2.2 deals with the RPV conditions that exist between cold shutdown and refuel modes. In cold shutdown the RCS will normally be intact and standard RPV inventory and level monitoring means are available. In the refuel mode the RCS is not intact and RPV level and inventory may be monitored by different means.
This EAL involves a decrease in RPV level below the top of the RPV flange that continues for 15 minutes due to an unplanned event. This EAL is not applicable to decreases in flooded reactor cavity level, which is addressed by AU2.1, until such time as the level decreases to the level of the vessel flange. If RPV level continues to decrease and reaches the Low-Low ECCS actuation setpoint then escalation to CA2.1 would be appropriate. JAFNPP Basis: The RPV flange is at 345 ft 8.75 in. elevation (ref. 1, 2) or 370 in. as indicated on 02-3Ll-86 Refuel Water Level (head off range +164.5 to +564.5 in.). (ref. 3) In Cold Shutdown mode, the RCS will normally be intact and standard RPV inventory and level mon'itoring means are available. In the Refuel mode, the RCS is not intact and RPV level and inventory may be monitored by different means, including the ability to monitor level visually. If RPV level cannot be monitored, drywell equipment or floor drain sump level increase must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of leakage (see EAL CU2.3). JAFNPP Basis Reference(s):
- 1. Drawing S02-069
- 2. OP-658 Shutdown Operation
- 3. Technical Support Guideline - 1 (TSG-1) Parameter Assessment
- Apri 2018 Page 67 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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FitzPatrick Annex Exelon Nuclear
- Category:
Subcategory: 2 - RPV Level ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Initiating Condition: RCS Leakage EAL: CU2.3 Unusual Event RPV level cannot be monitored with any unexplained RPV leakage indication, Table C-1 Table C-1 RPV Leakage Indications
- Drywell equipment drain sump tevel rise
- Drywell floor drain sump level rise
- ~eactor Buitding equipment drain sump level nse
- - Reactor Buitding floor drain sump tevel rise
- Torus tevel rise
- RPV make-up rate rise
- Observation of unisolabte RCS teakage Mode Applicability:
5 - Refuel NEI 99-01 Basis: This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant. Continued loss of RCS Inventory will result in escalation to the Alert emergency classification level via either CA2.1 or CA3.1
- Apri 2018 Page 69 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT 3 - EAL BASES This EAL addresses conditions in the refueling mode when normal means of core temperature indication and RPV level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.
Escalation to the Alert emergency classification level would be via either CA2.1 or CA3.1. JAFNPP Basis: RPV level in the Refuel mode is normally* monitored using 02-3Ll-86 Refuel Zone. For the purposes of this EAL, 'unexplained RPV leakage' is any leakage that has not been previously identified and cannot be attributed to known sources of c;;ollected fluid such as leakage from cooling water systems wit_hin the drywell. In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 2). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside *the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory. JAFNPP Basis Reference(s): ,
- 1. FSAR Update Section 4.10.3
- 2. OP 13D RHR - Shutdown Cooling
- Apri 2018 Page 70 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT3-EALBASES C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory EAL: CA2.1 Alert RPV level< 126.5 in. OR RPV level cannot be monitored for:::: 15 min. (Note 3) with any unexplained RPV leakage indication, Table C-1 Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Table C-1 RPV Leakage Indications
- Drywell equipment drain sump level rise
- Drywell floor drain sump level rise
- ~eactor Building equipment drain sump level nse
- Reactor Building floor drain sump level rise
- Torus level rise
- RPV make-up rate rise
- Observation of unisolable RCS leakage Mode Applicability:
4 - Cold Shutdown, 5 - Refuel
- Apri 2018 Page 71 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES This EAL serves as a precursor to a loss of ability to adequately cool the fuel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable .of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum emergency classification level of an Alert. 1*1 Condition The BWR Low-Low ECCS Actuation Setpoint/Level 2 was chosen because it is. a standard setpoint at which some available injection systems automatically start. The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier. 2nd Condition In the cold shutdown mode, normal RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available. . Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the operators would need to determine that RPV inventory loss
- was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.
The 15-minute duration for the loss of level indication was chosen because it is half of the CS2.3 Site Area Emergency EAL duration. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour per the analysis referenced in the CG2.2 basis. Therefore, this EAL meets the definition for an Alert. If RPV level continues to lower then escalation to Site Area Emergency will be via CS2, 1, CS2.2, or CS2.3 .
- Apri 2018 Page 72 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- JAFNPP Basis:
ATTACHMENT3-EALBASES The threshold RPV level of 126.5 in. is the low-low ECCS actuation setpoint (ref. 1). RPV level is normally monitored using the instruments in Figure C-3 (ref. 2). For the purposes of this EAL, 'unexpJained RPV leakage* i_s any leakage that has not been previously identified and cannot be attributed to known sources of collected fluid such as leakage from cooling water systems within the drywell. In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually. In the second condition of this EAL, all water level indication would be unavailable, and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 3). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 4). If the
- make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated' could be indicative of a loss of RPV inventory.
JAFNPP Basis Reference(s):
- 1. Technical Specifications Table 3.3.5.1-1
- 2. Drawing S02-069
- 3. FSAR Update Section 4.10.3
- 4. OP 130 RHR - Shutdown Cooling
- Apri 2018 Page 73 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS2.1 Site Area Emergency With Containment closure not established RPV level < 120.5 in. Mode Applicability: 4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis: Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted . Escalation to a General Emergency is via CG2.1 or AG1 .1. JAFNPP Basis: When RPV level decreases to 120.5 in., water level is six inches below the low-low ECCS actuation setpoint (ref. 1, 2). The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier. As applied to JAFNPP, Containment Closure is established when either Primary Containment or Secondary Containment is Operable per Section 3.6.1.1 or 3.6.4.1 of Technical Specifications (ref. 3). *
- Apri 2018 Page 75 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- JAFNPP Basis Reference(s):
ATTACHMENT3-EALBASES
- 1. Technical Support Guideline - 1 (TSG-1) Parameter Assessment
- 2. EOP-2 RPV Control
- 3. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
- 4. FSAR Update Section 4.10.3
- 5. OP 130 RHR - Shutdown Cooling
- Apri 2018 Page 76 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT3-EALBASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS2.2 Site Area Emergency With Containment closure established RPV level< 0 in. (TAF) Mode Applicability: 4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis: Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted .
. Escalation to a General Emergency is via CG2.1 or AG1 .1.
- JAFNPP Basis:
When RPV level drops below O in., the top of active fuel, core uncovery starts to occur (ref. 1, 2). As applied to JAFNPP, Containment Closure is established when either Primary Containment or Secondary Containment is Operable per Section 3.6.1.1 or 3.6.4.1 of Technical Specifications (ref. 3). JAFNPP Basis Reference(s):
- 1. Technical Support Guideline - 1 (TSG-1) Parameter Assessment
- 2. EOP-2 RPV Control
- 3. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
- 4. FSAR Update Section 4.10.3
- 5. OP-130 RHR - Shutdown Cooling
- Apri 2018 Page 77 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL: CS2.3 Site Area Emergency RPV level cannot be monitored for:::: 30 min. (Note 3) AND A loss of inventory as indicated by either of the following: D Any unexplained RPV leakage indication, Table C-1 D Erratic Source Range Monitor indication Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is dete,rmined that the condition will likely exceed the applicable time .
- Table C-1 RPV Leakage Indications*
- Drywell equipment drain sump level rise.
- Drywell floor drain sump level rise
- ~eactor Building equipment drain sump level nse
- Reactor Building floor drain sump level rise
- Torus level rise
- RPV make-up rate rise
- Observation of unisolable RCS leakage Mode Applicability:
4 - Cold Shutdown, 5 - Refuel
- Apri 2018 Page 78 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- NEI 99-01 Basis:
ATTACHMENT3-EALBASES Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach, pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted. Escalation to a General Emergency is via CG1 .2 or AG1 .1. In the cold shutdown mode, normal RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. The 30-minute duration allows sufficient time for actions to be performed to recover inventory control equipment.
- Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations
- Apri 2018 Page 79 of 309 EP-AA-1014 Addendum 3 (Revision 0)
. I
FitzPatrick Annex Exelon Nuclear
- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to RCS pressure boundary leakage or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted. For the purposes of this EAL, 'unexplained RPV leakage' is any leakage that has not been previously identified and cannot be attributed to known sources of collected fluid such as leakage from cooling water systems within the drywell. If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Sump level increases must be evaluated against other potential sources of leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 1). A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref. 2). If the make-up rate to the RPV unexplainably rises above the pre- established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory . JAFNPP Basis Reference(s):
- 1. FSAR Update Section 4.10.3
- 2. OP-130 RHR - Shutdown Cooling
- Apri 2018 Page 80 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: CG2.1 General Emergency RPV level< 0 in. (TAF) for;:: 30 min. (Note 3) AND Any Containment Challenge indication, Table C-5 Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time .
- Table C-5 Containment Challenge Indications Containment Closure not established Deflagration concentrations exist inside PC (H2 .::_ 6% AND 0 2 ~ 5%)
- Unplanned rise in PC pressure
- Secondary Containment area radiation > any Maximum Safe Operating Limit (EOP-5)
Mode Applicability: 4 - Cold Shutdown, 5 - Refuel
- Apri 2018 Page 81 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- NEI 99-01 Basis:
ATTACHMENT 3- EAL BASES This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV
- level. With the Containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE. The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers.
A number of variables can have a significant impact on heat removal capability challenging the fuel clad barrier. Examples include initial vessel level, shutdown heat removal system design. Analysis indicates that core damage may occur within an hour following continued core uncovery therefore, 30 minutes was conservatively chosen. If Containment Closure is re-established prior to exceeding the 30 minute core uncovery time limit then escalation to GE would not occur. JAFNPP Basis:
- Four conditions are associated with a challenge to Containment integrity:
D As applied to JAFNPP, Containment Closure is established when either Primary Containment or Secondary Containment is Operable per Section 3.6.1.1 or 3.6.4.1 of Technical Specifications (ref. 11)
- Apri 2018 Page 82 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT 3 - EAL BASES D In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in a deflagration concentration of dissolved gasses in the Primary Containment. However, PC monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that a deflagration concentration exists.
Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAOGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentrr;ition of approximately. 6% is considered the global deflagration concentration limit (ref. 2). Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 3) and readily recognizable because 6% hydrogen is well above the EOP-4, Primary Containment Control, entry condition (ref. 3) . The Drywell H2'02 Analyzer System samples the atmosphere in the primary . containment to detect concentrations of hydrogen and oxygen. This system consists of two redundant analyzers located in the Reactor Building with indication displayed on panels 27PCX-101A and B in the Relay Room. Each monitor is capable of detecting hydrogen with two ranges in concentrations
- from O to 10% and O to 30%, and oxygen in concentrations from O to 30%.
Control Room indication is supplied to EPIC display CAS1. Annunciation is provided on CRP 9-5 (ref. 4) .
- Apri 2018 Page 83 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES D Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability.
UnplannedPrimary Containment pressure increases indicate containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release. D Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Limit values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-5, Secondary Containment Control, (ref. 13) (see below). When RPV level drops below O in., the top of active fuel, core uncovery starts to occur (ref. 6, 7). JAFNPP Basis Reference(s):
- 1. NEI 99-01 Appendix C
- 2. BWROG EPG/SAG Revision 2, Sections PC/G
- 3. EOP-4a Primary Containment Gas Control
- 4 . FSAR section 5.2.3.14 5.* FSAR Update Table 5.2-1
- 6. Technical Support Guideline - 1 (TSG-1) Parameter Assessment
- 7. EOP-2 RPV Control
- 8. FSAR Update Section 4.10.3
- 9. OP-13D RHR - Shutdown Cooling
- 10. EOP-5 Secondary Containment Control, Table REACTOR BUILDING AREA RADIATION LEVELS
- 11. Technical Specifications Sections 3.6.1.1 or 3.6.4.1
- Apri 2018 Page 84 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT 3 - EAL BASES Table - EOP-5, Secondary Containment Control Reactor Building Area Radiation Levels REACTOR BUILDING AREA RADIATION LEVELS
,\.1AXIMUM MAXIMUM AREA INS"TRUMENT NORMAL SAFE Sp,ent Fuel Pool IBRIA-051-12 25 tnt/ht 10' .tut/ht Rooctot Building l BRIA*:>5 l-13 20 tnt/ht 10' Jnt/ht y.tJ ft eJev~t..bh Now FueJYoult l 6RIA*)5 l-14 10 tnt/ht 10' .tut/ht.
Cleilhuj) Preco.J.t.Area l BRIA*)5 l-1 5 80tnt,'ht 10' .tut/ht RWCUHeat I BR1A*)5 l-16 50 tnt/ht 10.. Jnt/ht Excha..hget R.ootn Fud Pod PurnJ> Rootn JBRIA-051-17 y:o 1nt/hr 10' .tut/ht C.obtntnihated E'.quip!neh.t l BRIA-05 l-!S 50 tnt/ht 10' .tut/ht Stoti1ge RWCU PUinpAtea l6RIA*.J51*19 30 tnt/ht 10' .tut/ht R.,: Bldg SJ.tuple Area l BRL'L*:'51-20 30 tnt/ht 10'- .tut/ht RBClCHeat JSRIA*051*21 5 .tut/ht 10-' .tut,'ht Exch>llget Atea Reactot BuildingAcce~5 lBRIA-)51-23 ~o tnt1ht- 1d' mr/ht 27.? fl eJe9.~i1ti,:,n TIP Cu bide lBRlA*0.51~24 125 .tnt/h.t 1/Y .tut/ht East HCU Area JSR1A*05 l*25 30 tnt/ht- HY .tut/ht We,tHCU~o l6RJA.('5 J-:M 3; tnt/ht- 10' .tut/ht E..ut Cte5cehl l6RIA-C'51-27 110 lnt/h.t 10' .tut/ht CRD ll.etnovnl Hiltcb IBRIA-051-18 25 tnt/ht 10~ Jnt/ht WcstC~c"!'bt 1BRIA-C'51*29 !CO 1nr;fl11 1c? rar/ht Refuel F.bot Weot l BR1A*.J5 l*:'I) JO~ mt/ht 2:dO~ tat/hi
- Apri 2018 Page 85 of 309 EP-AA-1014 Addendum 3 (Revision 0)
J
FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown I Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL: mergency RPV level cannot be monitored for~ 30 min. (Note 3) AND A loss of inventory as indicated by either of the following: D Any unexplained RPV leakage indication, Table C-1 D Erratic Source Range Monitor indication AND
- Any Containment Challenge indication, Table C-5 Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.
Table C-1 RPV Leakage Indications
- Drywell equipment drain sump level rise
- Drywell floor drain sump level rise
- ~eactor Building equipment drain sump level nse
- Reactor Building floor drain sump level rise
- Torus level rise
- RPV make-up rate rise
- *Observation of unisolable RCS leakage
- Apri 2018 Page 86 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT3-EALBASES Table C-5 Containment Challenge Indications
- Containment Closure not established
- Deflagration concentrations exist inside PC (H2 .=:_ 6% AND 0 2 .:': 5%)
- Unplanned rise in PC pressure
- Secondary Containment area radiation > any Maximum Safe Operating Limit (EOP-5)
Mode Applicability: 4 - Cold Shutdown, 5 - Refuel
- Apri 2018 Page 87 of 309 EP-AA-1014 Addendum 3 (Revision 0)
FitzPatrick Annex Exelon Nuclear
- ATTACHMENT 3 - EAL BASES
- NEI 99-01 Basis:
This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level. With the Containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE. The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers. A number of variables can have a significant impact on heat removal capability challenging the fuel clad barrier. Examples include: initial vessel level, shutdown heat removal system design. Analysis indicates that core damage may occur within an hour following continued core uncovery therefore, 30 minutes was conservatively chosen. If Containment Closure is re-established prior to exceeding the 30
- minute core uncovery time limit then escalation to GE would not occur.
Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage .
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- ATTACHMENT 3 - EAL BASES In the cold shutdown mode, normal RPV level instrumentation systems will usually be available. In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will usually be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. .
- Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
JAFNPP Basis: Four conditions are associated with a challenge to Containment integrity: D As applied to JAFNPP, Containment Closure is established when either Primary Containment or Secondary Containment is Operable per Section 3.6.1.1 or 3.6.4.1 of Technical Specifications (ref. 10)
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- D ATTACHMENT 3 - EAL BASES In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a cote uncovery could result in a deflagration concentration of dissolved gasses in the Primary Containment. However, PC . monitoring sampling should be performed to verify this assumption and a General and/or Emergency declared if it is determined that a deflagration concentration exists.
Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAOGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 2). Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion. The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 3) and readily recognizable because 6% hydrogen is well above the EOP-4, Primary Containment Control, entry condition (ref. 3) . The Drywell H2'0 2 Analyzer System samples the atmosphere in the primary containment to detect concentrations of hydrogen and oxygen. This system consists of two redundant analyzers located in the Reactor Building with
.indication displayed on panels 27PCX-101A and Bin the Relay Room. Each monitor is capable of detecting hydrogen with two ranges in concentrations from O to 10% and O to 30%, and oxygen in concentrations from O to 30%.
Control Room indication is supplied to EPIC display CAS1. Annunciation is provided on CRP 9-5 (ref. 4)
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- ATTACHMENT3-EALBASES D Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicate containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.
D Secondary Containment radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating Limit values are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-5, Secondary Containment Control, (ref. 9) (see below). If all means of RPV level monitoring are not available, the RPV inventory loss may be detected by the following: D If RPV level monitoring capability is unavailable, the RPV inventory loss must
. be detected by the leakage indications listed in Table C-1. Sump/tank level increases must be evaluated against other potential sources of leakage.
Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV (ref. 7). A Reactor Building
- equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage (ref.
8). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory .
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- JAFNPP Basis Reference(s):
- 1. NEI 99-01 Appendix C ATTACHMENT 3 - EAL BASES
- 2. BWROG EPG/SAG Revision 2, Sections PC/G
- 3. EOP-4a Primary Containment Gas Control
- 4. FSAR section 5.2.3.14
- 5. FSAR Update Table 5.2-1
- 6. Technical Support Guideline - 1 (TSG-1) Parameter Assessment
- 7. FSAR Update Section 4.10.3
- 8. OP-13D RHR - Shutdown Cooling
- 9. EOP-5 Secondary Containment Control, Table REACTOR BUILDING AREA RADIATION LEVELS
- 10. Technical Specifications Sections 3.6.1.1 or 3.6.4.1
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- ATTACHMENT3-EALBASES Table - EOP-5, Secondary Containment Control Reactor Building Area Radiation Levels REACTOR BUILrnNG AREA RADIATION LEVELS AREA INSTRLIMEi\ff MAXIMUM MAXIMUM NORl.\*1AL SAFE Spc:n,t l'Ud l'Dol !SRIA-'1)51-12 25 rru:1hr 10 ' IT!S/1:Jr Rc:;.aur ll*JHilir,;g_
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]:;l.-<;t HCUArca !8JUA-05 l -25 3,0mr/hr 10' rn:r/hr We:st EICL' A.n::-.1 18RLH)51-26 35 ffif,'hr ]I]' !Tif/hf l'::!5t cr~o:mt 18RLH)51-27 1 l(l ffili/llf l o' mr,,'llli CRD Rt!ffi;)VJI Batch !8RL',-05 l -28 25 mr,'hr 10' m:r/l1r Wc:~1 Crto~-CC!lt - ISRfA-051-29 100 m,/llr J 1}' mr/llr R.::Jlli!I fJoorwc..,1 !SJUA-'1)52-30 10' li!Ji/llf 2X)(Jl Jlli',/J:Jr
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- Category:
ATTACHMENT3-EALBASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL: CU3.1 Unusual Event Unplanned event results in RCS temperature > 212°F due to loss of decay heat removal capability Mode Applicability: 4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis: This EAL is a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the RPV usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered. Entry into cold shutdown conditions may be attained within hours of operating at power. Entry into the refueling mode procedurally may not occur for typically 100 hours or longer after the reactor has been shutdown. Thus, the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the RPV (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). In addition, the operators should be able to monitor RCS temperature and RPV level so that escalation to the alert level via CA3.1 or CA2.1 will occur if required .
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- ATTACHMENT3-EALBASES During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions that decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS/RPV temperatures depending on the time since shutdown.
JAFNPP Basis: Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212 9 F, ref. 1). These include (ref. 2): D RWR Loop Inlet Temp on 02TR-165 CRP 09-4 (if recirculation loop is in operation) D RWCU inlet Temperature on 12Tl-137 CRP 09-4 (position 1 on 12TSS-142) D FDWTR NOZZLE N4B INBD temperature on RX VESSEL TEMP recorder 02-3TR-89 [Point 1] at panel 09-21 JAFNPP Basis Reference(s):
- 1. Technical Specifications Table 1.1-1
- 2. AOP-30 Loss of Shutdown Cooling
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- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Loss of decay heat removal capability with irradiated fuel in the RPV EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for~ 15 min. (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Mode Applicability: 4 - Cold Shutdown, 5 - Refuel
- NEI 99-01 Basis:
This EAL is a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. Unlike the cold sh.utdown mode, normal means of core temperature indication and RCS level indication may not be available in the refueling mode. Redundant means of RPV level indication are therefore procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all level and temperature indication were to be lost in either the cold shutdown of refueling modes, this EAL would result in declaration of an Unusual Event if both temperature and level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via CA2.1 based on an inventory loss or CA3.1 based on exceeding its temperature criteria .
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- JAFNPP Basis:
ATTACHMENT 3*_ EAL BASES RPV water level is normally monitored using the instruments in Figure C-3 (ref. 3): Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212°F, ref. 1). These include (ref. 2): D RWR Loop Inlet Temp on 02TR-165 CRP 09-4 (if recirculation loop is in operation) D RWCU inlet Temperature on 12Tl-137 CRP 09-4 (position 1 on 12TSS-142) D FDWTR NOZZLE N4B INBD temperature on RX VESSEL TEMP recorder 02-3TR-89 [Point 1] at CRP 09-21 The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. JAFNPP Basis Reference(s):
- 1. Technical Specifications Table 1.1-1
- 2. AOP-30 Loss of Shutdown Cooling
- 3. Drawing S02-069
- Apri 2018 Page 97 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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FitzPatrick Annex Exelon Nuclear
- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL: CA3.1 Alert An unplanned event results in RCS temperature> 212°F for> Table C-3 duration OR Unplanned RPV pressure increase> 10 psig due to a loss of RCS cooling Table CsJ RCS Reheat Duration Time Limits
. If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable
- 1. RCS intact (Containment Closure N/A)
- 2. RCS not intact AND Containment Closure established
[____ 60 min.* 20 min.*
- 3. RCS not intact 0 min.
AND Containment Closure not established Mode Applicability: 4 - Cold Shutdown, 5 - Refuel
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- NEI 99-01 Basis:
ATTACHMENT3-EALBASES The first condition of this EAL addresses events in which RCS temperature exceeds the CU3.1 EAL threshold of 212DF (ref. 1) for the durations identified in Table C-3. The RCS Reheat Duration Threshold table addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when the RCS is intact. The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). The status of Containment Closure in this condition is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should. allow sufficient time to restore cooling without there being a substantial
, degradation in plant safety.
The RCS Reheat Duration Threshold table also addresses the complete loss of functions required for core cooling for greater than 20 minutes during refueling and cold shutdown modes when Containment Closure is established but the RCS is not intact. As discussed above, the RCS should be assumed to be intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible. Finally, complete loss of functions required for core cooling during refueling and cold shutdown modes when neither Containment Closure nor RCS being intact are established. The RCS is intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). No delay time is allowed because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment. The note (*) indicates that this EAL is not applicable if actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the specifi~d time frame .
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- ATTACHMENT 3 - EAL BASES In the 2nd condition, the 10 psi pressure increase addresses situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen should be 10 psi or* the lowest pressure that the site can read on installed Control Board instrumentation that is equal to or greater than 10 psi.
Escalation to Site Area Emergency would be via CS2.1, CS2.2, or CS2.3 should boiling result in significant RPV level loss leading to core uncovery. A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary unplanned excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available. The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. JAFNPP Basis:
- A 10 psig RPV pressure increase can read on:
D CRP 9-3: 06PR-61A/B and 06Pl-61A/B D CRP 9-5: 06Pl-90A/B/C and 06LR/PR-97 D EPIC page RXP
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- ATTACHMENT3-EALBASES Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specification cold shutdown temperature limit (212°F, ref. 1).
These include (ref. 2): *
- RWR Loop Inlet Temp on 02TR-165 CRP 9-4 (if recirculation loop is in operation)
- RWCU inlet Temperature on 12Tl-137 CRP 09-4 (position 1 on on 12TSS-142)
- FDWTR NOZZLE N48 INBD temperature on RX VESSEL TEMP recorder 02-3TR-89 [Point 1] afpanel 09-21 Containment Closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 4).
As applied to JAFNPP, Containment Closure is established when either Primary Containment or Secondary Containment is Operable per Section 3.6.1.1 or 3.6.4.1 of Technical Specifications (ref. 5)
- The note in the EAL is a reminder that any temperature increase above 212°F is an operating mode change from cold to hot conditions. Since each EAL is assigned one or more operating modes, the set of EALs that must be monitored must now include EALs associated with hot condition operating modes.
JAFNPP Basis Reference(s):
- 1. Technical Specifications Table 1.1-1
- 2. AOP-30 Lossof Shutdown Cooling
- 3. OP-130 RHR - Shutdwon Cooling
- 4. NEI 99-01 Appendix C
- 5. Technical Specifications Sections 3.6.1.1 and 3.6.4.1
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- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown I Refueling System Malfunction Subcategory: 4 - Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: CU4.1 Unusual Event Loss of all Table C-2 onsite (internal) communications capability affecting the ability to perform routine operations
- OR Loss of all Table C-2 offsite (external) communications capability affecting the ability to perform offsite notifications Table C-2 Communications Systems
- System Page/Party System (Gaitronics)
Onsite Offsite (internal) (external) X Sound Powered Phones X Control Room/Portable Radios X Wireless Phone System X Plant Telephone System X X RECS X Dedicated Phone Lines including X NRC Health Physics Network and FTS X 2001 Mode Applicability: 4 - Cold Shutdown, 5 - Refuel, DEF - Defueled
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- ATTACHMENT 3 - EAL BASES NEI 99-01 Basis:
The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate issues with off-site authorities. The loss of off-site communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50. 72. The availability of one method of ordinary off-site communications is sufficient to inform federal, state, and local authorities of plant issues. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to off-site locations, etc.) are being utilized to make communications possible. JAFNPP Basis: Onsite/offsite communications include one or more of the systems listed in Table S-2 (ref. 1, 2).
- Page/Party System (Gaitronics)
The page/party system (Gaitronics) is comprised of a page channel connected to loudspeakers throughout the plant and three channels. System functions allow multiple personnel to participate in a conversation on each of the channels. The page system is also used for announcements and plant alarms. The alarm mode must be initialized from the Control Room, but the conversation features are available in all emergency response facilities onsite and throughout the plant.
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- D Sound Powered Phones ATTACHMENT 3 ~ EAL BASES The sound-powered phone system allows point-to-point Communications as well as multi-point communication without interference from crosstalk. This system is normally used for maintenance and testing but can be used for conversations between individuals performing specialized tasks (e.g.,
individuals in the Control Room and a technical specialist in the Technical Support Center). This system is operational from the relay room and accessible from the TSC and Control Room. D Control Room/Portable Radios D Plant Telephone System The plant telephone systems can be used for inplant as well as outside communications. The system can be used for point-to- point or multipoint communications. Normal telephone lines are available at each emergency center. The phone systems include many automated or programmable features that improve notification and allow communications flexibility. Cellular or satellite phones are also available at various locations .
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- D RECS ATTACHMENT 3 - EAL BASES The Radiological Emergency Communications System is a dedicated telephone network to be used for communications pertaining to nuclear emergencies at JAFNPP. The RECS system is available 24 hours per day, 7 days per week and is tested by New York State periodically. The system consists of- dedicated transmission telephones providing multi-party communication in a conferencing mode. A station set. is normally located at each of the following locations:
- 1. . NY State Emergency Operations Center
- 2. NY State Warning Point
- 3. Alternate State Warning Point
- 4. State Department of Health
- 5. SEMO Regional Office
- 6. Oswego County EOC
- 7. Oswego County E-911 Center (Warning Point)
- 8. Nine Mile Point Control Rooms
- 9. Nine Mile Point TSC and EOF
- 10. JAFNPP Control Room
- 11. JAFNPP TSC
- 12. JAAFNPP EOF
- 13. SEMO Technical Resources
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- D ATTACHMENT3-EALBASES Dedicated Phone Lines Including NRC In addition to the RECS system, the following dedicated or special telephone connections exist.
- 1. Control Room to:
NRC TSC NMPNS EOF osc
- 2. TSC to:
NRC Control Room ENN Headquarters NMPNS EOF osc Alternate Operational Support Center
- 3. EOF to:
NRC TSC osc JAFNPP Radiological Coordinator Control Room
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- D ATTACHMENT3-EALBASES Health Physics Network and FTS2001 Phones This telephone system is part of the FTS2001. It is used to transmit health physics (radiological) data or other data to the NRG during an emergency.
JAFNPP facilities at which these telephones are located include:
- 1. TSC
- 2. EOF
- 3. Several FTS2001 phones at the TSC and EOF This EAL is the cold condition equivalent of the hot condition EAL SU4.2.
JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Section 7 Emergency Facilities and Equipment
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- Category:
Subcategory: ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction 5 - Inadvertent Criticality Initiating Condition: Inadvertent criticality EAL: CU5.1 Unusual Event Unplanned sustained positive period observed on nuclear instrumentation Mode Applicability:
- 4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:
This EAL addresses criticality events that occur in Cold Shutdown or Refueling modes such as fuel mis-loading events and inadvertent dilution events. This EAL indicates a potential degradation of the level of safety of the plant, warranting an Unusual Event classification. Escalation would be by Emergency Director Judgment.
- JAFNPP Basis:
Period meters A, B, C, and Don CRP 09-5 and 09-12 identify this condition as well as CRP 09-5 annunciator 09-5-2-41 SRM Period which is actuated by any one of the four SRM channels. JAFNPP Basis Reference(s):
- 1. ARP 09-5-2-41 SRM Period
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- Category:
ATTACHMENT 3 - EAL BASES C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Loss of DC Power Initiating Condition: Loss of required DC power for 15 minutes or longer EAL: CU6.1 Unusual Event
< 105 VDC bus voltage indications on all required 125 VDC buses for ~ 15 min. (Note 3)
Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Mode Applicability: 4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis The purpose of this EAL is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. The plant will routinely perform maintenance on a train related basis during shutdown periods. The required busses are the minimum allowed by Technical Specifications for the mode of operation. It is intended that the loss of the operating (operable) train is to be considered. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per CA3.1. 15 minutes was selected as a threshold to exclude transient or momentary power losses .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The 125 VDC system is illustrated in Figure C-2 (ref. 1). Two 125 V de systems are. provided to supply the station 125 VDC loads. One system includes 71 BCB-2A Battery Control Board, Battery 71 SB-1, and charger 71 BC-1A. The second system includes 71 BCB-2B Battery Control Board, Battery 71 SB-2, and charger 71 BC-1 B. (ref. 2) A low voltage condition on either Battery Bus is alarmed on CRP 09-8 annunciators 09-8-1-20 and 09-8-1-23 at <120 voe. (ref. 3, 4) If the loss of the required buses results in the inability to maintain Cold Shutdown, escalation to an Alert may be required by EAL CA3.1 due to inability to maintain the plant in cold shutdown. This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7.1. JAFNPP Basis Reference(s):
- 1. Drawing S71-068 2 . OP-43A 125 voe System
- 3. ARP 09-8-1-20 125 voe Batt A Volt Lo
- 4. ARP 09-8-1-23 125 voe Batt B Volt Lo
- 5. AOP-45 Loss of DC Power System 'A' 6 . AOP-46 Loss of DC Power System 'B'
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- ATTACHMENT 3 - EAL BASES Figure C-2: 125 VDC System (ref. 1) 71 MCC-252 71 MCC,262
(') =/\]mp. AC POWER
/IC POWER [
' 150 71 BC*lA (v) = Val'.* 71 BC-1!3 150 [ ]
I BATTERY El/\TTERY I _ta CHARGER CHARGER
~~
~
G:wJnds
'i __J p
[ rec;:]
~] 71 SB-2 71 S!l-1 125 VD DC PCT.VER 400
-:=- 125 voe C DC POWER V
-=-- BATTERY II BAT TER YB 71 BC!l 2A BATTERY CONTRQ BOARD 0 71 ilCil,2il BATTERY CONTRQ BOARD
- 09*8 09.a
~Jf-----+--i© 400 0£1-8 71 DC.A4 71 DC.El4
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- Category E - ISFSI ATTACHMENT 3 - EAL BASES EAL GROUP: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated. This includes classification based on a loaded fuel storage cask confinement boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage .
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- Category:
Sub-category: None ISFSI ATTACHMENT 3 - EAL BASES Initiating Condition: Damage to a loaded cask confinement boundary EAL: EU1 .1 Unusual Event Damage to a loaded cask confinement boundary as indicated by any Table E-1 sustained radiation reading for irradiated spent fuel in ISFSI. Table E-1 ISFSI Rad Reading Transfer Cask
- OVERPACK Overpack OVERPACK Overpack Average Surface Average Surface Dose Rates Serial Number Average Surface Dose Rates Dose Rates Serial Number (neutron+ gamma) (neutron+ gamma)
(neutron+ gamma) HI-STORM 100S(xxx) HI-STORM 1005 >80 mremlhr on the side >220 mrem/hr on the side
>20 mrern/hr on the top >260 mrem/hr on the side SIN - 0186, 0187, 0188
>40 mrem/hr on the top S/N -15, 16, 17 >32 mrem'hr at the inlet and >80 mrem/hr on the top outlel vent ducts SIN - 0307, 0308, 0309, 0310, 0311, 0312, 0690, 0691, 0692, 0693,
>100 mremlhron the side HI-STORM 1005(232) 0694, 0695 >600 mremlllr on the side
>20 mremlhr on the top >440 mrern/hr on the side
>90 mrem/hr at the inlet and >120 mremlhr on the top >60 mrem/hr on the top S/N
- 0169, 0170, 0171 outlet vent ducts S/N
- 0679, 0680, 0681, 0682, 0683 Mode Applicability:
All NEI 99-01 Basis: An Unusual Event in this EAL is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask Confinement Boundary is damaged or violated. This includes classification based on a loaded fuel storage cask Confinement Boundary loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage. JAFNPP Basis: The JAF ISFSI employs use of HI-STORM System casks for temporary onsite storage of JAF spent nuclear fuel. Degradation of the spent fuel cask radiation shield or degradation of the confinement boundary due to an operational event or environmental phenomena could result in dose rates and/or radionuclide releases that exceed normal doses. The dry storage modules are designed to standards identified in 10 CFR Part 72. The dry storage casks are routinely monitored by site Radiation Protection/Health Physics personnel, such that any
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- ATTACHMENT 3 - EAL BASES degradation would be detected. Increases in area or surface contamination levels may be indicative of degradation of the irradiated spent fuel or storage cask containment boundary.
A tip-over accident or a tornado missile could cause localized damage to the HI-STORM overpack shielding. This could cause overpack surface dose rates in the affected area to increase, but there is not expected to be a noticeable increase in the ISFSI site or controlled area boundary dose rates. Although the Cask System Safety Analysis Report (ref. 1) indicates that there are no man-made or natural phenomena that could cause failure of the MPC confinement boundary, this EAL addresses all mechanistically possible events that may occur. JAFNPP Basis Reference(s):
- 1. Holtec International Inc. Report HT-2002444, Revision 0, HI-STORM Cask System Final Safety Analysis Report, Docket 72-1014.
- 2. Holtec International Inc. Certificate of Compliance 72-1014, Amendment 0.
- 3. Holtec International Inc. Certificate of Compliance 72-1014, Amendment 1.
- 4. Holtec International Inc. Certificate of Compliance 72-1014, Amendment 2.
- 5. Holtec International Inc. Certificate of Compliance 72-1014, Amendment 5.
- 6. Holtec International Inc. Certificate of Compliance 72-1014, Amendment 8, Revision 1.
- 7. Holtec International Inc. Report HT-2002444, Revision 1, HI-STORM Cask System Final Safety Analysis Report, Docket 72-1014.
- 8. Holtec International Inc. Report HT-2002444, Revision 2, HI-STORM Cask System Final Safety Analysis Report, Docket 72-1014.
9, Holtec International Inc. Report HT-2002444, Revision 3, HI-STORM Cask System Final Safety Analysis Report, Docket 72-1014. to. Holtec International Inc. Report HT-2002444, Revision 7, HI-STORM Cask System Final Safety Analysis Report, Docket 72-1014.
- 11. Holtec International Inc. Report HT-2002444, Revision 11.1, HI- STORM Cask System Final Safety Analysis Report, Docket 72-1014 .
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- Category H - Hazards ATTACHMENT 3 - EAL BASES EAL GROUP: ANY (EALs in this category are applicable to any plant condition, hot or cold.)
Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. The events of this category pertain to the following subcategories:
- 1. Natural & Destructive Phenomena Natural events include hurricanes, earthquakes or tornados that have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.
Non-naturally occurring events that can cause damage to plant facilities and include aircraft crashes, missile impacts, etc.
- 2. Fire or Explosion Fires can pose significant hazards to personnel and reactor safety.
Appropriate for classifications are fires within the site Protected Area or which may affect operability of vital equipment.
- 3. Toxic & Flammable Gas Non-naturally occurring events that can cause damage to plant facilities and include toxic or flammable gas leaks.
- 4. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
- 5. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities .
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- 6. Judgment ATTACHMENT3-EALBASES The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:
HU1 .1 Unusual Event SEISMIC EVENT IDENTIFIED BY ANY TWO OF THE FOLLOWING: D FELT EARTHQUAKE D JAFNPP SEISMIC ACTIVITY ALARM (EPIC D-124) ACTUATED D CONFIRMATION OF EARTHQUAKE RECEIVED ON NMP-2 SEISMIC INSTRUMENTATION D NATIONAL EARTHQUAKE INFORMATION CENTER Mode Applicability: All NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Damage may be caused to some portions of the site, but should not affect
- ability of safety functions to operate.
As defined in the EPRl-sponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a "felt earthquake" is: An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant.
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES JAFNPP's seismic instrumentation consists of a laptop computer, Network Control Center (located in the Relay Room), and three recorder I accelerometer pairs with built in triggers. The recorder/ accelerometer pairs are located in the Reactor Building on the Refuel Floor, in the Drywell, and on the floor below the Torus. The recorders independently detect motion at the trigger level of 0.01 g and will start recording. When a trigger is met, an alarm is displayed via EPIC point D-124 (ref. 1). The laptop computer processes the seismic data to determine if an Operating Basis Earthquake (OBE) has occurred. The magnitude of the seismic event should be confirmed with NMP-2 based on their seismic instrumentation response or by analysis of the JAF seismic moitoring system. (ref. 2) Damage to some portions of the site may occur as a result of the felt earthquake but it should not affect the ability of safety functions to operate. This event escalates to an Alert under EAL HA 1.1 if the earthquake exceeds Operating Basis Earthquake (QBE) levels. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 2.6 Engineering Seismology
- 2. AOP-14 Earthquake
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive ph,enomena affecting the Protected Area EAL:
HU1 .2 Unusual Event Tornado striking within Protected Area boundary or winds > 90 mph Mode Applicability: All NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. This EAL is based on a tornado striking (touching down) or high winds within the Protected Area. Escalation of this emergency classification level, if appropriate, would be based on visible damage, or by other in plant conditions, via HA 1.2.
- JAFNPP Basis:
If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL HA1 .2.
- The design basis wind velocity for all structures at JAFNPP is 90 mph (ref. 1)
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- ATTACHMENT 3 - EAL BASES A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wine speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.
The Protected Area boundary is within the security isolation zone and is defined in the JAFNPP Site Security Plan (blue book). (ref. 2) JAFNPP Basis Reference(s):
- 1. FSAR Update Section 12.4.4
- 2. JAFNPP Site Security Plan
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:
HU1 .3 Unusual Event Turbine failure resulting in casing penetration or damage to turbine or generator seals Mode Applicability: All NEI 99-01 Basis: This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators .
- This EAL addresses main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant.
Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build up are appropriately classified via HU2.1 and HU3.1. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.
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- ATTACHMENT3-EALBASES Escalation of this emergency classification level, if appropriate, would be to HA 1.3 based on damage done by projectiles generated by the failure or by any radiological releases.
JAFNPP Basis: No additional JAFNPP Basis Reference(s): None
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:
HU1 .4 Unusual Event Flooding in any Table H-1 area that has the potential to affect safety-related equipment needed for the current operating mode Table H-1 Safe Shutdown Areas D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room
- D D
D D D Electric Bays Control Room AC Equipment Room Control Room Chiller Room Emergency Diesel Generator Building Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV/ADS) Mode Applicability: ALL
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps. Escalation of this emergency classification level, if appropriate, would be based on visible damage via HA 1.5, or by other plant conditions. JAFNPP Basis: Table H-1 contains the areas that contain the JAFNPP safe shutdown systems and components. (ref. 1) JAFNPP Basis Reference(s):
- 1. JAFNPP Safe Shutdown Analysis
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:
HU1 .5 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles. (Note 7) OR. ESW intake bay water level < 237 ft Note 7: A hazardous event does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability:
- ALL NEI 99-01 Basis:
This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up- river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. , This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. This EAL addresses other site specific phenomena (such as hurricane, flood, or seiche) that can also be precursors of more serious events .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The low level is based on ESW intake bay water level and corresponds to the design minimum lake level (ref. 2). JAFNPP Basis Reference(s): 1.. NEI 99-01 Revision 6 IC HU3.4
- 2. Safety Evaluation JAF-SE-93-034 "Evaluation of Maximum and Minimum Water Levels at Screenwell for Safe Operation of Class I Equipment"
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL:
HA1 .1 Alert Seismic event> QBE as indicated by EITHER: D Control Room indication of degraded performance of any system required for the safe shutdown of the plant 0 R D Analysis of the JAF seismic monitoring system AND Earthquake confirmed by ANY of the following: D Felt earthquake D National Earthquake Information Center D Confirmation of earthquake received on NMP-2 seismic instrumentation Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES This EAL escalates from HU1 .1 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other EALs (e.g., System Malfunction). Seismic events of this magnitude can result in a vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems. The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant. JAFNPP Basis:
- Ground motion acceleration of 0.08g horizontal or 0.053g vertical (2/3 the horizontal) is the Operating Basis Earthquake for JAFNPP (ref. 1).
JAFNPP's seismic instrumentation consists of a laptop computer, Network Control Center (located in the Relay Room), and three recorder/ accel,erometer pairs with built in triggers. The recorder/ accelerometer pairs are located in the Reactor Building on the Refuel Floor, in the Drywell, and on the floor below the Torus. The recorders independently detect motion at the trigger level of 0.01 g and will start recording. When a trigger is met, an alarm is displayed via EPIC point D-124 (ref. 1). The laptop computer processes the seismic data to determine if an Operating Basis Earthquake (QBE) has occurred .
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- ATTACHMENT3-EALBASES The magnitude of the seismic event should be based on indications of degraded safety system performance confirmed by one or more indications of a felt earthquake. Alternatively, determination of exceeding the QBE threshold can be performed by analysis of the JAFNPP seismic monitoring system. Procedure AOP-14 Earthquake provides the guidance for determining if the OBE earthquake threshold is exceeded. (ref. 2)
JAFNPP Basis Reference(s):
- 1. FSAR Update Section 2.6 Engineering Seismology
- 2. AOP-14 Earthquake
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive.phenomena affecting Vital Areas EAL:
HA1.2 Alert Tornado striking within Protected Area boundary or winds > 90 mph resulting in visible damage to any Table H-1 plant structures I equipment OR Control Room indication of degraded performance of Safe Shutdown systems Table H-1 Safe Shutdown Areas D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room D Electric Bays D Control Room AC Equipment Room D Control Room Chiller Room D Emergency Diesel Generator Building D Battery. Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MS IV/ADS) Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES This EAL escalates from HU1 .2 *in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems _in those structures evidenced by control indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs. This EAL is based on a tornado striking (touching down) or high winds that have caused visible damage to structures containing functions or systems required for safe shutdown of the plant. JAFNPP Basis: This EAL is based on the FSAR design basis wind velocity of 90 mph. (ref. 1)
- Personnel access to safe shutdown areas may be an important factor in monitoring and controlling equipment operability. Safe shutdown areas include structures that are in contact with or immediately adjacent to the areas that actually contains the equipment of concern (ref. 2).
The Alert classification is appropriate if relevant plant parameters indicate that the performance of safety systems in the affected safe shutdown areas has .been degraded. No attempt should *be made to fully inventory the actual magnitude of the damage or quantify the degradation of safety system performance prior to declaration of an Alert under this threshold. A tornado striking (touching down) within the Protected Area resulting in visible damage warrants declaration of an Alert regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm. The Protected Area boundary is within the security isolation zone and is defined in the JAFNPP Site Security Plan (blue book). (ref. 3)
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- JAFNPP Basis Reference(s):
- 1. FSAR Update Section 12.4.4 ATTACHMENT 3 - EAL BASES
- 2. JAFNPP Safe Shutdown Analysis
- 3. JAFNPP Site Security Plan
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL:
HA1 .3 Alert Vehicle crash resulting in visible damage to any Table H-1 plant structures or equipment OR Control Room indication of degraded performance of Safe Shutdown systems Table H-1 Safe Shutdown D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room
- D Electric Bays D Control Room AC Equipment Room D Control Room Chiller Room D Emergency Diesel Generator Building D Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MS IV/ADS)
Mode Applicability: ALL
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FitzPatrick Annex Exelon Nuclear ATTACHMENT3-EALBASES NEI 99-01 Basis: This EAL is based on the occurrence of a vehicle crash that resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs. This EAL addresses vehicle crashes within the Protected Area that results in visible damage to vital areas or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant. JAFNPP Basis:
- The Protected Area boundary is within the security isolation zone and is defined in the JAFNPP Site Security Plan (blue book). (ref. 1)
If the vehicle crash is determined to be hostile in nature, the event is classified under EAL HA4.1. JAFNPP Basis Reference(s):
- 1. JAFNPP Site Security Plan
- 2. JAFNPP Safe Shutdown Analysis
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL:
HA1 .4 Alert Turbine failure-generated projectiles result in any visible damage to or penetration of any Table H-1 area OR control room indication of degraded performance of Safe Shutdown systems Table H-1 Safe Shutdown D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room D Electric Bays D Control Room AC Equipment Room D Control Room Chiller Room D Emergency Diesel Generator Building D Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV/ADS) Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES These EALs escalate from HU1 :3 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs. This EAL addresses the threat to safety related equipment imposed by projectiles generated by main turbine. rotating component failures. Therefore, this EAL is consistent with the definition of an Alert in that the potential exists for actual or substantial potential degradation of the level of safety of the plant. JAFNPP Basis:
- The turbine generator stores large amounts of rotational kinetic energy in its rotor .
In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those h_ousing safety related equipment. In the event of missile ejection, the pmbability of a strike on a plant region is a function of the energy and direction of an ejected missile and of the orientation of the turbine with respect to the plant region. JAFNPP Basis Reference(s):
- 1. JAFNPP Safe Shutdown Analysis
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- Category: H - Hazards ATTACHMENT 3- EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL:
HA1.5 Alert Flooding in any Table H-1 area resulting in an electrical shock hazard that precludes access to operate or monitor safety equipment OR that results in degraded safety system performance as indicated in the Control Room Table H-1 Safe Shutdown D Reactor Building
- D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room
- D D
D D D Electric Bays Control Room AC Equipment Room Control Room Chiller Room Emergency Diesel Generator Building Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MSIV/ADS) Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES These EALs escalate from HU1 .4 in* that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs. This EAL addresses the effect of internal flooding caused by events sue~ as component failures, equipment misalignment, or outage activity mishaps. It is based on the degraded performance of systems, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety
. equipment. The inability to access, operate or monitor safety equipment represents an actual or substantial potential degradation of the level of safety of the plant.
Flooding as used in this EAL describes a condition where water is entering the room faster than installed equipment is capable of removal, resulting in a rise of water level within the room. Classification of this EAL should not be delayed while corrective actions are being taken to isolate the. water source
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- JAFNPP Basis:
ATTACHMENT3-EALBASES The one area identified as being vunerable to internal flooding in the JAFNPP IPE is the Relay Room. The most dominant internal flooding accident sequence is initiated by a rupture of fire protection piping in the relay room. Main steam isolation valves close due to damaged relay cabinets and possible loss of interlock relays for HPCI, RCIC, ADS and all low pressure coolant injection systems. (ref. 1) JAFNPP Basis Reference(s):
- 1. JAF-RPT-MULTl-02107 Rev. 3 JAF PSA Model 2007 Interim Update, Section 3.1
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 1 - Natural & Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting Vital Areas EAL:
HA1 .6 Alert Lake water level > 255 ft OR ESW intake bay water level < 235 ft Mode Applicability: ALL NEI 99-01 Basis: Escalation of this emergency classification level, if appropriate, would be based
- on System Malfunction EALs
- This EAL addresses other site specific phenomena such as hurricane, flood, or seiche that can also be precursors of more serious events.
JAFNPP Basis: This EAL covers high and low lake water level conditions that exceed levels which threaten vital equipment. The high lake level is based upon the revised design flood level for the screenwell interior walls and gates. (ref. 1) The low ESW intake bay water level corresponds to the top of the ESW and RHR Service Water pump suctions. (ref. 2) JAFNPP Basis Reference(s): 1.FSAR Section 2.4.3 2.Safety Evaluation JAF-SE-93-034 "Evaluation of Maximum and Minimum Water Levels at Screenwell for Safe Operation of Class I Equipmenf'
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within15 minutes of detection or explosion within Protected Area EAL:
HU2.1 Unusual Event Fire not extinguished within 15 minutes of control room notification or receipt of a valid Control Room alarm in any Table H-1 area (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed,
- but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.
Table H-1 Safe Shutdown Areas D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room D Electric Bays D Control Room AC Equipment Room D Control Room Chiller Room D Emergency Diesel Generator Building D Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MS IV/ADS) Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT3-EALBASES This EAL addresses the magnitude and extent of fires that may be potentially significant precursors of damage to safety systems. It addresses the fire, and not the degradation in performance of affected systems that may result. As used here, detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a fire is occurring, or indication of a fire detection system alarm/actuation: Verification of a fire detection system alarm/actuation includes actions that can be taken within the control room or other nearby site specific location to ensure that it is not spurious. An alarm is assumed to be an indication of a fire unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm. The intent of this 15 minute duration is to size the fire and to discriminate against small fires that are readily extinguished (e.g., *smoldering waste paper basket). Escalation of this emergency classification level, if appropriate, would be based on HA2.1 . JAFNPP Basis: No additional JAFNPP Basis Reference(s):
- 1. FSAR Update Section 12.3
- 2. JAFNPP Safe Shutdown Analysis
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- Category: H - Hazards Subcategory: 2 - Fire or Explosion ATTACHMENT 3 - EAL BASES Initiating Condition: Fire within the Protected Area not extinguished within 15 minutes of detection or explosion within Protected Area EAL:
HU2.2 Unusual Event Explosion within Protected Area boundary Mode Applicability: All NEI 99-01 Basis: This EAL addresses the magnitude and extent of explosions that may be potentially significant precursors of damage to safety systems. It addresses the explosion, and not the degradation in performance of affected systems that may result.
- This EAL addresses only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area.
No attempt _is made to assess the actual magnitude of the damage. occurrence of the explosion is sufficient for declaration. The The Emergency Director also needs to consider any security aspects of the explosion, if applicable. Escalation of this emergency classification level, if appropriate, would be based on HA2.1 .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The Protected Area boundary is within the security isolation zone and is defined in the JAFNPP Site Security Plan (blue book). (ref. 1) For this EAL, only those unanticipated explosions within the Protected Area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials.
- The occurrence of the explosion is sufficient for declaration.
- A steam line break, steam explosion or high energy line break (e.g. Feedwater) that damages surrounding permanent structures or equipment would be classified under this EAL. This does not mean the emergency is classified simply because the steam line break occurred. The method of damage is not as important as the potential degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EAL Category F).
If the explosion is determined to be hostile in nature, the event is classified under security based EALs. JAFNPP Basis Reference(s):
- 1. JAFNPP Site Security Plan
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 2 - Fire or Explosion Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown EAL:
HA2.1 Alert Fire or explosion resulting in visible damage to any Table H-1 area containing safety systems or components OR Control Room indication of degraded performance of those safe shutdown systems Table H-1 Safe Shutdown Areas D Reactor Building D Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading
- D D
D D Room Electric Bays Control Room AC Equipment Room Control Room Chiller Room Emergency Diesel Generator Building D Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MS IV/ADS) Mode Applicability: ALL
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- NEI 99-01 Basis:
ATTACHMENT3-EALBASES Visible damage is used to identify the magnitude of the fire or explosion and to discriminate against minor fires and explosions. The reference to structures containing safety systems or components is included to discriminate against fires or explosions in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the fire or explosion was large enough to cause damage to these systems. The use of visible damage should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform detailed damage assessments. The Emergency Director also needs to consider any security aspects of the explosion. Escalation of this emergency classification level, if appropriate, will be based on System Malfunctions, Fission Product Barrier Degradation or Abnormal Rad Levels / Radiological Effluent EALs .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Fire, as used in this EAL, means combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. The listed areas contain functions and systems required for the safe shutdown of the plant (ref. 2). The only explosions that should be considered are those of sufficient force to: damage permanent structures or equipment required for safe operation, or result in degraded performance of safety systems within the identified plant areas. An explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials. A steam line break or steam explosion that damages permanent structures or equipment would be classified under this EAL. The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EAL Category F).
- JAFNPP Basis Reference(s):
- 1. FSAR Update Section 12.3
- 2. JAFNPP Safe Shutdown Analysis
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 3 - Toxic & Flammable Gas Initiating Condition: Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to normal plant operations EAL:
HU3.1 Unusual Event Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect normal plant operations Mode Applicability: All NEI 99-01 Basis: This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect normal plant operations.
- The fact that SCBA may be worn does not eliminate the need to declare the event.
This EAL is not intended to require significant assessment or quantification. It assumes an uncontrolled process that has the potential to affect plant operations. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. An asphyxiant is a gas capable of reducing the level of oxygen
- in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.
Escalation of this emergency classification level, if appropriate, would be based on HA3.1 .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES As used in this EAL, affecting normal plant operations means that activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures have been impacted. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from normal plant operations and thus would be considered to have been affected. Some gases are toxic by their very nature. Others, like carbon dioxide, can be lethal if it reduces oxygen to low concentrations (asphyxiant) that are immediately dangerous to life and health (IDLH). Oxygen deficient atmospheres (less than 19.5% oxygen) are considered IDLH. NRC position is that anytime carbon dioxide is discharged in plant areas such that t.he area becomes uninhabitable, regardless of whether anyone is in the areas, conditions for classification exist. Releases occurring during planned surveillance activities or planned maintenance/tag-out activities are excluded if the release has been anticipated as a result of the planned activity and compensatory measures are in place. The following documents provide additional information on hazardous substances and spills. D AP-09.10 Hazardous Material Spill Response D AP-09.03 Oil Spill Prevention Control and Countermeasure Plan D
- Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) for Some Hazardous Chemicals Should the release affect plant safe shutdown areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1.
JAFNPP Basis Reference(s):
- 1. AP-09.10 Hazardous Material Spill Response
. 2. AP-09.03 Oil Spill Prevention Control and Countermeasure Plan
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 3 -*Toxic & Flammable Gas Initiating Condition: Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to normal plant operations EAL:
HU3.2 Unusual Event Recommendation by local, county or state officials to evacuate or shelter site personnel based on offsite event Mode Applicability: All NEI 99-01 Basis:
- JAFNPP Basis:
The release originated offsite and local, county or state officials have reported the need for evacuation or sheltering of site personnel. Offsite events (e.g., tanker truck accident releasing toxic gases, etc.) are considered in this EAL because they may adversely affect normal plant operations. State officials may determine the evacuation area for offsite spills by using the Department of Transportation (DOT) Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. Should the release affect plant safe shutdown areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1. JAFNPP Basis Reference(s): None
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 3 - Toxic & Flammable Gas Initiating Condition: Access to a vital area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of operable equipment required to maintain safe operations
- or safely shutdown the reactor EAL:
HA3.1 Alert Access to any Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or safely shutdown the reactor (Note 4) Note 4: If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event.
- D D
Reactor Building Table H-1 Safe Shutdown Control Room/ Relay Room/ Cable Run Rooms/ Cable Spreading Room
- D Electric Bays D Control Room AC Equipment Room D Control Room Chiller Room D Emergency Diesel Generator Building D Battery Rooms/Battery Room Corridor D RHRSW/ESW Pump Rooms D Cable Tunnels D Remote Safe Shutdown Panels 25ASP-4 and 25ASP-5 (for MS IV/ADS)
Mode Applicability: All
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES Gases in a Safe Shutdown Area can affect the ability to safely operate or safely shutdown the reactor. The fact that SCBA may be worn does not eliminate the need to declare the event. Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere poses an immediate threat to life and health or an immediate threat of severe exposure to gases. This could be based upon documented analysis, indication of personal ill effects from exposure, or operating experience with the hazards. If the equipment in the stated area was already inoperable, or out of service, before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shutdown beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death .
- An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene - used in welding). This EAL assumes concentrations of flammable gasses which can ignite/support combustion.
Escalation of this emergency classification level, if appropriate, will be based on System Malfunctions, Fission Product Barrier Degradation or Abnormal Rad Levels / Radioactive Effluent EALs .
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FitzPatrick Annex Exelon Nuclear le ATTACHMENT3-EALBASES JAFNPP Basis: Table H-1 safe shutdown areas contain systems that are operated to establish or maintain safe shutdown (ref. 1). This EAL does not apply to routine inerting of the Primary Containment. The following documents provide additional information on hazardous substances and spills. D AP-09.10 Hazardous Material Spill Response D AP-09.03 Oil Spill Prevention Control and Countermeasure Plan D R*egulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) for Some Hazardous Chemicals JAFNPP Basis Reference{s):
- 1. JAFNPP Safe Shutdown Analysis
- 2. AP-09.10 Hazardous Material Spill Response
- 3. AP-09.03 Oil Spill Prevention Control and Countermeasure Plan
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- Category: H - Hazards Subcategory: 4 - Security
- ATTACHMENT 3 - EAL BASES Initiating Condition: Confirmed security condition or threat which indicates a potential degradation in the level of safety of the plant EAL:
HU4.1
- Unusual Event A security condition that does not involve a hostile action as reported by the Security Shift Supervisor OR A credible site-specific security th_reat notification OR A validated notification from NRC providing information of an aircraft threat Mode Applicability:
- All NEI 99-01 Basis:
Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of
. effective Security EALs.
Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 1O' CFR 50.72. Security events assessed as hostile actions are classifiable under HA4.1, HS4.1 and HG4.1. . A higher initial classification could be made based upon the nature and timing of the security threat and_ potential consequences. The licensee shall consider upgrading the emergency response status and emergency classification level in accordance with the site's Safeguards Contingency Plan and Emergency Plan .
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- 1st Condition ATTACHMENT3-EALBASES Reference is made to site specific security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Safeguards Contingency Plan.
This threshold is based on site specific security plans. , Site specific Safeguards Contingency Plans are based on guidance provided by NEI 03-12. 2nd Condition This threshold is included to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a credible threat. Only the plant to which the specific threat is made need declare the Unusual Event. The determination of 'credible' is made through use of information found in the site specific Safeguards Contingency Plan. 3rd Condition The intent of this EAL is to ensure that notifications for the aircraft threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. It is not the intent of this EAL to replace existing non-hostile related EALs involving aircraft. This EAL is met when a plant receives information regarding an aircraft threat from NRC. Validation is performed by calling the NRC or by other approved methods of authentication. Only the plant to whi.ch the specific threat is made need declare the Unusual Event.
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- ATTACHMENT 3 - EAL BASES The NRG Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner (airliner is meant to be a large aircraft with the potential for causing significant damage to the plant). The status and size of the plane may be provided by NORAD through the NRG.
Escalation to Alert emergency classification level would be via HA4.1 would be appropriate if the threat involves an airliner within 30 minutes of the plant. JAFNPP Basis: This EAL is based on the JAFNPP Safeguards Contingency Plan (ref. 1). 1*1 Condition AOP-70, Security Threat (ref. 2) provides guidance for response to security related events based on contingency events at the JAFNPP. Reference is made to the Security Shift Supervisor because this individual is the designated on-site person who is qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the JAFNPP Safeguards Contingency Plan (Safeguards) (ref. 1, 2). 2nd Condition The determination of 'credible' is made through the use of information found in AOP-70, Security Threat (ref. 2). A threat is considered "credible' if notification of the threat has come from the NRG, FBI or local law enforcement and they consider the threat to be credible; or if the threat is deemed credible by the Security Shift Supervisor and the threat is specifically directed at the JAFNPP. Only the plant or site to which the specific threat is made need declare the Unusual Event. Guidance in these instances should be provided directly by JAFNPP Security and their sources. A higher initial classification could be made based upon the nature and timing of the threat and potential consequences .
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- 3rd Condition No additional.
ATTACHMENT 3 - EAL BASES A validated notification from the NRC of an airliner threat less than 30 min. away, or notification from the security force that a hostile action has occurred within the Owner Controlled Area would result in escalation to an Alert under EAL HA4.1 (ref. 3, 4). JAFNPP Basis Reference(s):
- 1. JAFNPP Safeguards Contingency Plan
- 2. AOP-70, Security Threat
- 3. NRC Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events*
- 4. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
- 5. AOP-70A, Airborne Security Threat*
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- Category: H - Hazards Subcategory: 4 - Security ATTACHMENT 3 - EAL BASES Initiating Condition: Hostile action within the owner controlled area or airborne attack threat EAL:
HA4.1 Alert A hostile action is occurring or has occurred within the Owner Controled Area as reported by the Security Shift Supervisor OR A validated notification from NRG of an airliner attack threat
< 30 min. away Mode Applicability:
All
- NEI 99-01 Basis:
Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs. These EALs address the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. They are not premised solely on the potential for a radiological release. Rather the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land or water attack elements. The fact that the site is under serious attack or is an identified attack target with minimal time available for further preparation or additional assistance to arrive requires a heightened state of readiness and implementation of protective measures that can be effective (such as on-site evacuation, dispersal or sheltering)
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- 1*1 Condition ATTACHMENT 3- EAL BASES This EAL addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the OCA. Those events are adequately addressed by other EALs.
Note that this EAL is applicable for any hostile action occurring, or that has occurred, in the Owner Controlled Area. This includes ISFSl's that may be outside the Protected Area but still within the Owner Controlled Area. Although nuclear plant security officers are well trained and prepared to protect against hostile action, it is appropriate for Offsite Response Organizations to be notified and encouraged to begin activation (if they do not normally) to be better prepared should it be necessary to consider further actions. 2"d Condition This EAL addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time.
- The intent of this EAL is to ensure that notifications for the airliner attack threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant.
This EAL is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is within 30 minutes of the plant. Only the plant to which the specific threat is made need declare the Alert .
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- - ATTACHMENT 3 - EAL BASES The NRG Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an airliner (airliner is meant to be a large aircraft with the potential for causing significant damage to the plant). The status and size of the plane may be provided by NORAD through the NRG.
If not previously notified by the NRG that the airborne hostile action was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRG. However, the declaration should not be unduly delayed awaiting Federal notification. JAFNPP Basis: Reference is made to the Security Shift Supervisor or designee because this individual is the designated on-site person qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the JAFNPP Safeguards Contingency Plan (Safeguar~s) information. (ref. 1) JAFNPP Basis Reference(s):
- 1. JAFNPP Safeguards Contingency Plan
- 2. AOP-70, Security Threat
- 3. NRG Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events'
- 4. NEI White Paper "Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005) -
- 5. AOP-70A, Airborne Security Threat
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- Category: H - Hazards Subcategory: 4 - Security ATTACHMENT 3 - EAL BASES Initiating Condition: Hostile Action within the Protected Area EAL:
HS4.1 Site Area Emergency A Hostile Action is occurring or has occurred within the Protected Area as reported by the Security Shift Supervisor Mode Applicability: All NEI 99-01 Basis: This condition represents an escalated threat to plant safety above that contained in the Alert in that a hostile force has progressed from the Owner Controlled Area to the Protected Area.
- This EAL addresses the contingency for a very rapid progression of events, such as that experienced on September 11*, 2001. It is not premised solely on the potential for a radiological release. Rather the issue includes the need for rapid assistance due to the possibility for significant and. indeterminate damage from
. additional air, land or water attack elements.
The fact that the site is under serious attack with minimal time available for further preparation or additional assistance to arrive requires Offsite Response Organizations readiness and preparation for the implementation of protective measures .
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- ATTACHMENT 3 - EAL BASES This EAL addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental *events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Protected Area. Those events are adequately addressed by other EALs.
Although nuclear plant security officers are well trained and prepared to protect against hostile action, it is appropriate for Offsite Response Organizations to be notified and encouraged to begin preparations for public protective actions (if they do not normally) to be better prepared should it be necessary to consider further actions. If not previously notified by NRC that the airborne hostile action was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification. Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack. JAFNPP Basis: Loss of plant control would result in escalation to a General
- Emergency under EAL HG4.1 .
Reference is made to the Security Shift Supervisor or designee because this individual is the designated on-site person qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the JAFNPP Safeguards Contingency Plan (Safeguards) information. (ref. 1)
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- JAFNPP Basis Reference(s):
ATTACHMENT3-EALBASES
- 1. JAFNPP Safeguards Contingency Plan
- 2. AOP-70, Security Threat
- 3. NRG Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events'
- 4. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
- 5. AOP-70A, Airborne Security Threat
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 4 - Security Initiating Condition: Hostile Action resulting in loss of physical control of the facility EAL:
HG4.1 General Emergency A hostile force has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions (i.e., reactivity control, RPV water level, or decay heat removal) OR A hostile action has caused failure of Spent Fuel Cooling Systems and imminent fuel damage is likely for a freshly off-loaded reactor core in pool Mode Applicability:
- ALL NEI 99-01 Basis:
1*1 Condition This EAL encompasses conditions under which a hostile action has resulted in a loss of physical control of safe shutdown areas (containing safe shutdown equipment or controls of safe shutdown equipment) required to maintain safety functions and . control of that equipment cannot be transferred to and operated from another location. Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions.
- If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the threshold is not met.
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- 2nd Condition ATTACH,MENT 3 - EAL BASES This EAL addresses failure of spent fuel cooling systems as a result of hostile action if imminent fuel damage is likely, such as when a freshly off-loaded reactor core is in the spent fuel pool.
JAFNPP Basis: If one train of a safety system is compromised, but the plant staff can fulfill a safety function with the redundant train, the EAL threshold is not met. Freshly off-loaded fuel (core) is defined as irradiated fuel that has been removed from the RPV within the previous 96 hours (ref. 5). JAFNPP Basis Reference(s):
- 1. JAFNPP Safeguards Contingency Plan
- 2. AOP-70, Security Threat
- 3. NRC Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events'
- 4. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
- 5. JAFNPP Technical Specification Bases 3.3.6.2
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- Category: H - Hazards ATTACHMENT3-EALBASES Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated EAL:
HA5.1 Alert Control Room evacuation required per AOP-43, Plant Shutdown from Outside the Control Room Mode Applicability: All NEI 99-01 Basis: With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facilities may be necessary. Inability to establish plant control from outside the control room will escalate this event to a Site Area Emergency. JAFNPP Basis: AOP-43, Plant Shutdown from Outside the Control Room, provides the instructions for scramming the unit and maintaining RCS inventory from outside the Control Room. The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. Inability to establish plant control from outside the Control Room escalates this event to a Site Area Emergency per EAL HS5.1. JAFNPP Basis Reference(s):
- 1. AOP-43, Plant Shutdown from Outside the Control Room
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- Category: H - Hazards ATTACHMENT 3 - EAL BASES Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated and plant control cannot be established EAL:
HS5.1 Site Area Emergency. Control Room evacuation has been initiated AND Control of the plant cannot be established per AOP-43, Plant Shutdown From Outside The Control Room, within 30 min. Mode Applicability: All NEI 99-01 Basis: The intent of this EAL is to capture those events where control of the plant cannot be
- reestablished in a timely manner. In this case, expeditious transfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated).
The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. Typically, these safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal- (ability to maintain a heat sink) for a BWR. The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director (ED) judgment. The. Emergency Director is expected to make a reasonable, informed judgment within the site specific time for transfer that the licensee has control of the plant from the remote shutdown panel. Escalation of this emergency classification level, if appropriate, would be by Fission Product Barrier Degradation or Abnormal Rad Levels/Radiological Effluent EALs .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The thirty minute time for transfer starts when the Control Room begins to be evacuated (not when AOP-43, Plant Shutdown from Outside the Control Room, is entered). The time interval is based on how quickly control must be reestablished without core uncovery and/or core damage. In Cold Shutdown and Refuel modes, operator concern is directed toward maintaining core cooling such as is discussed in Generic Letter 88-17, "Loss of Decay Heat Removal.' In Operating, and Hot Standby modes, operator concern is primarily directed toward maintaining safety functions and thereby assuring fission product barrier integrity. AOP-43, Plant Shutdown from Outside the Control Room, provides the instructions for scramming the unit, and maintaining RPV inventory from outside the Control Room. The Shift Manager determines if the Control Room is inoperable and requires evacuation. Control Room . inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions. JAFNPP Basis Ref~rence(s):
- 1. AOP-43, Plant Shutdown from Outside the Control Room
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- Category: H - Hazards Subcategory: 6 - Judgment ATTACHMENT 3 - EAL BASES Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of a UE EAL:
HU6.1 Unusual Event Other conditions exist that in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs Mode Applicability: All NEI 99-01 Basis:
. This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Unusual Event emergency classification level.
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- JAFNPP Basis:
ATTACHMENT3-EALBASES The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the JAFNPP Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1). JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Volume 1 Section 5, Organization
- 2. NRC Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events'
- 3. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
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- Category: H - Hazards Subcategory: 6 - Judgment ATTACHMENT 3 - EAL BASES Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an Alert EAL:
HA6.1 Alert Other conditions exist that in the judgment of the Emergency Director indicate that events are in progress or. have occurred which involve an actual or potential substantial degradation of the . level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels beyond the site boundary Mode Applicability: ALL
- NEI 99-01 Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency classification level.
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- JAFNPP Basis:
ATTACHMENT3-EALBASES The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the JAFNPP Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref.1 ). JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Volume 1 Section 5, Organization
- 2. NRG Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events*
- 3. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
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- Category: H - Hazards Subcategory: 6 - Judgment ATTACHMENT 3 - EAL BASES Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of Site Area Emergency EAL:
HS6.1 Site Area Emergency Other conditions exist that in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or likely major failures of plant functions needed for protection of the public or hostile action that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels (1 Rem TEDE and 5 Rem thyroid COE) beyond the site boundary Mode Applicability:
- All NEI 99-01 Basis:
This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for Site Area Emergency .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the JAFNPP Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of
. activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1).
JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Volume 1 Section 5, Organization
- 2. NRG Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events*
- 3. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revisetj November 1, 2005)
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- Category: H - Hazards Subcategory: 6 - Judgment ATTACHMENT 3 - EAL BASES Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of General Emergency EAL:
HG6.1 General Emergency Other conditions exist that in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels (1 Rem TEDE and 5 Rem thyroid COE) beyond the site boundary Mode Applicability: All NEI 99-01 Basis: This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for General Emergency .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the JAFNPP Emergency Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managing a major emergency (ref. 1). Releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the Site Boundary. JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Volume 1 Section 5, Organization
- 2. NRG Bulletin 2005-02 'Emergency Preparedness and Response Actions for Security-Based Events'
- 3. NEI White Paper 'Enhancements to Emergency Preparedness Programs for Hostile Action May 2005 (Revised November 1, 2005)
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- Category S - System Malfunction EAL Group:
ATTACHMENT3-EALBASES Hot Conditions (RCS temperature > 212DF); EALs in this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety. The events.of this category pertain to the following subcategories:
- 1. Loss of AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss. of onsite and offsite sources for 4160 V emergency buses.
- 2. ATWS I Criticality Events related to failure of the Reactor Protection System (RPS) to initiate and
- complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events.
For EAL classification however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to Fuel Clad, RCS and Primary Containment integrity. Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.
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- ATTACHMENT 3 - EAL BASES
- 3. Inability to Reach or Maintain Shutdown Conditions System malfunctions may lead to loss of capability to remove heat removal the reactor core and RCS.
One EAL falls into this subcategory. It is related to the failure of the plant to be brought to the required plant operating condition required by technical specifications if a limiting condition for operation (LCO) is not met.
- 4. Instrumentation/ Communications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Loss, of annunciators or indicators is in this subcategory.
Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 5. Fuel Clad Degradation During normal operation, reactor coolant fission product activity is very low.
- Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself.
Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling .
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- 6. CS Leakage ATTACHMENT 3 - EAL BASES The Reactor Vessel provides a volume for the coolant that covers the reactor core. The Reactor Vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail.
Excessive RCS leakage greater than.Technical Specification limits are utilized to indicate potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Primary containment integrity.
- 7. Loss of DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
This category includes loss of vital plant 125 VDC powersources .
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite AC power to emergency buses for 15 minutes or longer EAL: SU1 .1 Unusual Event Loss of all offsite AC power (Table S-3) to emergency buses 10500 and 10600 for~ 15 min. (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Table S-3 AC Power Scurces Offsite
- Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- Station Service TransformerT-4
{While backfeedingfrom Main Tra nsfo m1er) Onsite
- EDGA
- EDG B
- EDG C
- EDG D
- Main Generator via T-4 Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES Prolonged loss of off-site AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power to emergency busses. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power. JAFNPP Basis: A diagram of the JAFNPP electrical distribution system is given in Figure S-1 (ref. 1). The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loads are supplied by buses 10100, 10200, 10300, 10400, and 10700. The 115 KV system provides two independent sources of off-site power to the in-house distribution system, via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlleo through the position of Oil Circuit Breakers (OCB) 10012 and 10022. . A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2) The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700 .
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- ATTACHMENT 3 - EAL BASES The normal AC power (with the Main Generator on-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 10300; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively.
(ref. 4) The emergency AC power source provides power to auxiliaries required for safe shutdown of the plant in the event neither the normal nor the reserve power sources are available. It consists of two indel)endent on-site AC generating power sources, which supply the emergency buses 10500 and 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant safety. (ref. 5) An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115 kv system. It is recognized that *this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established. (ref. 6) JAFNPP Basis Reference(s):
- 1. Drawing 71-002 AC Distribution
- 2. OP-44 115 KV System
- 3. OP-45 345 KV System
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer I
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- ATTACHMENT 3 - EAL BASES Figure 5-1: JAFNPP Main Single Line Diagram (ref. 1)
N INt..-M lli. U N lf '
--7.---------L--------7.---
#4 oco IC.01:> *)
I) SYST EM A ONISI() 1 11 5 I KV I SYST'E VISION II 8 r I
#3 om
\.
- iOO!n' Ll<lH H ILL 1--~~~~~~~~~:--t..::;, /.
------y-~--------
1 345KV I 24KV :
*
- C RllA I I I.,~ T l.lo 1*10 .~. m- A GE:
- I I
=,. I .~ ,o
---11f--
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.... -~(:-----:1r:~:s:---~Lr"....
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- ~ *-
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to emergency buses EAL: SA1 .1 Alert AC power capability to emergency buses 10500 and 10600 reduced to a single power source (Table S-3) for~ 15 min. (Note 3) such that any additional single failure would result in loss of all AC power to emergency buses Note 3: The Emergency Director should not wait until the applicable time has elaP,sed, but shoula declare the event as soon as it is determined that the condition will likely exceed the applicable time .
- Table S-3 AC Power Sources Offs rte
- Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- Station Service TransformerT-4 (v\lhile backfeedingfrom Main Transformer)
Onsite
- EDGA
- EDG B
- EDG C
- EDG D
- Main Generator via T-4 Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
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- NEI 99-01 Basis:
ATTACHMENT3-EALBASES This EAL is intended to provide an escalation from EAL SU1 .1. The condition indicated by this EAL is the degradation of the off-site and on-site AC power systems such that any additional single failure would result in a loss of all AC power to the emergency buses. This condition could occur due to a loss of off-site power with a concurrent failure of all but one emergency generator to supply power to its emergency buses. Another related condition could be the loss of all off-site power and loss of on-site emergency generators with only one train of emergency buses being backfed from the unit main generator, or the loss of on-site emergency generators with only one train of emergency buses being backfed from off-site power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1 .1. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.
- JAFNPP Basis:
A basic diagram of the JAFNPP electrical distribution is given in Figure S-1 (ref. 1).
- The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loads are supplied by buses 10100, 10200, 10300, 10400, and 10700.
The 115 KV system provides two independent sources of off-site power to the in-house distribution system, via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlled through the position of Oil Circuit Breakers (OCB)- 10012 and 10022. A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2)
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- ATTACHMENT 3 - EAL BASES The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700.
The normal AC power (with the Main Generator on'-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 103Q_O; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively. (ref. 4) The emergency AC power source provides power to auxiliaries required for safe shutdown ofthe plant in the event neither the normal nor the reserve power sources are available. It consists of two indeJJendent on-site AC generating power sources, which supply the emergency buses 10500 and* 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant safety. (ref. 5)
- An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115KV system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which bac~feeding has already been established. (ref. 6)
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of emergency bus power is not restored within 15 minutes, an Alert is declared under this EAL.
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- JAFNPP Basis Reference(s):
ATTACHMENT 3 - EAL BASES
- 1. Drawing 71-002 AC Distribution
- 2. OP-44 115 KV System
- 3. OP-45 345 KV System
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer
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- ATTACHMENT 3 - EAL BASES Figure S-1: JAFNPP Main Single Line Diagram (ref. 1)
N I N ~-M I ~ UNIT T - ~~---------L--------7.!~ I SYSTEM A 11 5 I KV svsn: e I - ocn "i ONISI0"-1 1 I O VIS ION II r 0 a, LL<'.lHr loot*,; / \.
- IOU.":? H ILL
~-------------~=,
------T-~-------- ./.
1 345KV I 24KV : I *<Rn
- I I
**~-
I #1 o .~ , mm- A GE
,,,,-, --._ I II ... r \! I
=* */1 IIIG-;;;:-r--:. #1 .~ ,_,
--lit;
~.M - - -
\::::>
I l I / ,: ,
- [:OUSE*- " I , _ ,_
_ .1 _ - - - - ....!_T*- ~O::..:°"- - - - - _ L _ .. 10 11 1."
,~.. : I *~-
- 1) - - --I- - -- ~ )-,~,~-
I
) )
;-*,1~:I *-
) 1CM1 ~
I ** ) I J.: 10400 Ti< i O ) , ,_. ) 10.04 ) ) i Ou O t OL'.,14 ) ' ) uac.
'°"" J) I) 10000 10000
,,__ l ) l )
i
.. e I e o
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL: SS1 .1 Site Area Emergency Loss of all offsite and all onsite AC power (Table S-3) to emergency buses 10500 and 10600 for~ 15 min. (Note 3) Note 3: The Emergency Director should not wait until the apP.licable time has elaP,sed, but shoula declare the event as soon as it is determined that the condition will likely exceed the applicable time. Table S-3 AC Power Soorces Offsrte
- Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- Station Servrce TransformerT-4 (While backfeedrng from Main Transformer)
Onsfte
- EDGA
- EDG B
- EDG C
- EDG D
- Marn Generator via T-4 Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES Loss of all AC power to emergency buses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency buses will lead to loss of Fuel Clad, RCS, and Containment, thus this event can escalate to a General Emergency. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of off-site power. Escalation to General Emergency is via Fission Product Barrier Degradation or EAL SG1 .1. JAFNPP Basis: A basic diagram of the JAFNPP electrical distribution is given in Figure S-1 (ref. 1). Loss of all AC power compromises all plant safety systems requiring electric power. This EAL is indicated by: Loss of power for> 15 min. to all:
- D D
Reserve Station Transformer T-2 D Reserve Station Transformer T-3 T-4 back fed from Station Main Transformer T-1A/T-1 B, then Station Transformer Service AND Failure of all DGs to power any vital bus AND Failure to restore power to 10500 or 10600 in s 15 min. The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loads are supplied by buses 10100, 10200, 10300, 10400,and 10700
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- ATTACHMENT 3 - EAL BASES The 115 KV system provides two independent sources of off-site power to the in-house distribution system, via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlled through the position of Oil Circuit Breakers (OCB) 10012 and 10022. A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2)
The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700. The normal AC power (with the Main Generator on-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 10300; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively. (ref. 4)
- The emergency AC power source provides power to auxiliaries required for safe shutdown of the plant in the event neit~er the normal nor the reserve power sources are available. It consists of two independent on-site AC generating power sources, which supply the emergency buses 10500 and 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant safety. (ref. 5)
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- ATTACHMENT 3 - EAL BASES An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115KV system. It is recognized that this is not a readily available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established. (ref. 6)
The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. The interval begins when both offsite and onsite AC power are lost. Consideration should be given to operable loads necessary to remove decay heat or provide RPV makeup capability when evaluating loss of AC power to essential busses emergency buses. Even though an emergency bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or RPV makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus was the only energized bus then a Site Area Emergency per SS1 .1 should be declared .
- This EAL is the hot condition equivalent of the cold condition loss of all AC power EAL CA 1.1. When in Cold Shutdown, Refuel, or Defueled mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency buses, relative to that existing when in hot conditions .
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- JAFNPP Basis Reference(s):
ATTACHMENT3-EALBASES
- 1. Drawing 71-002 AC Distribution
- 2. OP-44 115 KV System
- 3. OP-45 345 KV System
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer
- 7. AOP-49 Station Blackout
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- Figure 5-1: JAFNPP Main Single Line Diagram (ref. 1)
#4 ATTACHMENT 3 - EAL BASES
#3 J. - - - - - - - - ?:- -
- N IN t.-M IL.t UNIF
- oca I -
SYST CM A 01V1SIO 1 11 5 I KV I SYST£M 8 VIS I~ 11 ,. I om L1 H r ~ 1001" . ) ) \,.
- 10 ~ HILL ea----------~=*
------T-~-------- --- /.
1 345KV I 24KV : II """'*
"~ TI it I I 1,10 .~. -m-- Q) AGe '
I I ....... - - I
- ,. I ~~ I I
*1 .1I:: I~
~ Y :t: ,~~ II 'J ,.. ,.
--------r--------~ "
. . . 1" .. :l :USE*- " I . . .
T* . .... IUI n - 1011' 10, w : 1 *~
-~---_- _- _ ~ - - .----~---.
- ;-**!- ], *-
I Ci.140
)
) laJ0.4 )
) I L~ _*-~~ _]
) *ro~
Tl* 10 4<0 )
)
)
1041 2' I CM40 I IX.14 ) a I I
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- Category:
ATTACHMENT3-EALBASES S -System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Prolonged loss of all offsite power and all onsite AC power to emergency buses EAL: SG1 .1 General Emergency Loss of all offsite and all onsite AC power (Table S-3) to emergency buses 10500 and 10600 AND EITHER: Restoration of at least one emergency bus within 4 hours is not likely OR RPV level cannot be restored and maintained> 0 in. (TAF) or cannot be determined Table S-3 AC Power Soorces Offsite
- Reserve Station Transformer T-2
- Reserve Station Transformer T-3
- station Service TransformerT-4 (While backfeeding from Main Transformer)
Onsite
- EDGA
- EDG B
- EDG C
- EOG D
- Main Generator via T-4 Mode Applicability:
1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
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- NEl-9901 Basis:
ATTACHMENT3-EALBASES Loss of all AC power to emergency busses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power to emergency busses will lead to loss of fuel clad, RCS, and containment, thus warranting 'declaration of a General Emergency.
- This EAL is specified *to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory:
The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.
- In addition, under these conditions, fission product barrier monitoring capability may be degraded.
JAFNPP Basis:
- A basic diagram of the JAFNPP electrical distribution is given in Figure S-1 (ref. 3).
The 4160 V buses are arranged with all AC operated emergency loads supplied by buses 10500 and 10600. The balance-of-plant loads are supplied by buses 10100, 10200, 10300, 10400, and 10700. The 115 KV system provides two independent sources of off-site power to the in-house distribution system; via Transformers T-2 and T-3. This power is the normal source of in-house power whenever the main generator is not on line. The two independent sources are Nine Mile Point Unit 1 and Lighthouse Hill Hydroelectric Station (Lighthouse-Hill 26 miles). Power flow in the 115 KV System is controlled through the position of Oil Circuit Breakers (OCB) 10012 and 10022. A Motor Operated Disconnect (MOD-10017) serves to cross-connect the two power supply lines upon the loss of one. This maintains power on both sides of the JAFNPP In-House distribution and on both 115 KV lines. (ref. 2) The reserve AC power is provided by two transformers, T2 and T3, which stepdown the 115KV to 4160 VAC. Reserve station service transformer T2 provides an alternate source of 4160 VAC power to buses 10200 and 10400. Reserve station service transformer T3 provides an alternate source of power to buses 10100 and 10300. No alternate source of power is provided for bus 10700 .
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- ATTACHMENT 3 - EAL BASES The normal AC power (with the Main Generator on-line) is provided from the main generator to transformer T4 which steps-down the 24KV to 4160 VAC. The T4 transformer supplies power to the 4160 VAC buses 10100, 10200; 10300; 10400; and 10700. Normal or reserve 4160 VAC power flows to the emergency buses 10500 and 10600 through bus tie connections from buses 10300 and 10400 respectively. (ref. 4)
The emergency AC power source provides power to auxiliaries required for safe shutdown of the plant in the event neither the normal nor the reserve power sources are available. It consists of two independent on-site AC generating power sources, which supply the emergency buses
- 10500 and 10600. Each of these independent generating sources consists of two emergency diesel generators (EDGs) operated in parallel; each source having sufficient capacity to safely shutdown the reactor, maintain the safe shutdown conditions and operate all auxiliaries necessary for plant safety. (ref. 5)
An Alternate AC power source is provided from the 345KV system for plant shutdown conditions. The power is supplied to the 4160 VAC buses by back feeding from the 345KV system via the main transformers, isolated phase bus duct, and the normal station service transformer T4. The main generator must have the disconnect links to the isolated phase bus duct removed to support this evolution. Backfeeding of the station transformer has been included to allow for those conditions in which maintenance is being performed on the station reserve transformers or 115KV system. It is recognized that this is not a readily
- available source of emergency power under emergency conditions and should only be taken credit for those conditions under which backfeeding has already been established. (ref. 6)
Indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on Emergency Director judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers. Indication of continuing core cooling degradation is manifested by an inability too restore and maintain RPV level above 0 in. (RPV level is below the top of active fuel (TAF)) (ref. 7, 8). When RPV level is at or above O in., the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below O in., the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to O in.; the level is indicative of a challenge to core cooling and the Fuel Clad barrier ..
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- ATTACHMENT 3 - EAL BASES When RPV level cannot be determined, EOPs require entry to EOP-7, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 7, 8). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted.
The instructions in EOP-7 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram-failure events) (ref. 9). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. JAFNPP Basis Reference(s):
- 1. Misc. Calculation JAF-CALC-89-012 "Determination of Required SBO Coping Duration Per NUMARC 8700
- 2. OP-44 115 KV System
- 3. Drawing 71-002 AC Distribution
- 4. OP-46A 4160 VAC and 600 VAC Normal AC Power Distribution
- 5. OP-22 Diesel Generator Emergency Power
- 6. OP-45A Backfeeding Normal Station Service Transformer
- 7. JAFNPP Plant-Specific Technical Guideline (PSTG)
- 8. EOP-2 RPV Control
- 9. EOP-7 RPV Flooding 10 . AOP-49 Station Blackout
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- ATTACHMENT 3 - EAL BAS ES Figure S-1: JAFNPP Main Single Line Diagram (ref. 3)
~::~~*~ --1---s~~ 15tKV- -
#4
°'" ) or,,, 1s10 1 I s~i7M"'; - -
o vis10N II 1- -
- r o"'
#3 L10H , so.
. ------T-~-------- ~=* /.
loot ) \.
- tO~ H ILL 1 345KV I 24KV :
1.,._,,. I 1*onc , I
- Tt;.
m-- JA GE 1 I I
- 10 .~.
woo 1... ~,.. ___Jd ~"""" I
- ..I I I ,cc
,1 *~ "- Y :f--. . woo I
- /I ,:
--------r--------~"
T1 1 1co:H r . .*
.l T4 ..- ~.... t (.t. I .
~
*** : Ji -:OUSE*- ~
,~., : 1 *-
----I_-_-_ ~ )-,.,.,-
- ,.., )
M LT 1...0 )
) ,....,
tCIJIO.t ) I
!.. - _*w
~
I TH UIC)f41Ul 0,. - l tll
,*.., )
10.0f ) )
) ,.. ,.
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/--j f-["" ,,,.., ,,... ,. -; r'-
t (.(.W ) .;) 1Dfo1 t 2
~
J; (--)
;:1
" e I
- a
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 2 - ATWS I Criticality Initiating Condition: Inadvertent criticality EAL: SU2.1 Unusual Event Unplanned sustained positive period observed on nuclear instrumentation Mode Applicability: 3 - Hot Shutdown NEI 99-01 Basis: This EAL addresses inadvertent criticality events. This EAL indicates a potential degradation of the level of safety of the plant, warranting an Unusual Event classification. This EAL excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated) . Escalation would be by the Fission Product Barrier Table, as appropriate to the operating mode at the time of the event.
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Period meters A, B, C, and Don CRP 09-5 and 09-12 identify this condition as well as CRP 09-5 annunciator 09-5-2-41 SRM Period which is actuated by any one of the four SRM channels. Escalation to higher emergency classification levels would be by the Category F, Fission Product Barrier Degradation EALs, or the judgment EALs in Category H (EAL HA6.1, HS6.1 or HG6.1 ). JAFNPP Basis Reference(s):
- 1. ARP 09-5-4-2-1 SRM Period
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 2 - A TWS / Criticality Initiating Condition: Automatic Scram fails to shutdown the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor EAL: SA2.1 Alert Automatic scram fails to reduce reactor power < 2.5% (APRM downscale) AND Manual scram actions taken at the reactor control console successfully shutdown the reactor as indicated by reactor power
< 2.5% (APRM downscale)
Mode Applicability: 1 - Power Operations, 2 - Startup
- NEI 99-01 Basis:
Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor. This condition indicates failure of the automatic protection system to scram the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient. Thus the plant safety has been compromised because design limits of the fuel may have been exceeded. An Alert is indicated because conditions may exist that lead to potential loss of fuel clad or RCS and because of the failure of the Reactor Protection System to automatically shutdown the plant. If manual actions taken at the reactor control console fail to shutdown the reactor, the event would escalate to a Site Area Emergency .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The first condition of this EAL identifies the need to cease critical reactor operations by actuation of the automatic Reactor Protection System (RPS) scram function. A
- reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints.
A reactor scram may be the result of manual or automatic action in response to any of the following parameters (ref. 2): D Neutron Monitoring System Scrams (SRM High High with shorting links removed, IRM upscale trip/lNOP, APRM flow-biased and fixed upscale/lNOP) D High reactor pressure D Reactor low water level D Turbine stop valve closure D Turbine control valve fast closure D Main steamline isolation valve closure D Scram discharge volume high level D High drywell pressure Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period.
- The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion to bring the reactor power below the APRM downscale setpoint.
This EAL indicates a failure of the automatic RPS scram function to rapidly insert a sufficient number of control rods to achieve reactor shutdown. The significance of this condition, therefore, is that a potential degradation of a safety system exists because a front line automatic protection system did not function in response to a plant transient. Thus, plant safety has been compromised .
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- ATTACHMENT 3 - EAL BASES Following any automatic RPS scram signal, AOP-1 (ref. 3) and EOP-3 (ref. 4) prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Alert.
This EAL is not applicable. if a manual scram is initiated and no RPS setpoints are exceeded. Taking the mode switch to shutdown is a manual scram action. When the Mode Switch is taken out of the Run position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated. In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 2.5% (ref. 5), the event escalates to the Site Area Emergency under EAL SS2.1. If by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed (such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine if the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, then consideration should be given to evaluating the fuel for potential damage, and the reporting requirements of 50.72 should be considered for the transient event. JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. Technical Specifications section 3.3.1.1 RPS Instrumentation
- 3. AOP-1 Reactor Scram
- 4. EOP-3 Failure to Scram
- 5. EOP-2 RPV Control
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 2 - ATWS I Criticality Initiating Condition: Automatic Scram fails to shutdown the reactor and manual actions taken from the reactor control console are not successful in _shutting down the reactor EAL:
$S2.1 Site Area Emergency Automatic scram fails to reduce reactor power < 2.5% (APRM downscale)
AND Manual scram actions taken at the reactor control console do not shutdown the reactor as indicated by reactor power~ 2.5% Mode Applicability: 1 - Power Operations, 2 - Startup NEI 99-01 Basis:
- Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS.
Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly
.inserted into the core and shuts down the reactor.
Manual scram actions are not considered successful if action away from the reactor control console is required to scram (trip) the reactor. This EAL is still applicable even if actions taken away from the reactor control console are successful in shutting the reactor down because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shutdown the plant. Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core- cooling or heat removal.
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- JAFNPP Basis:
ATTACHMENT3-EALBASES This *EAL addresses any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed. Reactor shutdown achieved by use of the EP-3 actions does not constitute a successful manual scram (ref. 2). For the purpose of emergency classification at the Site Area Emergency level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., manual scram pushbuttons, Mode Switch or ARI). The APRM downscale trip setpoint (2.5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine if reactor power is greater than 2.5% power (ref. 3). The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat
- poses a direct threat to the Fuel Clad and RCS barriers and warrants declaration of a Site Area Emergency. Although this EAL may be viewed as redundant to the Category F, Fission Product Barrier Degradation, EAL FS1 .1, its inclusion is necessary to better assure timely recognition and emergency response.
Escalation of this event to a General Emergency would be under EAL SG2.1 or Emergency Director judgment. JAFNPP Basis Reference(s):
- 1. EP-3 Backup Control Rod Insertion
- 2. EOP-3 Failure to Scram
- 3. EOP-2 RPV Control
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic Scram and all manual actions fail to shutdown the reactor and indication of an extreme challenge to the ability to cool the core exists EAL: SG2.1 General Emergency Automatic and all manual scrams were not successful after any RPS setpoint is exceeded AND Reactor power is ~ 2.5% (APRM downscale) AND EITHER: RPV level cannot be restored and maintained> MSCRWL (-19 in.) (EOP-3) or cannot be determined OR Torus temperature and RPV pressure cannot be maintained below the HCTL (EOP-11) .
- Mode Applicability:
1 - Power Operations, 2 - Startup NEI 99-01 Basis: Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. For BWRs, the extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level as described in the EOP bases.
- For BWRs, considerations include inability to remove heat via the main condenser, or via the suppression pool or torus (e.g., due to high pool water temperature).
In the event either of these challenges exists at a time that the reactor has not been brought below the power associated with the safety system design a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier table declaration to permit maximum off-site intervention time .
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- JAFNPP Basis:
ATTACHMENT 3 EAL BASES This EAL addresses the following: D Any automatic reactor scram signal followed by by failure of the automatic scram and all subsequent manual scrams that fail to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SS2.1 ), and D Indications that either core cooling is extremely challenged or heat removal is extremely challenged. The APRM downscale trip setpoint (2.5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative* of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to
- determine if reactor power is greater than 2.5% power (ref. 2, 3) .
The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers .
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- ATTACHMENT 3 - EAL BASES Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above -19 in. (or cannot be determined). -19 in. is the Minimum Steam Cooling RPV Water Level (MSCRWL). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence. When RPV level cannot be determined, EOPs require entry to EOP-7, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level* are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-7 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressures.
The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
- D Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR D Suppression chamber pressure above the Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
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- ATTACHMENT 3 - EAL BASES The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. This threshold is met when the final step of section Torus Temperature in EOP-4, Primary Containment Control, is reached (ref. 4, 6). In addition to the* Torus temperature and pressure limits, Torus water level must be within HCTL limits or HCTL is exceeded. This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature.
In the event the challenge to either core cooling or heat removal occurs at a time when the reactor has not been brought below the power associated with safety system design power, a core melt sequence may exist and rapid degradation of the fuel clad could begin. To permit maximum offsite intervention time, the General Emergency declaration is therefore appropriate in anticipation of an inevitable General Emergency declaration due to loss and potential loss of fission product barriers. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 7.2
- 2. EOP-3 Failure to Scram
- 3. EOP-2 RPV Control
- 4. EOP-4 Primary Containment Control
- 5. EOP-7 RPV Flooding 6 . EOP-11 EOP and SAOG Graphs
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 3 - Inability to Reach or Maintain Shutdown Conditions Initiating Condition: Inability to reach required shutdown within Technical Specification limits EAL: SU3.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO action statement completion time Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: Limiting Conditions of Operation (LCOs) require the plant to be brought to a required operating mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is. within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate Unusual Event is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of an Unusual Event is based on the time at which the LCD-specified action statement time period elapses under the site Technical Specifications and is not related to how- long a condition may have existed. JAFNPP Basis: No additional JAFNPP Basis Reference(s):
- 1. Technical Specifications
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 4 - Instrumentation / Communications Initiating Condition: Unplanned loss of safety system annunciation or indication in the control room for 15 minutes or longer EAL: SU4.1 Unusual Event Unplanned loss of> approximately 75% of the annunciators OR indicators associated with safety systems on the following panels (CRPs) for 2:: 15 min. (Note 3):
- D 09-3 D 09-4 D 09-5 D 09-6 D 09-7 D 09-8 D 09-75
- Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.
Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
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- NEI 99-01 Basis:
ATTACHMENT 3 - EAL BASES This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is further recognized that' most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on EAL SU3.1. 15 minutes was selected as a threshold to exclude transient or momentary power losses. This Unusual Event will be escalated to an Alert based on a concurrent loss of EPIC indications or if a significant transient is in progress during the loss of annunciation or indication .
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- JAFNPP Basis:
ATTACHMENT3-EALBASES The availability of computer-based monitoring capability (i.e., EPIC) is not a factor at the Unusual Event emergency classification level. Safety system annunciation and indication considered in this EAL is found on Control Room Panels (CRP) 09-3, 09-4, 09-5, 09-6, 09-7 09-8 and 09-75. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 1). The judgment of the Shift Manager should be used as the threshold for determining the severity of the plant conditions. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 7.19
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 4 - Instrumentation / Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL: SU4.2 Unusual Event
*Loss of all Table S-2 onsite (internal) communications capability affecting the ability to perform routine operations OR Loss of all Table S-2 offsite (external) communications capability affecting the ability to perform offsite notifications Table S-2 Communications Systems
- System Page/Party System (Gaitronics)
Sound Powered Phones Onsite Offs ite (internal) (external) X X Control Room/Portable Radios X Wireless Phone System X Plant Telephone System X X RECS X Dedicated Phone Lines including X NRC Health Physics Network and FTS X 2001
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- Mode Applicability:
ATTACHMENT3-EALBASES 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate issues with off-site authorities. The loss of off-site communications ability is expected to be significantly m,ore comprehensive than the condition addressed by 10 CFR 50.72. The availability of one method of ordinary off-site communications is sufficient to inform federal, state, and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from non-routine radio transmissions, individuals being sent to off-site locations, etc.) are being used to make communications possible. JAFNPP Basis: Onsite/offsite communications include one or more of the systems listed in
- Table S-2 (ref. 1, 2) .
D Page/Party System (Gaitronics) The page/party system (Gaitronics) is comprised of a page channel connected to loudspeakers throughout the plant and three channels. System functions allow multiple personnel to participate in a conversation on each of the channels. The page system is also used for announcements and plant alarms. The alarm mode must be initialized from the Control Room, but the conversation features are available in all emergency response facilities onsite and throughout the plant.
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- D Sound Powered Phones ATTACHMENT3-EALBASES The sound-powered phone system allows point-to-point Communications as well as multi-point communication without interference from crosstalk. This system is normally used for maintenance and testing but can be used for conversations between individuals performing specialized tasks (e.g.,
individuals in the Control Room and a technical specialist in the Technical Support Center). This system is operational from the relay room and accessible from the TSC and Control Room. D Control Room/Portable Radios D Plant Telephone System The plant telephone systems can be used for inplant as well as outside communications. The system can be used for point-to- point or multipoint communications. Normal telephone lines are available at each emergency center. The phone systems include many automated or programmable features that improve notification and allow communications flexibility. Cellular or satellite phones are also available at various locations .
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- D RECS ATTACHMENT3-EALBASES The Radiological Emergency Communications System is a dedicated telephone network to be used for communications pertaining to nuclear emergencies at JAFNPP. The RECS system is available 24 hours per day, 7 days per week and is tested by New York State periodically. The system consists of dedicated transmission telephones providing multi-party communication in a conferencing mode. A station set is normally located at each of the following locations:
- 1. NY State Emergency Operations Center
- 2. NY State Warning Point
- 3. Alternate State Warning Point
- 4. State Department of Health
- 5. SEMO Regional Office
- 6. Oswego County EOC
- 7. Oswego County E-911 Center (Warning Point)
- 8. Nine Mile Point Control Rooms
- 9. Nine Mile Point TSC and EOF
- 10. JAFNPP Control Room
- 11. JAFNPP TSC
- 12. JAAFNPP EOF
- 13. SEMO Technical Resources
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- D ATTACHMENT 3 - EAL BASES Dedicated Phone Lines Including NRC In addition to the RECS system, the following dedicated or special telephone connections exist.
- 1. Control Room to:
NRC TSC NMPNS EOF osc
- 2. TSC to:
NRC Control Room ENN Headquarters NMPNS EOF osc Alternate Operational Support Center
- 3. EOF to:
NRC TSC osc JAFNPP Radiological Coordinator Control Room
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- D ATTACHMENT 3 - EAL BASES Health Physics Network and FTS2001 Phones This telephone system is part of the FTS2001. It is used to transmit health physics (radiological) data or other data to the NRC during an emergency.
JAFNPP facilities at which these telephones are located include:
- 1. TSC
- 2. EOF
- 3. Several FTS2001 phones at the TSC and EOF This EAL is the hot condition equivalent of the cold condition EAL CU4.1.
JAFNPP Basis Reference(s):
- 1. JAFNPP Emergency Plan Section 7 Emergency Facilities and Equipment
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 4 - Instrumentation/ Communications Initiating Condition: Unplanned loss of safety system annunciation or indication in the control room with EITHER (1) a Significant Transient in progress, OR (2) compensatory indicators unavailable EAL: SA4.1 Alert Unplanned loss of> approximately 75% of the annunciators OR indicators associated with safety systems on the following panels (CRPs) for;;:: 15 min. (Note 3): D 09-3 D 09-4 D 09-5 D 09-6
- D 09-7 D 09-8 D 09-75 AND EITHER:
Any significant transient is in progress, Table S-1 OR EPIC is unavailable Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. 1 Table S-11 SDgniificant Transients 11 I Re actor scram Auto/Manua I R'u nb.ack > 2:5% ther:n"131 I p ewer Elecnical <load rejection > 25Sii !full electrical load ECCS injec:tiar:, Thermal po*1.ie r osdlla1ions > 1 0% (pe ak-'li::l -peak)
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- Mode Applicability:
ATTACHMENT 3 - EAL BASES 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a SIGNIFICANT TRANSIENT. Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation. It is further recognized that most plant designs provide redundant safety system
- indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50. 72. If the .shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on EAL SU3.1. "Compensatory indications" in this context includes computer based information such as EPIC. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required.
15 minutes was selected as a threshold to exclude transient or momentary power losses. This Alert will be escalated to a Site Area Emergency if the operating crew cannot monitor the transient in progress due to a concurrent loss of compensatory indications with a significant transient in progress during the loss of annunciation or indication .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES EPIC serves as a redundant compensatory indicator which may be utilized in lieu of normal control room indicators (ref. 1). Safety system annunciation and indication considered in this EAL is found on Control Room Panels (CRPs) 09-3, 09-4, 09-5, 09-7, 09-8, and 09-75. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 2). The judgment of the Shift Manager should be used as the threshold for determining the severity of the plant conditions.
- Significant transients are listed in Table S-1 and include response to automatic or .
manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater. If the operating crew cannot monitor the transient in progress, the Alert escalates to a Site Area Emergency under EAL SS4.1. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 7 .16
- 2. FSAR Update Section 7.19
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- Category: S - System Malfunction ATTACHMENT 3- EAL BASES Subcategory: 4 - Instrumentation I Communications Initiating Condition: Inability to monitor a significant transient in progress EAL:
SS4.1 Site Area Emergency Loss of > approximately 75% of the annunciators OR indicators associated with safety systems on the following panels (CRPs) for 2: 15 min. (Note 3): D 09-3 D 09-4 D 09-5 D 09-6 D 09-7 D 09-8 D 09-75
- AND EPIC is unavailable AND Any significant transient is in progress, Table S-1 Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.
Tab:f e S-1. Signficant Transients Reactor scram Auto/M.an'Ua~ Riu nlJ.ack ~=-- 25% tl:!ternu I p c,*11er Elecit!Jie3 I lf ,cad rejiectio,n > 25% run: el'e ctricaJ' lo.ad EGCS injecll.:ioo Th,ern:ial po'lteroscilla1ions > 10% {peak-ti-;pe.ak:1
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- Mode Applicability:
ATTACHMENT3-EALBASES 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: This EAL is intended to recognize the threat to plant safety associated with the complete loss of capability of the control room staff to monitor plant response to a significant transient.
"Planned" and 'UNPLANNED' actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.
Quantification is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not-intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation. It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the Unusual Event is based on EAL SU3.1. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public while a significant transient is in progress. Site specific indications needed to monitor safety functions necessary for protection of the public must include control room indications, computer generated indications and dedicated annunciation capability .
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- ATTACHMENT3-EALBASES "Compensatory indications" in this context includes computer based information such as EPIC. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.
15 minutes was selected as a threshold to exclude transient or momentary power losses. JAFNPP Basis: The availability of computer-based monitoring capability (i.e., EPIC) is a factor at the Site Area Emergency classification level because EPIC is a compensatory non-alarming indication (ref.1 ). EPIC serves as a redundant compensatory indicator which may be utilized in lieu of normal control room indicators. Safety system annunciation and indication considered in this EAL is found on Control Room Panels (CRPs) 09-3, 09-4, 09-5, 09-7, 09-8, and 09-75. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 2).
- The third condition of the EAL addresses the loss of ability to monitor any single EOP parameter such as the loss of all RPV water level indication or the loss of all suppression pool temperature indication. Such a loss would impact the ability to use the EOPs and to maintain plant safety functions.
Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater. Due to the limited number of safety systems in operation during Cold Shutdown, Refuel and Defueled modes, this EAL is not applicable during these modes of operation. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 7.16
- 2. FSAR Update Section 7.19
- 3. EOP-2 RPV Control
- 4. EOP-3 Failure to Scram
- 5. EOP-4 Primary Containment Control
- 6. EOP-5 Secondary Containment Control
- 7. EOP-6 Radioactivity Release Control
- 8.
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- Category:
ATTACHMENT 3 - EAL BASES S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL: SU5.1 Unusual Event Offgas radiation ~ hi-hi alarm Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. Escalation of this EAL to the Alert level is via the Fission Product Barrier EALs. This threshold addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity. JAFNPP Basis: Elevated off-gas radiation activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The Technical Specification allowable limit is 600,000 µCi/sec (recombiner discharge gross noble gases beta and/or gamma) (ref.3). The hi-hi radiation alarm setpoint is set at an equivalent s 500,000 µCi/sec (ref. 1). The hi-hi radiation alarm setpoint has been conservatively selected because it is operationally significant and is readily recognizable by Control Room operating staff. The hi-hi offgas radiation alarm is set at 1000 mR/hr on 17RM-150 A and B.
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- JAFNPP Basis Reference(s):
ATTACHMENT 3 - EAL BASES
- 1. DVP-01.02 Offsite Dose Calculation Manual Specification 3.6.1
- 2. OP-31 Process Radiation Monitoring
- 3. Technical Specification 3.7.5
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL: SU5.2 Unusual Event Coolant activity> 2 µCi/gm 1-131 equivalent Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.
- Escalation of this EAL to the Alert level is via the Fission Product Barrier EALs.
This threshold addresses coolant samples specifications for transient iodine spiking limits. exceeding coolant technical JAFNPP Basis: This EAL addresses reactor coolant samples exceeding coolant Technical Specifications for iodine spiking. The value of 2 µCi/gm 1-131 equivalent is the temporary upper limit for operation allowed by Technical Specifications (ref. 1) to ensure that should a MSLB accident occur resulting offsite doses remain within a small fraction of the 10CFR100 limits. JAFNPP Basis Reference(s):
- 1. Technical Specification 3.4.6
- 2. Technical Specifications Bases 3.4.6
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- ATTACHMENT3-EALBASES Category: S - System Malfunction Subcategory: 6 - RCS Leakage Initiating Condition: RCS leakage EAL:
SU6.1 Unusual Event Unidentified or pressure boundary leakage > 10 gpm OR Identified leakage> 25 gpm Note 5:See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due to RCS Leakage Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown
- NEI 99-01 Basis:
This EAL is included as an Unusual Event because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close per design should be considered applicable to this EAL if the relief valve cannot be isolated and the resulting leakge constitutes either identified or unidentified leakage per Technical Specifications. A stuck open SRV discharging to the suppresion pool does not constitute leakage for the purpose of this EAL. The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this EAL to the Alert level is via Fission Product Barrier Degradation EALs .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Leakage is monitored by utilizing the following techniques: D Sensing excess flow in piping systems D Sensing pressure and temperature changes in the primary containment D Monitoring for high flow and temperature through selected drains, D Sampling airborne particulate and gaseous radioactivity. D Drywell floor and equipment drain sump leak rate alarm system Identified leakage rate is that portion of the total leakage rate which is routed to the drywell equipment drain sump (ref. 1). Unidentified and pressure boundary leakage rate is that portion of the total leakage rate received in the drywell floor drain sump. The unidentified and pressure boundary leakage limit is less than the identified leakage limit because of the possibility that most of the leakage might be emitted from a single crack in the RCS barrier (ref. 1). Escalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FA 1.1. The note has been added to remind the EAL-user to
- review Table F-1 for possible escalation to higher emergency classifications .
JAFNPP Basis Reference(s):
- 1. FSAR Update Section 4.1 O
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- Category:
ATTACHMENT3-EALBASES S - System Malfunction Subcategory: 7 - Loss of DC Power Initiating Condition:Loss of all safety-related DC power for 15 minutes or longer EAL: SS7.1 Site Area Emergency Loss of all DC power based on 71 BCB-2A and 71 BCB-28 bus voltage indications < 105 VDC for~ 15 min. (Note 3) Note 3: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:
- Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system.
15 minutes was selected as a threshold to exclude transient or momentary power losses. Escalation to a General Emergency would occur by Abnormal Rad Levels/Radiological Effluent, Fission Product Barrier Degradation .
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES The 125 VDC system is illustrated in Figure S-2 (ref. 1). Two 125 V de systems are provided to supply the station 125 VDC loads. One system includes 71 BCB-2A Battery Control Board, Battery 71 SB-1, and charger 71 BC-1A. The second system includes 71 BCB-2B Battery Control Board, Battery 71 SB-2, and charger 71 BC-1 B. (ref. 2). A low voltage condition on either Battery Bus is alarmed on CRP 09-8 annunciators 09-8-1-20 and 09-8-1-23 at <120 VDC. (ref. 3, 4) Escalation to a General Emergency would occur by EALs in Category R - Abnormal Rad Release/ Rad Effluent, Category F - Fission Product Barrier Degradation, and Category H - Hazards EAL HG6.1 Gudgment). This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU6.1. JAFNPP Basis Reference(s):
- 1. Drawing S71-068
- 2. OP-43A 125 voe System
- 3. ARP 09-8-1-20 125 4.
5. 6. ARP 09-8-1-23 125 AOP-45 Loss of DC AOP-46 Loss of DC voe Batt A Volt Lo voe Batt B Volt Lo Power System 'A' Power System 'B'
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- ATTACHMENT3-EALBASES Figure S-2: 125 VDC System (ref. 1) 71 MCC-252 71 MCC-2d2 A)= Amp, AC Pa,VER [] AC roWER 150 71 BC*1A C0 =- Vot~~
71 BC*l!l 150 [ ]
]~ BATTERY CHARGER
!l.~TTERY CHARGER I
~I~
~ ~
G-ounds
~
- I
~
OC Pa,VER 400 71 SB*1 125 voe BATTERY A
~ OC POWER 71 BC8-2A BATTERY CONTRa. OOARO 71 BC8*2B BATTERY CONTRa. OOARD og.a [ J400 og.a 09*8 400 [ ]
0B01 u 0~8 71 OC-A4
~~
71 DC.84
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- ATTACHMENT3-EALBASES Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature> 212DF); EALs in this category are applicable only in one or more hot operating modes.
EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: A. Fuel Clad (FC): Zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the FC barrier. B. Reactor Coolant System (RCS): The reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations, and all primary systems directly connected to the RPV up to the outermost primary containment isolation valve comprise the RCS barrier. C. Primary containment (PC): The drywell, the torus, their
- respective interconnecting paths, and other connections up to and including the outermost primary containment isolation valves comprise the PC barrier .
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- ATTACHMENT 3 - EAL BASES The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). 'Loss' and 'Potential Loss' signify the relative damage and threat of damage to the barrier. 'Loss' means the barrier no longer assures containment of radioactive materials.
'Potential Loss' means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:
Unusual Event: Any loss or any potential loss of Primary Containment Alert: Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency: Loss or potential loss of any two barriers
- General Emergency:
Loss of any two barriers and loss or potential loss of third barrier
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- ATTACHMENT 3 - EAL BASES The logic used for emergency classification based on fission product barrier monitoring should reflect the following considerations:
D The Fuel Clad barrier and the RCS barrier are weighted more heavily than the Primary Containment barrier. UE EALs associated with RCS and Fuel Clad barriers are addressed under System Malfunction EALs. D At the Site Area Emergency level, there must be some ability to dynamically
- assess how far present conditions are from the threshold for a General Emergency. For example, if Fuel Clad and RCS barrier *1oss* EALs existed, that, in addition to offsite dose assessments, would require continual assessments of radioactive inventory and containment integrity. Alternatively, if both Fuel Clad and RCS barrier 'Potential Loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.
D The ability to escalate to higher emergency classes as an event deteriorates must be ma*intained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety .
- D The Primary Containment Barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment Barrier status is addressed by Technical Specifications.
Determine which combination of the three barriers are lost or have a potential loss and use FU1 .1, FA1 .1, FS1 .1 and FG1 .1 to classify the event. Also an event or multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is imminent (i.e., within 1 to 2 hours). In this imminent loss situation, use judgment and classify as if the thresholds are exceeded .
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- Category:
ATTACHMENT3-EALBASES Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of Primary Containment EAL: FU1 .1 Unusual Event Any loss or any potential loss of Primary Containment (Table F-1) Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: None JAFNPP Basis:
- Fuel Clad, RCS and Primary Containment comprise the fission product barriers.
Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. Fuel Clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Fuel Clad and RCS barriers, the loss of either of which results in an Alert (EAL FA 1.1 ), loss of the Primary Containment barrier in and of itself does not result in the relocation of radioactive materials or the potential for degradation of core cooling capability. However, loss or potential loss of the Primary Containment barrier in combination with the loss or potential loss of either the Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1. JAFNPP Basis Reference(s): None
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- Category:
Subcategory: N/A ATTACHMENT 3 - EAL BASES Fission Product Barrier Degradation Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1 .1 Alert Any loss or any potential loss of either Fuel Clad or RCS (Table F-1) Mode Applicability: 1 -Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: None JAFNPP Basis: Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Primary Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Primary Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1. JAFNPP Basis Reference(s): None
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- *Category:
ATTACHMENT3-EALBASES Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL: FS1 .1 Site Area Emergency Loss or potential loss of any two barri,ers (Table F-1) Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: None
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- JAFNPP Basis:
ATTACHMENT 3 - EAL BASES Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefor!:! appropriate for any combination of the following conditions: D One barrier loss and a second barrier loss (i.e., loss - loss) D One barrier loss and a second barrier potential loss (I.e., loss
- potential loss)
D One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Primary Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent. JAFNPP Basis Reference(s): None
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- Category:
ATTACHMENT 3 - EAL BASES Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL: FG1 .1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1) Mode Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis: None
- JAFNPP Basis:
Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions: D Loss of Fuel Clad, RCS and Primary Containment barriers D Loss of Fuel Clad and RCS barriers with potential loss of Primary Containment barrier D Loss of RCS and Primary Containment barriers with potential loss of Fuel Clad barrier D Loss of Fuel Clad and Primary Containment barriers with potential loss of RCS barrier JAFNPP Basis Reference(s): None
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- Introduction Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Primary Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Loss column) lists the
- categories (types) of fission product barrier thresholds. The fission product barrier categories are:
A. RPV Level B. PC Pressure/ Temperature C. Isolation D . Rad
- E. Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories. The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word
'None' is entered in the cell.
Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its category title and number. For example, the first Fuel Clad barrier Loss in Category A would be assigned 'FC Loss A.1,' the third Primary Containment barrier Potential Loss would be assigned 'PC P-Loss B.3,' etc .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL-user determines that any threshold has been exceeded, by definition, the
- barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded; only that one barrier is lost or potentially lost. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if Primary Containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Primary Containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, FA1 .1 and FU1 .1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Primary Containment barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category A, then B, E.
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FitzPatrick Annex Exelon Nuclear Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Table F-1 Fission Product Barrier Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss A. RPV Level [] Primary Containment D 1. RPV level c.annotbe ]1. RPV level cannot be restored[filid None D None D 1. PrimaryContainmentFloodinglsrequired Floodingisrequired restored and maintained> 0 in.(TAF) or cannot D maintained > diJn. be determined D (TAF) or cannot be determined D B.PCPressure 2. PC pressure> 2.7psig due 1. PCpressurerisefoUowedbyarap!d 2. PC pressure > 56 psig and rising
/Temperature to RCS leakage unexplaineddropinPCpressure D D D 3. Denag1ationconcentrationsexistinsidePC Nane Nooo None 2. PC pressure response not consistent with (H 2 > 6%AND 0 2 > 5%)
LOCAconditions D D _ 4. Torus temperature and RPV pressure cannot be maintained below the HCTL(EOP-11) C.lsolation 3. Releasepalhwayexistsoutside 1. RCS leakage > 50 gpm 3. Failure of any valve in any one line lo clos~ primaryco11tai11me11tresu!ting insidethedryweU AND D fromisolationfailureinany611the D Direct downsbeam pathway lo the envi1onmenl following(excludingnormal existsafterPCisolalionsignal, processsystemnowpalhsfroman 2. Unisolableprimarysystem unisolab!esystem): discharge outside primary
- Mainsteamfine D containment as indicated by 4. Intentional PC venting per EOPs None None - HPCI steam tine None Secondary Containment
- RC[Csteamline area radiation or D
- RWCU temperature above 5. Unisolableprimarysystemdischargeoutside
- Feedwater any Maximum Normal primarycontainmentasindicatedby Operating Limit (EOP[E Secondary Containment area radiation or
- 4. Emergency RPV Dep1essurizalion Tables) temperature above any Maximum Safe is required Operating Limit (EOP-5 Tables)
- 2. DryweUradiation D 5. Drywen radiation 5. Drywell radiation > 250,DOD R/hr
>3000 R/hr >30DR/hr D.Rad D 3. Primaryeoolanlactivity Nooo D None None D
>300µCi/gm 1-131do&e equivalent
~
- 4. Any condition in the opinion 2. Any condition in the 6. Anyconditioninthe 3. Any condition in the 6. Arly condition intheopinionoflheEmergency 6. Anycondition!ntheoplnionoftheEmergency cflheEmergencyDirectcr opinionoflhe opinionoftheEmergency opinionoftheEmergency Direc!orlhatindicateslossofthePrimary Directorlhatindicatespotenlial!ossoflhe E.Judgment that indicates Emergency Director Directorthalindicalesloss Director that indicates Containment barrier Primary Containment barrier D loss cftheFuelCladliarrier that indicatesO of the RCS barrier D potential Joss of the RcB D potenliallossofthe barrier Fuel Clad barrier Apri 2018 Page 246 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: A. RPV Level DegradationThreat: Loss Threshold:
- 1. Primary Containment Flooding is required due to any of the following:
D RPV water level cannot be restored and maintained > -19 in. (MSCRWL) and no core spray subsystem flow can be restored and maintained~ 4725 gpm D RPV water level cannot be restored and maintained~ ~44.5 in. D RPV water level cannot be determined and core damage is occurring NEI 99-01 Basis: The "Loss* threshold value corresponds to the level used in EOPs to indicate challenge of core cooling. This is the minimum value to assure core cooling without further degradation of the clad .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases EOP-2, EOP-3 and EOP-7 specify the requirement for Primary Containment Flooding and entry to the SAOGs when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Primary Containment Flooding is required when (ref. 1): D RPV water level cannot be restored and maintained above -19 in. (MSCRWL) and no core spray subsystem flow can be restored and maintained equal to or greater than 4725 gpm (design core spray flow) (ref. 2, 4) D RPV water level cannot be restored and maintained at or above -44.5 in. (elevation of the jet pump suction) (ref. 2) D RPV water level cannot be determined and core damage is occurring (ref. 3) The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad. This threshold is also a Potential Loss of the Primary Containment barrier (PC P-Loss A.1 ). Since Primary Containment Flooding occurs after core uncovery has occurred a Loss of the RCS barrier exists (RCS Loss A.1 ). Primary Containment Flooding, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. EOP-7 RPV Flooding
- 4. EOP-3 Failure to Scram
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: A. RPV Level Degradation Threat: Potential Loss Threshold: 1.RPV level cannot be restored and maintained > 0 in. (TAF) or cannot be determined NEI 99-01 Basis: This threshold is the same as the RCS barrier Loss threshold A.1 and corresponds to the site specific water level at the top of the active fuel. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of RCS barrier that appropriately escalates the emergency classification level to a Site Area
- Emergency .
JAFNPP Basis: An RPV level instrument reading of O in. indicates RPV level is at the top of active fuel (TAF) (ref. 1, 2, 3). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be restored and maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. Since core uncovery begins if RPV level drops below TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases When RPV level cannot be determined, EOPs require entry to EOP-7, RPV Flooding. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 1, 2, 4). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-7 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressures (in scram-failure events) (ref. 4). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. Note that EOP-3, Failure to Scram, may require intentionally lowering RPV water level to TAF (0 inches) and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL)(-19 inches) and TAF (ref. 5). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - RPS Failure EALs.
- JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. TSG-1 Parameter Assessment
- 4. EOP-7 RPV Flooding 5 . EOP-3 Failure to Scram
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: B. PC Pressure I Temperature Degradation Threat: Loss Threshold: None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold: I None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss/ Fuel Clad Potential Loss Matrix and Bases Category: C. Isolation Degradation Threat: Loss Threshold: None Apri 2018 Page 253 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: C. Isolation Degradation Threat: Potential Loss Threshold: None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss/ Fuel Clad Potential Loss Matrix and Bases Category: D. Rad Degradation Threat: Loss Threshold:
- 2. Drywell radiation > 3000 R/hr NEI 99-01 Basis:
3,000 R/hr is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the drywell. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage.
- This value is higher than that specified for RCS barrier Loss threshold D.S.
- Thus, this threshold indicates a loss of both Fuel Clad barrier and RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
There is no Potential Loss threshold associated with this item. JAFNPP Basis: The Containment High-Range Radiation Monitor (1?RE-104A or B) reading of 3000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage. The 3000 R/hr value was conservatively selected from EAP-4.1 Attachment 4 based on Case #4 (1 % noble gas release and 0.5% halogen release) one hour after shutdown. (ref. 1) JAFNPP Basis Reference(s):
- 1. EAP-4.1 Release Rate Determination
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- Barrier:
Attachment 4 - Fission Product Barrier Loss/ Fuel Clad Potential Loss Matrix and Bases Category: D. Rad Degradation Threat: Loss Threshold:
- 3. Primary coolant activity> 300 µCi/gm 1-131 dose equivalent NEI 99-01 Basis:
Assessment by the EAL Task Force indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.
- There is no Potential Loss threshold associated with this item.
Basis: No Additional JAFNPP Basis Reference(s): None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: D. Rad Degradation Threat: Potential Loss Threshold: None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss I Fuel Clad Potential Loss Matrix and Bases Category: E. Judgment Degradation Threat: Loss Threshold:
- 4. Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss I Potential Loss Matrix and Bases The Emergency Director judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences. D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term 'imminent' refers to recognition of the inability to reach safety acceptance criteria before completion of all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.
- D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
JAFNPP Basis Reference(s): None
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- , Barrier:
Attachment 4 - Fission Product Barrier Loss / Fuel Clad Potential Loss Matrix and Bases Category: E. Judgment Degradation Threat: Potential Loss
- Threshold:
- 2. Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost.
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases The Emergency Director judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences. D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety acceptance criteria before completion of all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. JAFNPP Basis Reference(s): None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: A. RPV Level Degradation Threat: Loss Threshold:
- 1. RPV level cannot be restored and maintained> O in. (TAF) or cannot be determined NEI 99-01 Basis:
The Loss threshold for RPV water level corresponds to the level that is used in EOPs to indicate challenge of core cooling. This threshold is the same as Fuel Clad Barrier Potential Loss threshold A.1 and corresponds to a challenge to core cooling. Thus, this threshold indicates a Loss
- of RCS barrier and Potential Loss of Fuel Clad barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
There is no Potential Loss threshold associated with this item .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases An RPV level instrument reading of O in. indicates RPV level is at the top of active fuel (TAF) (ref. 1, 2, 3). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers, and initiation of all ECCS. If RPV level cannot be restored and maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA.
- By definition, a LOCA event is a Loss of the RCS barrier.
When RPV level cannot be determined, EOPs require entry to EOP-7, RPV Flooding (ref. 4). The instructions in EOP-7 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss C.4). The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A.1 ). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification.
- Note that EOP-3, Failure to Scram, may require intentionally lowering RPV water level to TAF (0 inches) and control level between the Minimum Steam Cooling RPV Water Level (MSCRWL)(-19 inches) and TAF (ref. 5). Under these conditions, a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - RPS Failure EALs.
JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. TSG-1 Parameter Assessment
- 4. EOP-7 RPV Flooding
- 5. EOP-3 Failure to Scram
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- Attachment 4 - Fission Product Barrier Loss I Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Potential Loss Threshold:
None
- Apri 2018 Page 264 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: B. PC Pressure/ Temperature Degradation Threat: Loss Threshold:
- 2. PC pressure> 2.7 psig due to RCS leakage NEI 99-01 Basis:
The threshold pressure value is the drywell high pressure scram setpoint and is indicative of a LOCA event that requires ECCS response. There is no Potential Loss threshold associated with this item. JAFNPP Basis:
- The drywell high pressure scram setpoint is an entry condition to EOP-2, RPV Control, and EOP-4, Primary Containment Control (ref. 1, 2, 3). Normal primary containment pressure control functions (e.g., operation of drywell coolers, vent through SBGT, etc.)
are specified in EOP-4 in advance of less desirable but more effective functions (e.g., operation of drywell or torus sprays, etc.). In the JAFNPP design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the Primary Containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (ref. 4). The threshold phrase *... due to RCS leakage' focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. PC pressure greater than 2.7 psig with corollary indications (e.g., drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 2.7 psig should not be considered an RCS barrier Loss .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. EOP-4 Primary Containment Control, Entry Conditions
- 4. FSAR Update Chapter 6 Emergency Core Cooling Systems
- Apri 2018 Page 266 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: B. PC Pressure/ Temperature Degradation Threat: Potential Loss Threshold:
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:
- 3. Release pathway exists outside primary containment resulting from isolation failure in any of the following (excluding normal process system flowpaths from an unisolable system):
- Main steam line
- HPCI steam line RCIC steam line
- RWCU
- Feedwater NEI 99-01 Basis:
An unisolable RCS break outside Primary Containment is a breach of the RCS barrier. Thus, this threshold is included for consistency with the Alert emergency classification level. Large high-energy line breaks such as HPCI, Feedwater, RWCU, or RCIC that are unisolable represent a significant loss of the RCS barrier and should be considered as MSL breaks for purposes of classification .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases The conditions of this threshold include required primary containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside primary containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represents the loss of both the RCS and Primary Containment (see PC Loss C.3) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Even though RWCU and Feedwater systems do. not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems
- directly connected to the RCS .
JAFNPP Basis Reference(s):
- 1. FM-29A Main Steam System Flow Diagram
- 2. FM-298 Main Steam System Flow Diagram
- 3. FM-25A High Pressure Coolant Injection System Flow Diagram
- 4. FM-22A Reactor Core Isolation Cooling System Flow Diagram
- 5. FM-34A Feedwater System Flow Diagram
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:
- 4. Emergency RPV Depressurization is required NEI 99-01 Basis:
Plant symptoms requiring Emergency RPV Depressurization per the site specific EOPs are indicative of a loss of the RCS barrier. If Emergency RPV depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a loss of the RCS should be considered to exist due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.
- JAFNPP Basis:
Plant symptoms requiring Emergency RPV Depressurization are specified in the EOPs (ref. 2 through 7). JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. EOP-3 Failure to Scram
- 4. EOP-4 Primary Containment Control
- 5. EOP-5 Secondary Containment Control
- 6. EOP-6 Radioactivity Release Control
- 7. EOP-7 RPV Flooding
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: C. Isolation Degradation Threat: Potential Loss Threshold: I 1. RCS leakage > 50 gpm inside the drywell I NEI 99-01 Basis: This threshold is based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak, however, break propagation leading to significantly larger loss of inventory is possible .
- If primary system leak rate information is unavailable, other indicators of RCS leakage should be used.
JAFNPP Basis: If primary system leak rate information is unavailable, other indicators of RCS leakage should be used (ref. 3-5). Inventory loss events, such as a stuck open SRV, should not be considered when referring to 'RCS leakage' because they are not indications of a break, which could propagate. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 4.10.3
- 2. Technical Specification 3.4.4
- 3. ARP 09-5-1-34 OW Press Alarm Hi or Lo
- 4. AOP-9 Loss of Primary Containment
- 5. AOP-39 Loss of Coolant Apri 2018 Page 271 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Category: C. Isolation Degradation Threat: Potential Loss Threshold:
- 2. Unisolable primary system discharge outside primary containment as indicated by Secondary Containment area radiation or temperature above any Maximum Normal Operating Limit (EOP-5 Tables)
NEI 99-01 Basis: Potential loss of RCS based on primary system leakage outside the Primary Containment is determined from temperature or area radiation Max Normal Operating Limits in the areas of the main steam line tunnel, main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside Primary Containment. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage warrant an Alert classification. An unisolable leak which is indicated by a high alarm setpoint escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold C.5 (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Normal Operating Limit values define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-5, Secondary Containment Control Tables (ref. 1, 2) (see below). In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area
- radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.
- Apri 2018 Page 273 of 309 EP-AA-1014 Addendum 3 .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Table - EOP-5, Reactor Building Area Temperatures REACTOR BUILDING AREA TEMPERATURES MAXIMUM MAXIMUM MAXIMUM MAXIMUM AREA INSTRUMENT AREA INSTRUMENT NORMAL SAFE NORMAL SAFE Reactor Building Reactor Building 272 Ji 369 Ji elention ele\*ation southeast 104"F 112"F 104"F 153"F 66RTD-!o6 6611-106. Panel 09-75 23RTD-02C 2:\-204A. P,Ulel 09-95 66RTD-!08 6611-108. Panel 09-75 23RTD-02D 23"204B. P.inel 09-96 Outside 'A' LPCJ HPCI Drywell Entrance Bmery Enclosure J04"F ll3"F L3RTD-102C I 3-202C. Parnel 09-95 120"F 25l"F 66RTD-!J5 EPIC Only 13RTD-!02D l:\-202D. P,rnel 09-96 Below Refuel Floor RCIC Drywell Entr.tnce Exhaust 104'F l 13"F 13RTD-102A l 3-202A. Panel 09-95 12ll'F 218'F 66RTD-!05 6611-105. l"anel 09-75 13RTD-107II 13-2078, Panel 09-96 Outside 'B' LPCI Reactor Building 272 ft Battery Enclosure J04'F 115°F ete,*ation soutwest
!04'F !96'F 66RTD-!!6 EPIC Only 23RTD-O!C 23-202.\., Panel 09-95 23RTD-O!D 23-202B. P.inel 09-96 SLC PumpArea 66RTD-l 14 EPIC Only 0 10,l'F 133'F 'A' RHRHeat Exchanger Room 130'F 2-il'F Fuel Pool Cooling 23RTD-01A 23-201A,Panel 09-95 Pump Room !04'F 133'F 23RTD-O!B 23-20 !B, Panel 09-96 66RTD-l 13 EPIC Only Torus Room - South Reactor Building 300 ft HPCI Steamline
!20'F 2SO'F 13RTD-!07C 13-207C. P.rnel 09-95 elevation northeast 0 10.J'F 158°F 13RTD-!07D l 3-207D. P.rnel 09-96 66RTD-l 12 EPIC Only R\'i'CU Heat E.,,;changer Torus Room - Southwest Room RCIC Steamline J20'F 280'F ll5'F 203'F 13RTD-107A 13-207A.Panel 09-95
!2T[~l 17E P,mel 09-21
!2T[~l 17F Panel 09-21 13RTD-!02Il 13-202B, Panel 09-96 1
B' R\li:'CU Pump Room East Crescent 10.J'F 13TF
!ZTE-l !7C Panel 09-2 I 135'F 225'F 66RTD-!09ll 6611-!09B, Panel 09-75 l2TE-!J7D Panel 09-21 HPCI Room
'A' RWCU Pump Room 23RTD-94A 23-29-lA. Panel 09-95 12TE-l 17A Panel 01}.2! 125'F 225'F 23RTD-94B 23-294B, P,1nel 09-96 !04'F 13TF 12TE-l 17B P,mel 01}.2 l 23RTD-l 17A 23"217A.Panel U().95 23RTD-1 l 7 !J 23-2178, P,rnel 09-96 Reactor Building 300 n elevation southwest 0 !04'F 173'F RCICRoom 66RTD-l ll EPIC Only 13RTD-89A 13-289A. Panel 09-95 10.J'F 13TF 13RTD-89B 13-28911, P,rncl 09-96
'B' RHR Heat Exchanger Room \'?e~t Crescent 130'F 242'F 23RTD.02A 23-203A,Panel 09-95 13RTD-76A 13-276A.Panel 09-95 10.J'F 13TF 23RTD.0211 23-203B. P.rnel 09-96 l3RTD-7(,B 13-27611. P.inel 09-96
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- Attachment 4 - Fission Product Barrier Loss/
Potential Loss Matrix and Bases Table - EOP-5, Reactor Building Area Radiation Levels REACTOR BUILDaNG AREA RADIATION LEVELS AREA MAX!MU,:...*1 MAXIMUM INSTRUMENT NOR.\1AL SAFE Spmt J;ue, J'DDI !8RLH:t:iJ-l2: 25 mr/hr J o' rri:r/ll, Rl?'2m1r B!i!lldlJl;g
!BRLH:*'il-13 20 mr.thr 10' lll:i/hli 3-*U It el:-!!i*2Ua!!1 Ne\;' Fud \';;;111L !SRL',.-051-1,t 20 cur/hr 10' rru/h, Cle-JJ111.!i.P Pr~-coat Area 681UA-{'.*51-J5 80 mr:1.u 10' m:rmir R',X'CU El eat
!8RLH*5 l-l 6 5Dmr/hr 10~* m:r/hr E.'Kh3fil£12lf Room l*!i!cl l~JrJ~ J',.J.."ll[l' Rcwm !8RL',,<::*5J-J7 301] Itlf/hf 10' !J'i:f/hf c.;:,r.t;,;m:.ll!illcd Eqjllipmen.1
!8RIA-{:,;l-18 50 m[..'hf JO' rru/hr SLOr.JSf!
RWCU Pump Are:! Il8RL',.-{)51-19 30 i:IU:/hf' 10** m:r,,'111: Rx Bldg s:::.mple Are-.;. !8R[A-{'J51-20 30 mr:ihr 1o' mr/11.- 1mnc: ltl?'Jt l:H:lliA-'l:*; 1-21 ; ffif/hi" Jo** m:r.m, E.,Cllillilf;~1i Are.:.! Rc.:2cta:.- Br,JHl,:liill:gACC<l!!i'f-18RL'a.-'l:151-23 ,to ,111:,,-nr Jo' rr.s,111r 272 J1 Ci'.e,*.i:UrJ:!1 TIP OutJJct,e !81UAJJ'il-2ci 125 ffif/l:lr 1n' rri:r/11, El.-sc I lCll 1'.Ji,i::1 18RU-{)51-25 30mr/hr 10-' rru,,'hC
\Ve!>1 EIC1J Are-.! 18RU-<:i5 l-26 3:> mr1hr 10' m:r/h, E::5t cri;:;ceot !8RL\.-{)5 l-27 lHI mt/hr 1a' m:r,,*111r CRD R.cm::,,*a! El:J.1Cll 18R[.',.-{}5 l-28 25 mr.lhr 10' ms,,'11!
Vi'csl Cre-<=...c.c!ilt ISRLH); 1-29 lOOmr/hr 10' _flti,,'llr Rcilri!I FJoorW,1!!-.1 !8RU1.-'l:*;i2-}0 10' ffi!i/h! 2xHI' mr/hir
- Apri 2018 Page 275 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-5 Secondary Containment Control
- Apri 2018 Page 276 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Barrier: Reactor Coolant System Category: D. Rad Degradation Threat: Loss Threshold: Is. Drywell radiation > 300 R/hr NEI 99-01 Basis: 300 R/hr is a value which indicates the release of reactor coolant to the primary containment. This reading is less than that specified for Fuel Clad barrier Loss threshold D.2. Thus, this threshold would be indicative of a RCS leak only. If the radiation monitor reading increased to that value specified by Fuel Clad Barrier threshold, fuel damage would also be indicated. There is no Potential Loss threshold associated with this item. JAFNPP Basis: The Containment High-Range Radiation Monitors are 17RE-104 A or B. The 300 R/hr value was conservatively selected from EAP-4.1 Attachment 4 based on Case #5 (1/10 of 1% noble gas release) one hour after shutdown. (ref. 1) It is important to recognize that the radiation monitor may be sensitive to shine from the RPV or RCS piping. JAFNPP Basis Reference(s):
- 1. EAP-4.1 Release Rate Determination
- Apri 2018 Page 277 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: D. Rad Degradation Threat: Potential Loss Threshold:
- Apri 2018 Page 278 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Reactor Coolant System Category: E. Judgment Degradation Threat: Loss Threshold:
- 6. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost. Basis:
- The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations. JAFNPP Basis Reference(s): None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss I Potential Loss Matrix and Bases Reactor Coolant System Category: E. Judgment Degradation Threat: Potential Loss Threshold:
- 3. Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost. Basis:
- The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.
D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to the inability to reach final safety acceptance criteria before completing all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations. JAFNPP Basis Reference(s):
- None Apri 2018 Page 280 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: A. RPV Level Degradation Threat: Loss Threshold: I None
- Apri 2018 Page 281 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: A. RPV Level Degradation Threat: Potential Loss Threshold:
- 1. Primary Containment Flooding is required NEI 99-01 Basis:
The potential loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be established and maintained and that core melt is possible. Entry into Primary Containment Flooding procedures is a logical escalation in response to the inability to maintain adequate core cooling. Severe Accident Operating Guidelines (SAOGs) direct the operators to perform Containment Flooding when Reactor Vessel Level cannot be restored and maintained greater than a specified value or RPV level cannot be determined with indication that core damage is occurring. The condition in this potential loss threshold represents a potential core melt sequence which, if not corrected, could lead to vessel failure and increased potential for containment failure. In conjunction with Reactor Vessel water level "Loss* thresholds in the Fuel Clad and RCS barrier columns, this threshold will result in the declaration of a General Emergency -- loss of two barriers and the potential loss of a third .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAFNPP Basis: EOP-2, EOP-3 and EOP-7 specify the requirement for Primary Containment Flooding and entry to the SAOGs when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Primary Containment Flooding is required when (ref. 1): D RPV water level cannot be restored and maintained above -19 in. (MSCRWL) and no core spray subsystem flow can be restored and maintained equal to or greater than 4725 gpm (design core spray flow) (ref. 2, 4) D RPV water level cannot be restored and maintained at or above -44.5 in. (elevation of the jet pump suction) (ref. 2) D RPV water level cannot be determined and core damage is occurring (ref. 3) The above EOP conditions, if not restored and maintained, represent a potential core melt sequence which could lead to RPV failure and increased potential for Primary Containment failure. This threshold is also a Loss of the Fuel Clad barrier (FC Loss A.1 ). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss A.1 ). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-2 RPV Control
- 3. EOP-7 RPV Flooding
- 4. EOP-3 Failure to Scram
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- Barrier:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: B. PC Pressure/ Temperature Degradation Threat: Loss Threshold:
- 1. PC pressure rise followed by a rapid unexplained drop in PC pressure NEI 99-01 Basis:
Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase from a high energy line break indicates a loss of containment integrity. This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition . JAFNPP Basis: No additional JAFNPP Basis Reference(s): None
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- Barrier:
Attachment 4 - Fission Product Barrier Loss/ Potential Loss Matrix and Bases Primary Containment Category: B. PC Pressure/ Temperature Degradation Threat: Loss Threshold:
- 2. PC pressure response not consistent with LOCA conditions NEI 99-01 Basis:
Primary Containment pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, Primary Containment pressure not increasing under these conditions indicates a loss of containment integrity. This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition. JAFNPP Basis: The calculated pressure and temperature responses of the Primary Containment are shown in Figures 14.6-6 through 14.6-8. Figure 14.6-8 shows that the maximum calculated drywell pressure is 45 psig, which is well below the design allowable pressure of 56 psig (ref. 1, 2, 3). After the initial blowdown of the primary coolant from the RPV into the drywell (which occurs in the first 30 seconds), the temperature of the suppression chamber water approaches 130°F (from an initial 90°F) (Figure 14.6-7). The primary containment pressure stabilizes at about 26 psig, as shown on Figure 14.6-8 (ref. 2). Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate. JAFNPP Basis Reference(s):
- 1. FSAR Update Figure 14.6-6
- 2. FSAR Update Figure,14.6-8
- 3. FSAR Update Section 14.6.1.3.3 Apri 2018 Page 285 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAMES A. FITZPATRICK FSARUPDATE TRANSi ENT RES UL TS FROM LOCA - DRYWELL TEMPERATURE- INITIAL CORE 250 REV.O JULY, 1982 I FIGURE NO. 14.6-6 d 200 0 G:' l
- UJ a::
F?* 150
~
UJ a.
- I:
w
.J ..J w ~
a: Q 100 so a= FULL RHR CAPACITY 4 RHR PUMPS, 2HX, ICS W/ CONT. SPRAY b = 1 LOOP, 2 RHR PUMPS, 1 HX, 1CS W/CONT.SPRAY C = 1 LOOP, 1 RHR PUMP, 1 HX, 1CS W/ CONT. SPRAY d =l LOOP, I RHR PUMP, lHX,lCS W/o CONT. SPRAY 105
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAMES A. FITZPATRICK FSAR UPDATE TRANSIENT RESULTS FROM LOCA - PRESSURE SUPPRESSION POOL TEMPERATURE - [NITIAL CORE c,d 200 REV. 0 JULY,1982 I FIGURE NO. 14.6-7 150 C 02
*UJ ci: =>
I-
~
w 11.. i5 I-
..J 0
- r 100 5
~
w a:: 11.. 11.. a = FULL RHR CAPACITY 4 RHR PUMPS, 2 HX, 1 CS, W/ CONi. SPRAY
=>
"' b :; 1 LOOP, 2 RHR PUMPS, I HX, 1 CS W/ CONT. SPRAY c = l LOOP, l RHR PUMP, 1 HX, l CS W/ CONT. SPRAY d =1 LOOP, 1 RHR PUMP, 1 HX, l CS W/o CONT. SPRAY 50 0'--_ _ __.__ _ ____.__ _ ___._ _ _ _....___ _ _~ - - - - - - ' ' - - - - - - - '
-1 O I 2 3 4 5 6 10 10 10 10 10 10 10 10 TIME Apri 2018 Page 287 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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- Attachment 4 - Fission Product Barrier Loss/
Potential Loss Matrix and Bases a= FULL RHR CAPACITY 4 RHR PUMPS, 2 HX, l CS W/CONT.SPRAY b =1 LOOP, 2 RHR, l HX, l CS W/ CONT. SPRAY c:: 1 LOOP, l RHR, l HX, 1 CS W/CONT.SPRAY d = 1 LOOP, l RHR, l HX, l CS W/o CONT. Sli'RAY JAMES A. FITZPATRICK FSARUPDATE 40 TRANSi ENT RESULTS FROM LOCA - CONTAINMENT PRESSURE - INITIAL CORE REV. 0 JULY, 1982 I FIGURE NO. 14.6-8 QI 30 a. l UJ a::* UJ a:: II.. t-z UJ
- e z
~ 20
!u d
10 0 ------"-::------L----....l------1----...L-----.I.-----...I
-1 0 1 2 3 5 10 10 10 10 10 10 TIME (~sec)
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: B. PC Pressure I Temperature Degradation Threat: Potential Loss Threshold: 1 2. PC pressure > 56 psig and rising NEI 99-01 Basis: The value of 56 psig for Potential Loss of containment is based on the Primary Containment design pressure. JAFNPP Basis: When the primary containment exceeds the maximum allowable value (56 psig) (ref. 1), primary containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The drywell and suppression chamber maximum allowable value of 56 psig is based on the primary containment design pressure as identified in the JAFNPP accident analysis (ref. 1, 3). If this threshold is exceeded, a challenge to the Primary Containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Primary Containment barrier even if a Primary Containment breach has not occurred. JAFNPP Basis Reference(s):
- 1. FSAR Update Section 5.2.3
- 2. EOP-4 Primary Containment Control
- 3. UFSAR 14.6.1.3.3
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: 8. PC Pressure/ Temperature Degradation Threat: Potential Loss Threshold:
- 3. Deflagration concentrations exist inside PC (H 2 ~ 6% AND 02 ~ 5%)
NEI 99-01 Basis: BWRs specifically define the limits associated with explosive (deflagration) mixtures in terms of deflagration concentrations of hydrogen and oxygen. For Mk 1/11 containments the deflagration limits are '6% hydrogen and 5% oxygen in the drywell or suppression chamber". JAFNPP Basis:
- Deflagration (explosive) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal- water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1) .
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- Attachment 4 - Fission* Product Barrier Loss I Potential Loss Matrix and Bases Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inertion.
The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 2) and readily recognizable because 6% hydrogen is well above the EOP-4, Primary Containment Control, entry condition (ref. 3). The minimum global deflagration hydrogen/oxygen concentrations (6% and 5%, respectively) require intentional primary containment venting, which is defined to be a Loss of Primary Containment (PC Loss C.4). The Drywell H2'02 Analyzer System samples the atmosphere in the primary containment to detect concentrations of hydrogen and oxygen. This system consists of two redundant analyzers located in the Reactor Building with indication displayed on panels 27PCX-101A and B in the Relay Room. Each monitor is capable of detecting hydrogen with two ranges in concentrations from O to 10% and O to 30%, and oxygen in concentrations from O to 30%. Control Room indication is supplied to EPIC display CAS1. Annunciation is provided on CRP 9-5 (ref. 4). JAFNPP Basis Reference(s):
- 1. BWROG EPG/SAG Revision 2, Sections PC/G
- 2. EOP-4a Primary Containment Gas Control
- 3. EOP-4 Primary Containment Control
- 4. FSAR section 5.2.3.14
- 5. FSAR Table 7.3-6
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: B. PC Pressure / Temperature Degradation Threat: Potential Loss Threshold:
- 4. Torus temperature and RPV pressure cannot be maintained below the HCTL (EOP-11)
NEI 99-01 Basis: The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise: D Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR D Suppression chamber pressure above the Primary Containment Pressure Limit, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCTL is a function of RPV pressure and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. JAFNPP Basis: his threshold is met when the final step of section Torus Temperature in EOP-4, Primary Containment Control is reached (ref. 2, 3). In addition to the Torus temperature and pressure limits, Torus water level must be within HCTL limits or HCTL is exceeded. JAFNPP Basis Reference(s):
- 1. BWROG EPG/SAG Revision 2, Section 17
- 2. EOP-4 Primary Containment Control
- 3. EOP-11 EOP and SAOG Graphs Apri 2018 Page 292 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold:
- 3. Failure of any valve in any one line to close AND Direct downstream pathway to the environment exists after PC isolation signal NEI 99-01 Basis:
These thresholds address incomplete containment isolation that allows direct release to the environment. The use of the modifier 'direct* in defining the release path discriminates against release paths through interfacing liquid systems (e.g. Reactor Building Closed Loop Cooling). The existence of an in- line charcoal filter does not make a release path indirect since the filter is not effective at removing fission product noble gases. Typical filters have an efficiency of 95-99% removal of iodine. . Given the magnitude of the core inventory of iodine, significant releases could still occur. In addition, since the fission product release would be driven by boiling in the reactor vessel, the high humidity in the release stream can be expected to render the filters ineffective in a short period .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAFNPP Basis: This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic Primary Containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of primary containment integrity. For the purposes of this EAL threshold, "environment' is any location that will ultimately provide a pathway to outside without going through an interfacing liquid system - e.g. a release from primary containment to the reactor building through an open penetration meets this definition based on the release then being exhausted to the outside via the Standby Gas
.Treatment System, Stack, Reactor Building Vent or other unmonitored pathways.
As stated above, the adjective 'Direct' modifies 'downstream pathway" to discriminate against release paths through interfacing liquid systems. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable Main steam line, HPCI steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisloable Primary Containment atmosphere vent paths. If the main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using the Category R, Abnormal Rad Release/ Rad Effluent, EALs. The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If operator actions from the Control Room are successful, this threshold is not applicable. Credit is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach. EOP-4, Primary Containment Control, Section Primary Containment Pressure may specify primary containment venting and intentional bypassing of the Primary Containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid Primary Containment isolation signal, the Primary Containment barrier should be considered lost. JAFNPP Basis Reference(s):
- 1. EOP-4 Primary Containment Control Apri 2018 Page 294 of 309 EP-AA-1014 Addendum 3 (Revision 0)
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold:
- 4. Intentional PC venting per EOPs NEI 99-01 Basis:
The EOPs may direct containment isolation valve logic(s) to be intentionally bypassed, regardless of radioactivity release rates. Under these conditions with a valid containment isolation signal, the containment should also be considered lost if containment venting is actually performed. Intentional venting of Primary Containment for Primary Containment pressure or combustible gas control per EOPs to the secondary containment and/or the
- environment is considered a loss of containment. Containment venting for pressure when not in an accident situation should not be considered.
JAFNPP Basis: EOP-4, Primary Containment Control, Section Primary Containment Pressure (the final step using EP-6) may specify primary containment venting and intentional bypassing of the Primary Containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). The threshold is met when the operator begins venting the primary containment in accordance with EOP-4 and EP-6, not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in the first step of Section Primary Containment Pressure to control drywell pressure below the drywell high pressure scram setpoint does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below Technical Specification LCO limits .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases EOP-4a, Primary Containment Gas Control, (using EP-6) may specify primary containment venting and purging, and intentional bypassing of the Primary Containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 3). The threshold is met when the operator begins venting the primary containment in accordance with EOP-4a and EP-6, not when actions are taken to bypass interlocks prior to opening the vent valves. JAFNPP Basis Reference(s):
- 1. EOP-4 Primary Containment Control
- 2. EP-6 Post Accident Containment Venting and Gas Control
- 3. EOP-4a Primary Containment Gas Control
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: C. Isolation Degradation Threat: Loss Threshold:
- 5. Unisolable primary system discharge outside primary containment as indfcated by Secondary Containment area radiation or temperature above any Maximum Safe Operating Limit (EOP-5 T..,.hl--\
- - --'J NEI 99-01 Basis:
The presence of area radiation levels or area temperatures above any Maximum Safe Operating Limit indicates unisolable primary system leakage outside the primary containment are addressed after a containment isolation. The indicators should be confirmed to be caused by RCS leakage. There is no Potential Loss threshold associated with this item . JAFNPP Basis: The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The Maximum Safe Operating Limit values define this Primary Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-5, Secondary Containment Control Tables (ref. 1, 2) (see below). In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Table - EOP-5, Reactor Building Area Temperatures REACTOR BUILDING AREA TEMPERATURES MAXIMUM MAXIMUM MAXIMUM MAXIMUM AREA INSTRUMENT AREA INSTRUMENT SAFE NOR."1AL SAFE NORMAL Reactor Building Reactor Building 272 ft 369 fc ele,*ation ele,*;1tion southeast
!04.F 112°F 104.F 153*p 66RTD-l06 6611-106. l'Jncl 09-75 23RTD-D2C 2:\-204A, Panel 09-95 66RTD-108 6611-!0S. l'.mcl 09-75 23RTD-D2D 23-204B, !'.me! 09-96 Outside 'A' LPCI HPCI Drywell Entmnce Battery Enclosure JO.j*F 113°F I3RTD-l02C 13-202C. l'.mel 09-95 l2o*F 25l.F 66RTD-l 15 EPIC Only 13RTD-!OlD 13-202D, Rmcl 09-96 Below Refuel Floor RCIC Drywell Entr.tnce Exhaust 104.F l l3'F 13RTD-102A 13-202A, Panel 09-95 12o*F 21s*F 66RTD-l05 6611-105, Panel 09-75 13RTD-10711 13-207B, l',mcl 09-%
Outside 'B' LPCI Reactor Building 272 fr Battery Enclosure 104*p 113°F eleYation sounvcst 104-F 196.F 66RTD-l 16 EPIC Only 23RTD-DIC 23-202A, Panel 09-95 23RTD-OID 23-202B, !'.mcl 09-96 SLC PumpArea 66RTD-114 EPIC Only 0 JO.j*F 133*F 'A' RHRHeat Exchanger Room 13o*F 242.F Fuel Pool Cooling 23RTD-DIA 2:l-201A,Panel 09-95 Pump Room 104.F 133°F 23RTD-OIB 23-20 lB. Panel 119-96 66RTD-l 13 EPIC Only Torns Room - South Reactor Buildin]l 300 ft HPCI Steamline 120'F 280'F l3RTD-l07C l3"207C, !'.me! 09-95 elevation northeast 0 JO.j*p ISS'F 13RTD-l07D l:l-207D, l'ancl 09-96 66IlTD-ll2 EPIC Only Torus Room - Southwest RWCU Heat E,changer RCIC Ste,unline RcXJm 120-F 2SO'F ll'i*F 203'F 13RTD-107A 13-207A,Panel 09-95 12TE-ll7E Panel 09-2l I3RTD-l02II 13-202B. Rrnel 09-96 12TE-l 17F Panel 09-21
'B' RW'CU Pump Room East Crescent 104.F l37'F 12Tf.ll 7C Panel 09-21 I35'F 22'i'F 66RTD-l09B 6611-10911. l'anel 09-75 l2TE-ll7D Panel 09-21 HPCJ Room
'A' RWCU Pump Room 23RTD-94A 2;\-294A. Panel 09.95 l2TE-l l7A Panel 09-2 l 125'F 225'F 23RTD-94II 23-29,'8. Panel 09-96 104'F 137'F I2TE-l l7B Panel 09-2 l 2_,RTD-l 17A 23-217A, Panel 09-95 23RTD-l l7B Z3-2 l 7B. Panel 09-96 Reactor Building 300 ft elevation southv.*cst 0 104-F 173'F RCIC Room 66RTD-l ll EPIC Only 13RTD.S9A l 3"289A, Panel 09-95 104-F 137'1' l3RTD~~9B l 3-289B. Rmel 09-96
'll' rum Heat Exchanger Room \Vest Crescent 130-F 242'F 104'F 23RTD-D2A 23-203A, Panel 09-95 13RTD-76A I 3-276A. Panel 09-95 13TF 23RTD-D2II 23-203B, P.mel 09-96 13RTD-76B l3-276B, Panel 09-96
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Table - EOP-5, Reactor Building Area Radiation Levels REACTOR BUILDING AREA RADIATION LEVELS MAXlMW,:1 I\MXtMt..JM AREA. INSTRUl\*1ENT NOR,MAL SAFE Spcnl l;Ui:i l'DOI Il8RtA-<:15J.J:! 25 1:111:/hr 10' mr/ll!i Re2.cra:r OU!ld:.ri.;g, Il:8RLHi51-13 :wmom ]I} ' llli/h!:" n 3H (f,e,*;1Ua!i1 Nc*.v Fuel \'.!!Ill IlSIUA.,f:051*1 *1 20 mr:,,ihr 10** llli/h, (J(8HEp Pm-coat An::a !fiRLa\.~::.51.15 BO Ell(lihr 10 ' mi/11, RWCU E-IE:al lTJ.1/1:Jli HSRLHJ;il-16 ~.o mr/ihr ]I]*. Exdlam:;;eir ll(:om Fuel I\Jl]j ]';TJffiP, IWtJm U:!tRU-{)51-17 300 ffil!"/hr HI~ rn.r/hli
- cmrt:11J12111atc,1 15:J:U!pmrn:1 Storn:;,-c RWCU Pump Af(;J; Rx Bid£ S.1mplr Arr.;
iSRU-'1)51-18
!8RLH)5 l-l 9 HSRtA-051-20 50 mr/hr 30 mr:lhr 30 mr.ihr JO'lll:i/1:Jf 1o' llli/11, rn' m:r/h!i RUG.C Hcill ) I],. rn:r,,'1111" 1SRU-4i5l*2ft 5 mir/llr Ex.dl:I!lgE-r Are::
RC2ctor [lnJHd!n;gAcceiS 272 fl r1-!!*,;1Um1
!:!i:R[A-05 l *23 4o mr:&r ] o'* !TC/hii TIP Cul:!Jd:,e Il:!i:RU-{)5J-2,1 125 mr/1:!r ] I]' IJ1.::T/11C
( E.:l..-9: I !CU Alic:! nsm:.v.:i,1-25 30 mr/hr ] I]' [Il.:!"/1:I h""
\Vt:St EKU Are:;; !SRLa\.-{)51-26 35 mr/hr 10,. llli/11(
E:!st Crt::SCE:Dl H:!i:RU-<)51-27 110 m,/llf ] I],. rn:r/llf CRD R.cm::,,*an Hatch ISRLH)'i l-28 25 mr:lhr lO~ mr/hri
\1;.'E:~1 Cn:'"Set?flt USRU-051-29 LOO mi./hr ]I]~ Ol:i/hf 10 ~ Illi'/1:!li I
RUIN!! l':JoorW61 K:!i:RL"i...;)52*.:,0 2:r. HI Ilil!/hli
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases JAFNPP Basis Reference(s):
- 1. EP-1 EOP Entry and Use
- 2. EOP-5 Secondary Containment Control
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: D. Rad Degradation Threat: Loss Threshold:
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: D.Rad Degradation Threat: Potential Loss Threshold: Js. Drywell radiation> 250,000 R/hr NEI 99-01 Basis: 250,000 R/hr is a value that indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted . There is no Loss threshold associated with this item .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases The Containment High-Range Radiation Monitor (1?RE-104A or B) reading of 250,000 R/hr indicates the release of reactor coolant into the drywell with elevated activity indicative of fuel damage. The 250,000 R/hr value was conservatively selected from EAP-4.1 Attachment 4 based on Case #3 one hour after shutdown. This is based on the following: D 10% noble gas release and 10% halogen release for airborne D 5% halogen content and 0.1 % fission product content for liquid borne (ref. 1) A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure into the reactor coolant has occurred. Regardless of whether the primary containment barrier itself is challenged, this amount of activity in Primary Containment could have severe consequences if released. It is, therefore, prudent to treat this as a potential loss of the Primary
- Containment barrier and upgrade the emergency classification to a General Emergency.
In order to reach this Primary Containment barrier Potential Loss threshold, a loss of the RCS barrier (RCS Loss D.5) and a loss of the Fuel Clad barrier (FC Loss D.2) have already occurred. This threshold, therefore, represents at a General Emergency classification. JAFNPP Basis Reference(s):
- 1. EAP-4.1 Release Rate Determination
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Barrier: Primary Containment Category: E. Judgment Degradation Threat: Loss Threshold:
- 6. Any condition in the opinion of the Emergency Director that indicates loss of the Primary Containment barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.
- The Containment barrier should not be declared lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Containment barrier.
When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical Specifications .
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- JAFNPP Basis:
Attachment 4 - Fission Product Barrier Loss/ Potential Loss Matrix and Bases The Emergency Director judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences. D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent' refers to recognition of the inability to reach safety acceptance criteria before completion of all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results. D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be
- mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
JAFNPP Basis Reference(s): None
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Attachment 4 - Fission Product Barrier Loss / Potential Loss Matrix and Bases Primary Containment Category: E. Judgment Degradation Threat: Potential Loss Threshold:
- 6. Any condition in the opinion of the Emergency Director that indicates potential loss of the Primary Containment barrier NEI 99-01 Basis:
This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost. The Containment barrier should not be declared potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS) the Containment barrier status is addressed by Technical Specifications .
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- Attachment 4 - Fission Product Barrier Loss /
Potential Loss Matrix and Bases Basis: The Emergency Director judgment threshold addresses any other factors relevant to determining if the Primary Containment barrier is* potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring
- capability and dominant accident sequences.
D Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term 'imminent' refers to recognition of the inability to reach safety acceptance criteria before completion of all checks. D Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results .
- D Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
JAFNPP Basis Reference(s): None
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- 1.
Attachment 5 - ADDITIONAL GUIDANCE FOR CLASSIFICATION The 15-minute criterion commences when plant instrumentation, plant alarms, computer displays, or incoming verbal reports that correspond to an EAL first become available to any member of the ERO who by virtue of training and experience is qualified to assess. indications or reports against
. the EALs (EDs, SROs, ROs Key facility staff).
- 2. Validation or confirmation of plant indications, or reports to the plarit operators, is to be accomplished within the 15-minute period as part of the assessment.
- 3. The ;15-minute period encompasses all assessment, classification, and declaration actions associated with making an emergency declaration and ends when the Shift Manager or Emergency Director verbally declares the emergency classification level.
- 4. Emergency declarations shall be made promptly. The 15-minute criterion is not to be construed as a grace period in which an attempt to restore plant conditions to avoid declaring an EAL that has already been exceeded. This does not preclude actions to correct or mitigate an off-normal condition, but once an EAL has been exceeded, the emergency declaration shall be made promptly without waiting for the 15-minute period to elapse. This is particularly the case when the EAL threshold is exceeded based on occurrence of a condition, rather than the duration of a condition.
- 5. For' EAL thresholds that specify duration of the off-normal condition, the
- emergency declaration process runs concurrently with the specified threshold duration. Once the off-normal condition has existed for the duration specified in the EAL, no further effort on this declaration is necessary-the EAL has been exceeded. Example: The EAL 'fire which is not extinguished within 15 minutes of detection.' On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts. The event must be declared as soon as it is known the fire will not be suppressed within 15 minutes of the time CR becomes aware of fire; NOT 15 minutes after the initial 15 minutes has passed .
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FitzPatrick Annex Exelon Nuclear
- 6.
Attachment 5 - ADDITIONAL GUIDANCE FOR CLASSIFICATION For thresholds with specific durations, if it is known that no actions can prevent the time period being met, then the event should be declared immediately. Example: There is a 15 minute time duration with a loss of offsite power. A call from Power Control could provide information that the lines will NOT be restored within 15 minutes therefore the clock stops there. There is, under the circumstances, no need to wait 15 minutes to declare 7.A small number of EAL thresholds are related to the results of analyses (e.g., dose assessments, chemistry sampling, and/or inspections) that are necessary to ascertain whether a numerical EAL. threshold has been exceeded, rather than confirming or verifying an alarm or a received report. In most of these cases, the basis of the EAL will identify the analysis necessary and its scope. In these limited cases, the 15-minute declaration period starts with the availability of analysis results that show the threshold to be exceeded; this is the time that the information is first available .
- Apri 2018 Page 309 of 309 EP-AA-1014 Addendum 3 (Revision 0)}}