ML20079J095

Emergency Implementation Procedure 2-001 Revision 28

ML20079J095
Person / Time
Site:	River Bend
Issue date:	01/22/2020
From:	Entergy Operations
To:	Office of Nuclear Reactor Regulation
References
RBG-48012 EIP-2-001, Rev 028
Download: ML20079J095 (175)

Text

Enclosure RBG-48012 Emergency Implementation Procedure 2-001 R~vision 28

REFERENCE USE ENTERGY RIVER BEND STATION STATION OPERATING MANUAL

• EMERGENCY IMPLEMENTING PROCEDURE

• CLASSIFICATION OF EMERGENCIES PROCEDURE NUMBER: *EIP-2-001 REVISION NUMBER: *028 Effective Date: *1/22/2020 NOTE : SIGNATURES ARE ON FILE.

Tern Rev 2 AddCounter 1 Att Enc DS MSet REGULAR KWN OFF REFERENCE USE

• INDEXING INFORMATION I

REFERENCE USE

.TABLE OF CHANGES LETTER DESIGNATIO N DETAILED DESCRIPTION OF CHANGES TRACKING NUMBER EIP-2-001 REV -028 PAGE 1 OF 144

REFERENCE USE TABLE OF CONTENTS SECTION PAGE NO.

1 PURPOSE ................................................................................................................................. 3 2 REFERENCES ......................................................................................................................... 3 3 DEFINITIONS ......................................................................................................................... 4 4 RESPONSIBILITIES ............................................................................................................... 8 5 GENERAL ................................................................................................................................ 8 6 PROCEDURE ....................................................................................................................... :.11 7 DOCUMENTATION ............................................................................................................. 12 ATTACHMENT 1 - INITIATING CONDITION MATRIX ........................................................ 13 ATTACHMENT 2 - ABNORMAL RADIATION LEVELS / RADIOLOGICAL EFFLUENT ................................................................................................................... 21 ATTACHMENT 3 - FISSION PRODUCT BARRIER................................................................. 23 ATTACHMENT 4 - HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ........................................................................................................................ 25 ATTACHMENT 5 - SYSTEM MALFUNCTION ........................................................................ 28 ATTACHMENT 6- COLD SHUTDOWN /REFUELING .......................................................... 32 ATTACHMENT 7 - EVENTS RELATED TO ISFSI .................................................................. 35 ATTACHMENT 8-EAL BASES ................................................................................................ 36 ATTACHMENT 9 - EIP-2-001 CLASSIFICATION USER AID .............................................. 143 EIP-2-001 REV -028 *PAGE 2 OF 144

REFERENCE USE I PURPOSE 1.1 This procedure provides guidelines for properly classifying emergencies.

2 REFERENCES 2.1 River Bend Station (RBS) Emergency Plan 2.2 EIP-2-002, Classification Actions 2.3 NEI 99-01 Rev 5, Methodology for Development of Emergency Action Levels 2.4 NUREG-1022, Event Reporting Guidelines: IOCFR5b.72 and IOCFR50.73 1 2.5 NRC Bulletin 2005-02, Emergency Preparedness and Response Actions for Security-Based ~vents 2.6 NRC RIS 2003-18 Supp 2, Use ofNuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels 2.7 10 CPR 50 Appendix E IV.C.2, Emergency Declaration Timeliness EIP-2-001 REV-028 PAGE 3 OF 144

REFERENCE USE 3 DEFINITIONS 3.1 AFFECTING SAFE SHUTDOWN: Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN condition. Plant condition applicability is determined by Technical Specification LCOs in effect.

3 .1.1. Example 1: Event causes damage that results in entry into an LCO that requires the plant to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is not "AFFECTING SAFE SHUTDOWN."

3.1.2. Example 2: Event causes damage that results in entry into an LCO that requires the plant to be placed in COLD SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is "AFFECTING SAFE SHUTDOWN."

3.2 ALERT

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 PAG exposure levels.

3.3 BOMB: Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

3.4 CIVIL DISTURBANCE: A group of persons violently protesting station operations or activities at the site.

3.5 CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment. (ISFSI MPC Confinement Boundary) 3.6 CONTAINMENT CLOSURE: A containment condition where at least one integral barrier to the release of radioactive material is provided.

3.7 EXPLOSION

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.

EIP-2-001 REV-028 PAGE 4 OF 144

REFERENCE USE

3.8 EXTORTION

An attempt to cause an action at the station by threat of force.

3.9 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.

3.10 GENERAL EMERGENCY: 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 PAO exposure levels offsite for more than the immediate site area.

3.11 HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

3.12 HOSTILE ACTION: An act toward a Nuclear Power Plant 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 the NPP. Non-terrorism-based EALs should be used to address such activities, (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA.).

3.13 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.

3 .14 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.

3.15 INDEPENDENT SPENT FUEL STORAG E 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.

3.16 INTRUSION: A person(s) present in a specified area without authorization. Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE FORCE.

EIP-2-001 REV -028 PAGE 5 OF 144

REFERENCE USE 3.17 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 offnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONS.

3.18 NOTIFICATION OF UNUSUAL EVENT (NOUE): 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.

3.19 OWNER CONTROLLED AREA: The area within the EOI property boundary.

3.20 PROJECTILE: An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

3.21 PROTECTED AREA: Encompasses all controlled areas within the security protected area fence.

3 .22 SABOTAGE: Deliberate damage, mis-alignment, 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.

3.23 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.

EIP-2.,001 REV - 028 PAGE 6 OF 144

REFERENCE USE 3.24 SIGNIFICANT TRANSIENT: An UNPLANNED event involving one or more of the following:

3 .24.1. Automatic turbine run back >25% thermal reactor power, 3.24.2. Electrical load rejection >25% full electrical load, 3.24.3. Reactor Trip, 3.24.4. Safety Injection Activation or 3 .24.5. Thermal power oscillations > 10%.

3.25 SITE AREA EMERGENCY: Events are in progress or have occurred which involve 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 PAO exposure levels beyond the SITE BOUNDARY.

3.26 SITE BOUNDARY: For classification and dose projection purposes, the site boundary is the area defined as exclusion area or exclusion zone in IOCFRl00.3 (a) which is a boundary of approximately 3,000 feet (or 0.5748 mile) from the RBS reactor centerline.

3.27 STRIKE ACTION: A work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands made on Entergy or its affiliates. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONS.

3.28 UNISOLABLE: A breach or leak that cannot be promptly isolated.

3.29 UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

3.30 VALID: An indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, (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.

EIP-2-001 REV - 028 PAGE 7 OF 144

REFERENCE USE 3.31 VISIBLE DAMAGE: Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concere regarding the continued operability or reliability of the affected 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.

3.32 VITAL AREA: Any area, normally within the PROTECTED 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.

4 RESPONSIBILITIES 4.1 Operations Shift Manager (OSM) - It is the responsibility of the OSM to:

4.1.1. Recognize and properly classify emergency conditions, and 4.1.2. Assume the responsibilities of the Emergency Director (ED) until relieved by the designated Emergency Director.

4.2 Control Room Supervisor (CRS) - It is the responsibility of the CRS to assume the responsibility of the OSM if the OSM becomes incapacitated.

4.3 Designated Emergency Director - It is the responsibility of the designated Emergency Director to assist the OSM as requested and if the emergency is classified at ALERT or higher, relieve the OSM of the ED duties arid responsibilities as soon as practical.

5 GENERAL

5. I Anytime Emergency Operating Procedures (EOPs) or Abnormal Operating Procedures (AOPs) are initiated, this procedure should be reviewed to determine if an emergency action level has been reached.

5 .2 This procedure, with Attachment I through Attachment 8, is a guideline for classifying emergencies. In a situation not covered by the Emergency Action Levels, the OSM (Emergency Director) must use his best judgment in determining the appropriate emergency classification.

5.2.1. Attachment I is a matrix that is useful as a quick review to determine if an EAL INITIATING CONDITION is met.

EIP-2-001 REV-028 PAGE 8 OF 144

REFERENCE USE 5.2.2. The Emergency Action Levels and bases in Attachments 2 - 8 are consistent with the definitions and INITIATING CONDITIONs in the RBS Emergency Plan.

5.2.3. Attachment 9 is the user aid that presents the EALs in chart format.

5.3 For Emergency Action Levels based on plant instrumentation, the indication shall be a VALID indication. When all indications for a certain parameter have been lost, the Emergency Director should use his best judgment and other plant indications to classify the emergency (e.g., loss oflevel trend on all RPV level instrumentation).

5.3.1. EPP-2-503, RBS Equipment Important to Emergency Preparedness, lists the instruments used for EAL identification and provides guidance for compensqtory measures when an instrument is out of service.

5 .4 The assessment, classification, and declaration of an emergency condition is expected to be completed within 15 minutes after the availability of indications (i.e. plant instrumentation, plant alarms, computer displays, or incoming verbal reports) to plant operators that an EAL has been exceeded.

5 .4.1. The 15 minute criterion is not to be construed as a grace period to restore plant conditions to avoid declaring the event.

5.4.2. The emergency declaration should be made promptly without waiting for the 15 minute period to elapse once the EAL is recognized as being exceeded.

5.4.3. For EALs that specify duration of the off-normal condition, such as fire lasting 15 minutes, loss of power for 15 minutes, etc.:

1. The Emergency Director shall make the declaration at the first available opportunity when the time has elapsed (not after an additional 15 minutes).

2. The declaration should be made before the EAL is met (before the time duration has elapsed) when the Emergency Director has information that the off-normal condition will not be corrected within the specified time duration.

EIP-2-001 REV -028 PAGE 9 OF 144 I_

REFERENCE USE 5 .5 The plant operating mode that existed at the time that the event occurred, prior to any protective system or operator action initiated in response to the condition, is compared to the mode applicability of the EALs. If an event occurs, and a lower or higher plant operating mode is reached before the emergency classification can be made, the declaration shall be based on the mode that existed at the time the event occurred.

5 .6 Initiating condition and EAL Information is presented by recognition category:

5.6.1. A -Abnormal Rad Levels/ Radiological Effluent 5.6.2. C - Cold Shutdown/ Refueling System Malfunctions 5.6.3. E- Events Related to Independent Spent Fuel Storage Installations \

5.6.4. F - Fission Product Barrier Degradation 5.6.5. H- Hazards and Other Conditions Affecting Plant Safety 5.6.6. S- System Malfunction 5.7 ICs and EALs are numbered as follows:

NOTE All sequential numbers are not used in some !Cs to maintain standardization with NE! numbering and Entergy numbering system .. (For example, there is no SU2, SU3, SU4, and SU5 between SUI and SU6) 5.7.1. Initiating Conditions: X 1 X2 X 3 1

Category (A, C, E, F, H, S) 2 Classification (U-NOUE, A-Alert, S-SAE, G-GE) 3 Sequential IC number for classification level (e.g., AUi, AU2, HAI, HA2, etc) 5.7.2. EALs: sequential number for EAL in each IC XXX-# (e.g., AUl-1, AUI-2, etc.)

EIP-2-001 REV-028 PAGE 10 OF 144

REFERENCE USE 6 PROCED URE NOTE The assessment, classification, and declaration of an emergency condition is expected to be completed within 15 minutes after the availability of indications (i.e. plant instrumentation, plant alarms, computer displays, or incoming verbal reports) to plant operators that an EAL has been exceeded 6.1 Anytime an event occurs that has the potential of causing or resulting in a hazard to personnel, onsite or offsite, the Emergency Director:

6.1.1. Should review INITIATING CONDITIONs and EALs to determine if the event should be classified as an emergency.

6.1.2. Shall classify the emergency in accordance with this procedure and implement EIP-2-002, Classification Actions, if criteria are met.

6.2 River Bend Station Senior Management or designated alternate shall:

6.2.1. Provide assistance to the OSM, as requested, if the emergency is classified as an Unusual Event (NOUE).

6.2.2. Relieve the OSM of the responsibilities of Emergency Director as soon as practical for an ALERT or higher classification and implement applicable EIP procedures.

6.2.3. The Emergency Director will review this procedure and upgrade the emergency to a SITE AREA EMERGENCY or GENERAL EMERGENCY when warranted.

6.3 Declaration of an emergency class may not be necessary if it is discovered that an event or condition had existed that met an EAL threshold but that no emergency had been declared and the basis for the emergency class no longer exists at the time of the discovery. (REF 2.4) 6.3 .1. Cases of this nature, discovered well after the fact, may be due to a rapidly concluded event or an oversight in the emergency classification made during the event or it may be determined during a post-event review (e.g., routine log or record review).

6.3.2. Reporting requirements of 10CFR50.72 are applicable and the guidance ofNUREG -1022 may be applied.

6.3.3. Notify the St~te and local agencies by phone.

EIP-2-001 REV-02 8 PAGE 11 OF 144 I

1 __

REFERENCE USE 6.4 For some events, the condition may be corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g., coolant radiochemistry sampling, may be necessary). Classify the event as indicated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met. (REF 2.3) 6.5 Existing guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate Operator actions.

6.6 When two or more Emergency Action Levels are determined, declaration will be made on the highest classification level for the plant.

6.7 Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert 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.

6.8 EPP-2-503, RBS Equipment Important to Emergency Preparedness, provides guidance when planning to take an instrument used to determine EAL conditions out of service or following an UNPLANNED loss of the instrument. The OSM/CRS should perform the following:

6.8.1. Evaluate out-of-service equipment and determine if other instruments or compensatory measures are in place to assess for the associated EAL entry condition.

6.8.2. Evaluate site effects and implement a contingency plan if applicable.

6.9 Attachment 9 contains the USER AIDS available to the OSM / ED to use in determining the EAL.

7 DOCUMENTATION 7.1 NONE EIP-2-001 REV -028 PAGE 12 OF 144

--~---------------------------,

I REFERENCE USE ATTACHMENT 1 PAGE 1 OF8 INITIATING CONDITION MATRIX

' ~EC0:GNI%1bN/ ,,* ' * . GENERAL:* . .. '

. '\

~A:I:EGOJ.tY ... , . :, , El\iIBRGENCY' > . .ISITE,AREAEMERGENCY * * * .. ALERT;' .. ,.o_N()UE Abnormal AGl AS1 AAl AUl Rad Levels/ Offsite dose resulting from Offsite dose resulting from an Any release of gaseous or Any release of gaseous or Radiological an actual or IMMINENT actual or IMMINENT release of liquid radioactivity to the liquid radioactivity to the Effluent release of gaseous gaseous radioactivity > I 00 mR environment> 200 times the environment > 2 times the radioactivity > 1000 mR TEDE or 500 mR thyroid COE ODCM limit for::: 15 minutes ODCM limit for::: 60 minutes TEDE or 5000 mR thyroid for the actual or projected o'P 1v10 Lr d e: 1, 2, 3, 4, 5, Op Mode: 1, 2, 3, 4, 5, COE for the actual or duration of the release DEFUELED projected duration of the DEFUELED Op Mode: 1, 2., 3, 4, 5, release using actual DEFUELED meteorology Op Mode: 1, 2, 3, 4, 5, DEFUELED AA2 AU2 Damage to irradiated fuel or UNPLANNED rise in plant loss of water level that has radiation levels resulted or will result in the Op Mode: 1, 2, 3, 4, 5, uncovering of irradiated fuel DEFUELED outside the reactor vessel Op Mode: 1, 2, 3, 4, 5, DEFUELED AA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions OpMode: 1, 2, 3, 4, 5, DEFUELED EIP-2-001 REV-028 PAGE 13 OF 144

REFERENCE USE ATTACHMENT 1 PAGE20F8 INITIATING CONDITION MATRIX

• GENERAL' ,, ~*-

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~RECOGNITION' ~, :"

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CATEGORY: *EMERGENCY

-BITE AREA'ElVIERGENGY ' ,' ALERT* ',' Ii ':NODE.'

Fission FGl FS1 FAl FUl

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Product

....=

"' Loss of ANY two barriers Loss or potential Joss of ANY ANY loss or ANY potential ANY Joss or ANY potential Barrier Degradation

...-- :~ AND loss or potential loss of 0

two barriers loss of EITHER fuel clad or loss of containment

~

0

... the third barrier. Op Mode: I, 2, 3 RCS Op Mode: I, 2, 3

,-l Op Mode: I, 2, 3 Op Mode: I, 2, 3

~

Hazards and HGl HS1 HAl HUl Other HOSTILE ACTION HOSTILE ACTION within the HOSTILE ACTION within Confirmed SECURITY Conditions "'

.,;,. resulting in loss of physical PROTECTED AREA the OWNER CONTROLLED CONDITION or threat which Affecting =

r.l control of the facility AREA or airborne attack indicates a potential Plant Safety 0 threat degradation in the level of

• Op Mode

  I, 2, 3, 4, 5,

, Op Mode: I, 2, 3, 4, 5, safety of the plant

.,u DEFUELED Op Mode: I, 2, 3, 4, 5, 00 DEFUELED DEFUELED Op Mode: I, 2, 3, 4, 5, DEFUELED EIP-2-001 REV -028 PAGE 14 OF 144

I REFERENCE USE ATTACHMENT 1 PAGE3 OF8 INITIATING CONDITION MATRIX

':RECOGNifio:N*:: * * *  :.GENER.Ai.;" * * '*' <

r:t ,CAXiGORY,' ,;, ** . ,* EMERGENCY*:_

,, ~ - ' . <

... ,;Sl~E AREA EMERGENCY . ALERT r *NOUE Hazards and HG2 HS2 HA2 HU2 Other Other conditions exist which Other conditions exist which in Other conditions exist which Other conditions exist which Conditions ~ in the judgment of the the judgment of the Emergency in the judgment of the in the judgment of the Affecting =

o Emergency Director warrant Director warrant declaration of Emergency Director warrant Plant Safety it; Emergency Director warrant declaration of a GENERAL a SITE AREA EMERGENCY declaration ofan ALERT. declaration of a NOUE S EMERGENCY Op Mode: I, 2, 3, 4, 5, Op Mode: I, 2, 3, 4, 5, Op Mode: I, 2, 3, 4, 5, Op Mode: I, 2, 3, 4, 5, DEFUELED DEFUELED DEFUELED DEFUELED HS3 HA3 Control Room evacuation has Control Room evacuation has been initiated and plant control been initiated cannot be established Op Mode: I, 2, 3, 4, 5, OpMode: I, 2, 3, 4, 5, DEFUELED DEFUELED HA4 HU4 FIRE or EXPLOSION FIRE within PROTECTED affecting the operability of AREA boundary not plant safety systems required extinguished within 15 to establish or maintain safe minutes of detection or shutdown EXPLOSION within the Op Mode: I, 2, 3, 4, 5, PROTECTED AREA DEFUELED Op Mode: I, 2, 3, 4, 5, DEFUELED EIP-2-001 REV -028 PAGE 15 OF 144

REFERENCE USE ATTACHMENT 1

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NP:W1 Hazards and HAS Other Access to a VITAL AREA is Release of toxic, corrosive, Conditions prohibited due to toxic, asphyxiant or flammable gases Affecting corrosive, asphyxiant or deemed detrimental to Plant Safety flammable gases which NORMAL PLANT jeopardize operation of OPERATIONS operable equipment required

... Op Mode: 1, 2, 3, 4, 5, 0

to m~ntain safe operations or DEFUELED safely shutdown the reactor Op Mode: 1, 2, 3, 4, 5, DEFUELED

~ HA6 HU6

=

e 0 Natural or destructive Natural or destructive

.,= phenomena affecting VITAL phenomena affecting the

.c:

~

-;; AREAS PROTECTED AREA

~

= OpMode: 1, 2, 3, 4, 5, Op Mode: 1, 2, 3, 4, 5, z DEFUELED DEFUELED System SGl SSl SAl SUl Malfunction Prolonged loss of all offsite Loss of all offsite and all onsite AC power capability to Loss of all offsite AC power and all onsite AC power to AC power to emergency busses emergency busses reduced to a to emergency busses for 2::. 15 emergency busses for 2::. 15 minutes single power source for 2::. 15 minutes minutes such that any Op Mode: 1, 2, 3 OpMode: 1, 2, 3 Op Mode: 1, 2, 3 additional single failure would result in station blackout OpMode: 1, 2, 3 EIP-2-001 REV -028 PAGE 16 OF 144

REFERENCE USE ATTACHMENT 1 PAGES OF8 INITIATING CONDITION MATRIX System SG3 SS3 SA3 Malfunction Automatic scram and all Automatic scram fails to Automatic scram fails to manual actions fail to shutdown the reactor and the shutdown the reactor and the shutdown the reactor and manual actions taken from the manual actions taken from the indication of an extreme reactor control console are not reactor control console are challenge to the ability to successful in shutting down the successful in shutting down cool the core exists reactor the reactor OpMode: I, 2 OpMode: 1, 2 OpMode: I, 2 u SS4

~

0 Loss of all vital DC power for 2::

"'"'0 15 minutes

..;i Op Mode: I, 2, 3 EIP-2-001 REV -028 PAGE 17 OF 144

REFERENCE USE ATTACHMENT 1 PAGE60F8 INITIATING CONDITION MATRIX System =

0 SS6 SA6 SU6

.::o=

Malfunction Inability to monitor a UNPLANNED loss of safety UNPLANNED loss of safety

a"' SIGNIFICANT TRANSIENT system annunciation or system annunciation or

....=

--.s

"... I

'! in progress indication in the control room with either ( 1) a indication in the Control Room for ~ 15 minutes o=

Op Mode: 1, 2, 3 SIGNIFICANT TRANSIENT Op Mode: 1, 2, 3

=

= in progress, or (2)

=

=

< compensatory non-alarming 0

indicators are not available

"'"'0

...:l Op Mode: 1, 2, 3 SU7 Cl) i RCS leakage

~1...

Op Mode: 1, 2, 3 1 SUS

'="' :1: Loss of all onsite or offsite jJ communications capabilities.

l Op Mode: 1, 2, 3 C

SU9

.si Fuel clad degradation

~] Op Mode: 1, 2, 3

\

SUlO ij Inadvertent criticality

~1 "O

• o=

....=

Op Mode: 3 EIP-2-001 REV-028 PAGE 18 OF 144

REFERENCE USE ATTACH MENT 1 PAGE7 0F8 INITIAT ING CONDIT ION MATRIX System SUH Malfunction Inability to reach required operating mode within Technical Specification limits Op Mode: I, 2, 3 Cold CGl CSl CAl CUl Shutdown /

Loss of RCS/RPV inventory Loss of RCS/RPV inventory Loss of RCS/RPV inventory RCS leakage Refueling affecting fuel clad integrity affecting core decay heat with containment challenged OpMode:4, 5 OpMode:4 removal capability OpMode:4, 5 OpMode:4, 5 CO2 UNPLANNED loss of RCS/RPV inventory Op Mode: 5 CA3 CU3 Inability to maintain plant in UNPLANNED loss of decay cold shutdown heat removal capability with OpMode:4, 5 irradiated fuel in the RPV Op Mode: 4, 5 CA5

,

Loss of all offsite and all CU5 AC power capability to 0

i:i.. onsite AC power to emergency busses reduced to a u emergency busses for 2: 15 single power source for 2: 15

~

0 minutes minutes such that any

"'"'0 Op Mode: 4, 5, Defueled additional single failure would

~

result in station blackout Op Mode: 4, 5 EIP-2-001 REV -028 PAGE 19 OF 144

REFERENCE USE ATTACHMENT 1 PAGE80F8, INITIATING CONDITION MATRIX Cold CU6 u

Shutdown/ A

.... Loss of required DC power for 0

Refueling "' 2: 15 minutes

"'0

....;i Op Mode,' 4, 5 CU7 Inadvertent criticality Op Mode: 4, 5 CU8

....o. Loss of all onsite or offsite -

"' communications capabilities

"'0

....;i Op Mode: 4, 5, Defueled ISFSI "O"=

0::

E-HUl Damage to a loaded cask

=

0

~ CONFINEMENT

=.,

=

BOUNDARY

=

Op Mode: All

=0 u

EIP-2-001 REV-028 PAGE 20 OF 144

REFERENCE USE ATTACHMENT 2 PAGE 1 OF2 ABNORMAL RADIATION LEVELS/ RADIOLOGICAL EFFLUENT

~"' ,.. """' ....

AG!

l l l l*lslnl Offsite dose resulting from an actual or IMMINENT release 1 2 3 AS!

l l l l+lnl 1 2 3 AAI ltl2l,l,lslnl AU!

l1 l2131,lslnl Offsite dose resulting from an actual or IMMINENT release of Any release of gaseous or liquid radioactivity to the environment Any release of gaseous or liquid radioactivity to the of gaseous radioactivity> 1000 mR TEDE or 5000 mR gaseous radioactivity> I 00 mR TEDE or 500 mR thyroid CDE > 200 times the ODCM limit for 2: 15 minutes environment> 2 times the ODCM limit for 2: 60 thyroid CDE for the actual or projected duration of the for the actual or projected duration of the release minutes release using actual meteorology Ernergencl'. Action Level(s}: (1 or 2 or 3)

Emergency Action Level(s} (I or 2 or 3} NOTE: The Emergency Director .\'hould not wail until the Emergency Action Level(s}: (I or 2 or 3)

Emergencv Action Level(s}: (1 or 2 or 3) NOTE: The Emergency Director should not wail until the applicable time has elapsed, hut should declare the event NOTE: The Emergency Director should not wait until NOTE: The Emergency Director should not wait until the applicable lime has elapsed, but should declare the eve/J/ a.\* as soon as it is determined that the release duration has the applicable time has elapsed, but should declare the i applicable time has elapsed, but should declare the event as soon as ii is determined that the condition will likely exceed the exceeded, or will likely exceed, the applicable lime. In the event as soon as ii is determined that the release e soon as it is determined that the condition will likely exceed applicable linie. Ifdose assessment results are available, the absence <~fdata to the contrary, assume that the release duration has exceeded, or will likely exceed, the

~

the applicable time. If dose assei*unent results are available, the c/assijicalionshou/d be based on EAL #2 instead ofEAL c/assijication should be based on EAL #2 instead ofEAL#/. Do not delay declaration awaiting dose assessment results.

duration has exceeded the applicable time ifan ongoing release is detected and the release start time is unknown.

applicable time. In the absence ofdata to the contrary,

-~

~

1. I. Do not delay declaration awaiting dose assessment results.

assume that the release duration has exceeded the applicable lime ifan ongoing release is detected and 0 I. VALID reading on any of the radiation monitors in Table RI I. VALID reading on any of the radiation monitors in Table RI

a

~

I. VALID reading on any of the radiation monitors in Table > the SITE AREA EMERGENCY reading for 2: 15 minutes the release start lime is unknown.

I>: > the ALERT reading for 2. 15 minutes RI > the GENERAL EMERGENCY reading for 2: 15 OR OR 1. VALID reading on any of the radiation monitors minutes 2. Dose assessment using actual meteorolOb,Y indicates doses 2. For RMS-REI07 effiuent monitor: in Table RI > the NOUE reading for 2. 60 minutes OR > I 00 mR TEDE or 500 mR thyroid CDE at or beyond the EITHER

2. Dose assessment using actual meteorology indicates doses OR SITE BOUNDARY VALID reading> 200 times the alarm setpoint established 2. VALID reading on RMS-REI 07 effluent monitor

> 1000 mR TEDE or 5000 mR thyroid CDE at or beyond OR by a current radioactivity discharge permit for::'.:. 15 minutes > 2 times the alarm setpoint established by a the SITE BOUNDARY 3. Field survey results indicate closed window dose rates> 100 OR OR current radioactivity discharge permit for 2. 60 mR/hr expected to continue for 2:. 60 minutes; or analyses of VALID reading> 1.27E-01 µCi/ml for :o:_ 15 minutes

3. Field sufVey results indicate closed window dose rateS > minutes field survey samples indicate thyroid CDE > 500 mR for one OR I 000 mR/hr expected to continue for:::_ 60 minutes; or OR hour of inhalation, at or beyond the SITE BOUNDARY 3. Confirmed sample analyses for gasequs or liquid releases analyses of field survey samples indicate thyroid CDE > 3. Confirmed sample analyses for gaseous or liquid indicate concentrations or release rates> 200 times the releases indicate concentrations or release rates 5000 mR for one hour of inhalation, at or beyond the ODCM limit for 2:. 15 minutes > 2 times the ODCM limit for:::_ 60 minutes SITE BOUNDARY 4.50E+o8 fLCi/sec 4GE125 4.50E+o7 fLCi/sec 4GE125 3.06E+o7 µCi/sec 4GE!25 3.06E+05 µCi/sec NIA IGE!26 2. 82E-O 1 µCi/ml IGE!26 5.26E-03 µCi/ml l.OOE+09 /LCi/sec 4GE005 1.00E+o8 µCi/sec 4GE005 2. l 9E+o6 µCi/sec 4GE005 2. l 9E+o4 ftCi/sec NIA 5GE005 2.82E-Ol µCi/ml 5GE005 4.65E-03 µCi/ml NIA 4GE006 2.58E+o6 ftCi/sec 4GE006 2.58E+o4 µCi/sec 5GE006 6.84E-02 tCi/ml 5GE006 6.84E-04 Ci/ml Plant Modes (white boxes indicate applicable modes) 1 Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdmvn Refuel D Defueled EIP-2-001 REV-028 PAGE 21 OF 144

REFERENCE USE ATTACHMENT2 PAGE20F2 ABNORMAL RADIATION LEVELS/ RADIOLOGICAL EFFLUENT AA2 AU2 l1 l2 l3 i41 5 1°1 Damage to irradiated fuel or loss of water level that has resulted UNPLANNED rise in plant ~adiation levels or will result in the uncovering of irradiated fuel outside the reactor vessel Emergency Action Level(s): (1 or 2)

I. a. UNPLANNED water level drop in a reactor Emergency Action Level(s): (I or 2) refueling pathway as indicated by any of the

l. A water level drop in the reactor refueling cavity, spent fuel following:

pool or fuel transfer canal that will result in irradiated fuel a. Water level drop in-the reactor refueling becoming uncovered cavity, spent fuel pool, or fuel transfer OR canal indication on Control Room Panel

2. A VALID reading on any of the following radiation monitors 870 due to damage to irradiated fuel or loss of water level: b. Personnel observation by visual or remote RMS-REl40 2000 mR/hr means.

RMS-REl41 2000 mR/hr AND RMS-REl92 2000 mR/hr b UNPLANNED VALID area radiation monitor RMS-RE 193 2000 mR/hr alarm on any of the following:

RMS-RESA 1.64E+o3 fLCi/sec RMS-REI40 RMS-RE5B (GE) 5.29E-04 µCi/ml RMS-RE141 RMS-REl92 RMS-REl93 OR

2. UNPLANNED VALID area radiation monitor readings or survey results indicate a rise by a factor of 1000 over normal* levels NOTE: For area radiation monitors with ranges incapable ofmewmring 1000 limes normal* level\',

dassijkation shall he based on VALlD fit/I scale indications unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes o/lhe area radiation monitor indication.\*

~

goingfilll scale.

E 0

=

..c *Normal can be considered the highest reading in the

< past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> excluding the current peak value.

AA3 Rise in radiation levels within the facility that impedes operation of systems required to maintain plant s~fety functions Emergency Action Level(s):

I. Dose rate> 1S mR/hr in any of the following areas requiring continuous occupancy to maintain plant safety functions:

Main Control Room CAS Plant Modes (white boxes indicate applicable modes) l Power Operations 2 Startup 3 Hot Shutdown ..J Cold Shutdown 5 Refuel D Dcfuelcd EIP-2-001 REV-028 PAGE 22 OF 144

REFERENCE USE A TIACHMENT 3 PAGE I OF2 FISSION PRODUCT BARRIER FGI FSI FA! FU!

Loss or potential loss of ANY two barriers ANY loss or ANY potential loss of EITHER fuel clad or RCS ANY loss or ANY potential loss of containment Emergency Action Level(s):

Emergency Action Level(s): Emergency Action Level(s):

Emergency Action Level(s): I. Loss or potential loss of any two barriers I. Loss of any two barriers l. Any loss or any potential loss of fuel clad I. Any loss or any potential loss of containment AND OR Loss or potential loss of the third barrier Any loss or any potential loss of RCS

• FUEL CLAD (FC)'Banier . . ' ,'.*,- :j'., *'. REAC:TOR CO~L.,.NT SYST)lM. (RC) Barrier**

I .. .. . / ,:> I ,.

PRIMARY.~ONTAIN,MENT(PC) Barrier

• Parameter* Potential-Loss, . , Par?ITieter ;, ~

. <Potential LOSS.~ '.

FCl Primary Coolant activity None RCI Drywell Drywell pressure> 1.68 psid with None PC I Primary containment I. Rapid unexplained loss of PC pressure I. PC pressure> 15 psig and rising coolant > 300 µCi/gm dose pressure indications of reactor coolant leak in conditions following initial pressure rise OR activity JeyeJ equivalent I-131 d1ywell OR 2. a. PC hydrogen in lhe unsafe zone

2. PC pressure response not consistent with ofHDOL cun*e LOCA conditions OR

b. DW hydrogen concentration > IJ%

OR

3. RPV pressure and suppression pool temperature cannot be maintained below the HCTL FC2 Reactor vess_el RPV water level cannot be RPV water level cannot be RC2 Reactor RPV water level cannot be restored None PC2 Reactor vessel water None Entry into PC flooding procedures water level restored and maintained restored and maintained abm*e Yessel and maintained abO\'e -162 inches or le,*el SAP-I and SAP-2 above -187 inches -162 inches or cannot be waterleYel cannot be detennined determined FC3 Primary Containment radiation None RC3 RCS Leak I. UNISOLABLE main steam I. RCS leakage > 50 gpm PC3 Primary containment I. a. Failure of all valves in any one line to None Containment monitor RMS-RE16 reading Rate line break as indicated bv the inside the drywell isolation failure or close radiation > 3,000 R/hr failure of both MS1Vs i~ any OR bypass monitors ~

one line to close 2. UN1SOLABLE RCS h. Direct do\\nstream pathway to the AND leakage outside PC as environment exists after PC isolation indicated by exceeding Sib'Ilal High MSL now annunciator either of the following: OR (P60l-19A-A2)

a. Ma.x Normal Operating 2. Intentional PC venting per EOPs or SAPs OR Temperature (Table F2) OR

2. Indication of an OR 3. UNISOLABLE RCS leakage outside PC as UNlSOLABLE

b. Max Normal Area indicated by exceeding either of the HPCS, feedwater, RWCU or Radia1ion (Table F2) following:

RClC break OR a. Ma'\. Safe Operating Temperature (Table

3. Emergency RPV Fl) depressurization is required OR

b. Ma.x Safe Area Radiation (Table Fl)

RC4 Drywell Drywell radiation monitor RMS- None PC4 Primary containment None Containment radiation monitor radiation RE20 reading > I 00 R/hr due to radiation monitors RMS-REl6 reading> 10,000 R/hr reactor coolant leakage FC4 Emergency Any condition in the Any condition in the opinion of RC5 Emergency Any condition in the opinion of the Any condition in the opinion PCS Emergency Director Any condition in the opinion of the Any condition in the opinion of the Director opinion of the Emergency the Emergency Director that Director Emergency Director that indicates of the Emergency Director judgment judgment Emergency Director that indicates loss of the Emergency Director that indicates Director that indicates loss indicates potential loss of the judgment loss of the RCS barrier that indicates potential loss of Primary Containment harrier potential loss of the Primary of the Fuel Clad barrier Fuel Clad barrier the RCS barrier Containment barrier Plant Modes (white boxes indicate applicable modes) I Power Operations 2 S!artup 3 Hot Shutdown 4 Cold Shutdown Refuel D Dcfuclcd EIP-2-001 REV-028 PAGE 23 OF 144

REFERENCE USE ATTACHMENT 3 PAGE20F2 FISSION PRODUCT BARRIER

" \/,'*

,~

"TABLEFl ,'. '

TABLEF2' PC:3 Loss of Primary,Contilinment' ,

• C RC 3 Potential Loss of RCS Parameter Area Temperature Area Radiation Level Parameter Area Temperature Area Radiation Level Max Safe Oneratinrr Value DRMSGrid2 Max Safe Oneratin., Value (isolation temgeraiure DRMSGrid2 Max Normal Oger!:lting RHR A eauioment area 200° F 1213 9.5E-+-03 mR/hr alarm) Value RHR B eauinment area 200' F 1214 9.5E-+-03 mR/hr RHR A equipment area 117'F 1213 8.2E+Ol mR/hr RHR C eouioment area NIA 1215 9.5E+03 mR/hr (P601-20A-B4)

RCIC room 200°F 1219 9.5E+03 mR/hr RHR B equipment area 117'F 1214 8.2E+Ol mR/hr MSL Tunnel 200'F NIA IP601-20A-B4)

RWCU pump room I 200°F NIA RHR C eauioment area NIA 1215 8.2E+Ol mR/hr (A)l2(B) RCIC room 182'F 1219 l.20E+02 mR/hr

'/P601-21 A-B6)

MSLTunnel 173' F NIA (P601-l 9A-AIIA3IB I/B3)

RWCU pump room I (A) I 2 (B) 165'F NIA (P680-I A-A2/B2)

EIP-2-001 REV-028 PAGE 24 OF 144 L__

REFERENCE USE ATTACHMENT 4 PAGE 1 OF3 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY

. , SITE AREA EMERGENCY ALERT .' NOUE HG!

HOSTILE ACTION resulting in loss of physical control of the HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED Confirmed SECURITY COND1TION or threat which facility AREA or airborne attack threat Emergency Action Level(s): indicates a potential degradation in the level of safety of Emergencv Action Levcl(s): (I or 2) Emergency Action Level(s): (1 or 2) the plant

1. A HOSTILE ACTION is occurring or has occurred

1. A HOSTILE ACTION has occurred such that plant personnel within the PROTECTED AREA as reported by the I. A HOSTILE ACTION is occurring or has occurred within the Emergency Action Level(s): (1 or 2 or 3) are unable to operate equipment required to maintain safety RBS security shift supervision OWNER CONTROLLED AREA as reported by the RBS 1. A SECURITY CONDITION that does NOT involve a functions security shift supervision HOSTILE ACTION as reported by the RBS security shift OR supeivision

2. A HOSTILE ACTION has caused failure of Spent Fuel 2. A validated notification from NRC of an airliner attack threat OR Cooling Systems and I!\.1MINENT fuel damage is likely for within 30 minutes of the site 2. A credible site specific security threat notification a freshly off-loaded reactor core in pool OR

3. A validated notification from NRC providing information of an aircraft threat HG2 HS2 HA2 HU2 Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Emergency Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency Director warrant declaration of a SITE Director warrant declaration of an ALERT Emergency Director warrant declaration of a NOUE Emergency AREA EMERGENCY Emergency Action Level(s): Emergency Action Level(s):

Emergency Action Level(s): Emergency Action Level(s): I. Other conditions exist which in the judgment of the I. Other conditions exist which in the judgment of the

1. Other conditions exist which in the judgment of the 1. Emergency Director indicate that events are in Emergency Director indicate that events are in progress or Emergency Director indicate that events are in Emergency Director indicate that events are in progress progress or have occurred which involve actual or have occurred which involve an actual or potential substantial progress or have occurred which indicate a potential or have occurred which involve actual or IMMINENT likely major failures of plant functions needed for degradation of the level of safety of the plant or a security degradation of the level of safety ofthe plant or substantial core degradation or melting with potentiaJ for protection of the public or HOSTILE ACTION that event that involves probable life threatening risk to site indicate a security threat to facility protection has been loss of containment integrity or HOSTILE ACTION results in intentional damage or malicious acts; (I) perso1U1el or damage to site equipment because of HOSTILE initiated. No releases of radioactive material requiring that results in an actual Joss of physical control of the toward site personnel or equipment that could lead to ACTION. Any releases are expected to be limited to small offsite response or monitoring are expected unless facility. Releases can be reasonably expected to exceed the likely failure of or: (2) that prevent effective fractions of the EPA Protective Action*Guideline exposure further degradation of safety systems occurs EPA Protective Action Guideline exposure levels offsite access to equipment needed for the protection ofthe levels for more than the immediate site area public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY Plant Modes (white boxes indicate applicnblc modes) I Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown Refuel D Defueled EIP-2-001 REV-028 PAGE 25 OF 144

REFERENCE USE ATTACHMENT 4 PAGE20F3 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Control room evacuation has been initiated and plant control carutot Control room evacuation has been initiated be established Emergency Action Level(s):

Emergency Action Level(s): I. AOP-Ol31,Shu!dmmfiom0u1sidetheMainCmtrolRoom"'11Jll'SCClltrol L a Control room evacuation has been initiated Rocmevocuaticn e0 0

~

b. Control of the plant cannot be established in accordance e with AOPP003 l, Shutdown from Outside the Main Control

=

8 Room, within 15 minutes HA4 FIRE or EXPLOSION affecting the operability of plant safety systems FIRE within the PROTECTED AREA not extint,ruished within 15 required to establish or maintain safe shutdown minutes of detection or EXPLOSION within the PROTECTED AREA Eme1:gency Action Level(s): Emergency Action Level(s): (I or 2)

1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any N11te: The Emergene,y Director should not wait until the applicable time of the structures or areas in Table H2 containing safety systems has elapsed, but should dec:lare the event as soon as ii is determined or components or Control Room indication of degraded that the duration has exceeded, or will likely exceed, the applicable performance of those safety systems time.

I. FIRE not extinguished within 15 minutes of Control Room notification or verification ofa Control Room FIRE alarm in any Table H2 structure or area

2. EXPLOSION within the PROTECTED AREA HAS HU5 Access to a VITAL AREA is prohibited due to toxic, Release of toxic, corrosive, asphyxiant or flammable gases deemed corrosive, asphyxiant or flammable gases which jeopardize detrimental to NORMAL PLANT OPERATIONS operation of operable equipment required to maintain safe operations or safely shutdown the reactor Eme1:gency Action Level(s): (I or 2)

I. Toxic, corrosive, asphyxiant or flammable gases in Emergency Act_ion Level(s):

amounts that have or could adversely affect NORMAL PLANT Note: If the equipment in the stated area was already OPERATIONS inoperable, or 0111 ofservice, before the event occurred, then OR this EAL should not be declared as it will have no adverse 2. Report by West Feliciana Parish for evacuation or sheltering of impact on the ability ofthe plant to safely opera~e or safely site personnel based on an offsite ev shutdown beyond that already allowed by Technical

!:JJ1ecijications at the time ofthe event.

1. Access to Main Control Room, Auxiliary Building, or 95' Control Buildinga VITAL AREA (Table HZ) 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 PlantModes(whiteboxcsindicateapplicablemodcs) l PowcrOperations 2 Startup 3 HotShutdown 4 ColdShutdown 5 Refuel D Defuelcd EIP-2-001 REV-028 PAGE 26 OF 144

REFERENCE USE ATTACHMENT 4 PAGE 3 OF3 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY Natural or destructive phenomena affecting VITAL ARE Emergency Action Level(s): (I or 2 or 3 'or 4 or 5 or 6)

I. a Seismic eyent > Operating Basis Earthquake (OBE) as indicated by: Emergency Action Level(s): (1 or2 or3 or 4 or 5)

Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P68ti-o2A-D06) I. Seismic event identified by any 2 of the following; AND Seismic event confirmed by activated seismic switch as ERS-NBR3D TRIGGER (RECORD STARl) is yellow indicated by receipt of EITHER a OR b:

AND a Anmmciator'*SEISMIC SYS RECORDING/

Receipt of EITHER I OR 2; TROUBLE" (P68CHl2A-D06)

I. Annunciator '*Seismic E,,ent High" (P680-02A-C06)

b. Event Indicator on ERS-N8R3D TRIGGER (RECORD

2. Annunciator '*Seismic Eyent High-High" START) is yellow (P680-02A-B06) AND ERS-NBRJA QBE (HI) yellow light Earthquake felt in plant AND National Earthquake Center

b. Earthquake confirmed by any of the following: OR

~ Earthquake felt in plant

= 2. Tornado striking within the PROTECTED AREA batmdary e

0 Parameter Value/ Indicator Reactor Standby National Earthquake Center

• Control Room indication of degraded performance of OR 5 Aux Bldg Crescent 6 inches above floor Building Cooling systems required for the safe shutdown of the plant 3. Internal flooding that has the potential to affect safety related o'.; Area 70'EL (must be verified locally) Tower OR equipment rt.XJuired by Technical. Specifications for the current

~

2. Tornado striking resulting in VISIBLE DAMAGE to any of the operating mode in any Table HI area

~ HPCS Room 70'EL 4 inches above floor Auxiliary Diesel Table H2 structures or areas containing safety systems or OR

~ (P870-51A-G4) Building Generator components or Control Room indication of degraded 4. Turbine failure resulting in casing penetration or damage to g RHRARoom 70'EL 4 inches above floor Building OR performance of those safety systems OR turbine or generator seals

~

0 (P870-51A-G4) Control Tunnels (B, D, 3. Internal flooding in Auxiliary Building 70 ft elevation resulting in

5. Severe weather or hunicane conditions with indication ofSUSTAINED an electrical shock hazard that precludes access to operate or E RHRB Room 70'EL 4 inches above floor Building E,F,G) monitor safety equipment or Control Room indication of degraded high \\inds,:: 74 mph within the PROTECTED AREA boundruy E

~ (P870-5 l A-G4) Fuel Building performance of those safety systems z OR RHR C Room 70'EL 4 inches above floor 4. Turbine failure-generated PROJECTILES resulting in VISIBLE (P870-51A-G4) DAMAGE to or penetration of any of the Table H2 structures or areas containing safety systems or components or Control Room LPCS Room 70'EL 4 inches above floor indication of degraded performance of those safety systems (P870-51A-G4) OR RCIC Room 70'EL 4 inches above floor 5. Vehicle crash resulting in VISIBLE DAMAGE to any of the Table H2 structures or areas containing safety systems Or components or 870-51A-G4 Control Room indication of degraded performance of those safety systems OR G. Hurricane or high SUSTAINED wind conditions 2: 74 mph within the PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the Table H2 structures or areas containing safety systems or components or Control Room indication of degraded performance of those safety systems EIP-2-001 REV-028 PAGE 27 OF 144

REFERENCE USE ATTACHMENTS PAGE 1 OF4 SYSTEM MALFUNCTION

. """'" I ii, h I !,'In!

SGI I, h h l,T.fol SSI I I, I l>lslnl SAl 1,1,h!,liih SUI Prolonged loss of all offsite and all onsite AC power to Loss of all offsite and all onsite AC power to emergency AC power capability to emergency busses reduced to a single Loss of all offsite AC power to emergency busses emergency busses busses for 2:. I 5 minutes power source for 2:. 1S minutes such that any additional single for 2:. 15 minutes failure would result in station blackout Emergency Action Level(s): Emergency Action Level{s): Emergency Action Level{s):

I. a Loss of all offsite and all onsite AC power to Div I, II and III Emergency Action Level(s):

Note: The Emergency Director should not wait until the Note: The Emergency Director should not wail until ENS busses applicable time has elapsed, but should declare the event a.,* Note: 171e Emergency Director should not wait until the the applicable lime has elapsed, but should declare AND soon as ii is determined that the condition has exceeded, or applicable time has elap.~ed, but should declare the event as soon the event as soon as ii is determined that the will likely exceed, the applicable time. a.,* it is determined that the condition has exceeded, or will likely condition has exceeded, or will likely exceed, the

b. Either of the following:

exceed, the applicable time. applicable time.

• Restoration of at least one emergency bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely I. Loss of all offsite and all onsite AC power to Div I, II and III

1. a. AC power capability to Div I and Div II ENS busses ENS busses for 2: 15 minutes I. Loss of all offsite AC power to Div I and II ENS OR reduced to a single power source for 2:. 1S minutes busses for::::_ I 5 minutes

• RPV level cannot be maintained AND

> -162 inches b. Any additional single failure will result in a station blackout SG3 11121:il*l~'liil SS3 Automatic scram and all manual actions fail to shutdown the AutomatiC scram fails to shutdown the reactor and the manual Automatic scram fails to shutdown the reactor and the manual reactor and indication ofan extreme challenge to the ability to cool actions taken from the reactor control console are not actions taken from the reactor control console are successful in the core exists successful in shutting down the reactor shutting down the reactor Emergency Action Level(s):

I . a. An automatic scram failed to shutdown the reactor Emergency Action Level{s): Emergency Action Level(s):

AND I. a. An automatic scram failed to shutdown the reactor I. a. An automatic scram failed to shutdown the reactor

b. All manual actions do not shutdown the reactor as indicated by reactor power:::, 5%

b. Manual actions taken at the reactor control console b. Manual actions taken at the reactor control console AND do not shutdown the reactor as indicated by reactor successfully shutdown the reactor as indicated by

c. Either of the following exist or have occurred due to power:::,5% reactor power < 5%

continued power generation:

• Core cooling is extremely challenged as indicated by RPV level can not be maintained> -187 inches OR

• Heat removal is extremely challenged as indicated by RPV pressure and Suppression Pool temperature cannot be maintained in the EOP Heat Capacity Temperature Limit (HCTL) Safe Zone SS4 I, l,hl,lslnl Loss of all vital DC power for:::, I 5 minutes

~

~

C Emergency Action Level(s):

0..

u Note: The Emergency Director :,lwuld not wait until the Q

~ applicable lime has elapsed, .but should declare the e1,*ent as C

soon as ii is determined that the condition has exceeded, or C

..l will likely exceed, the applicable lime.

1. < 105 VDC on all vital DC busses for 2: 15 minutes EIP-2-001 REV-028 PAGE 28 OF 144

REFERENCE USE ATTACHMENT 5 PAGE20F4 SYSTEM MALFUNCTION

~,_:.:: :t.* .. _:. ~_(?~N~RAL _JE~I_ER£~NCY~ _ ; __ *- --E AA SITE AREA EMERGENCY ALERT NOUE SS6 SA6 SU6 lil2hl,lslnl I, 121,blslnl I, I, h L lslnl Inability to monitor a SIGNIFICANT TRANSIENT in UNPLANNED loss of safety system annunciation or indication UNPLANNED loss of safety system progress in the control room with either (I) a SIGNIFICANT annunciation or indication in the Control Room TRANSIENT in progress, or (2) compensatory non-alarming for?. 15 minutes Emergency Action Level(s): indicators are not available Note: The Emergency Director should not wail until the Emergency Action Level{s}*

applicable lime hm* elapsed, but .,*hould declare the event as Emergency Action Level(s):

= soon as it is determined that the condition has exceeded, or Note: The Emergency Director should no/ wait until i:a will likely exceed, the applicable time.

I. a . UNPLANNED loss of> approximately 75% of the Note: The Emergency Director should not wail until the applicable time has. elapsed, hut should declare the event as :won as it is determined that the condition ha.\' exceeded, or will likely the applicable time has elap...ed, but should declare the event as soon as ii is determined that the

--~..5= following for:::, 15 minutes:

exceed, the applicable time.

I. a. UNPLANNED loss of> approximately 75% of the condition has exceeded, or will likely exceed, the applicable time.

I. UNPLANNED loss of> approximately 75% of

.g Control Room safety system annunciation following for:::, I 5 minutes:

-(

. =

=

= .

OR Control Room safety system indication

. Control Room safety system annunciation the following for:::, I 5 minutes:

a. Control Room safety system annunciation OR

~

Q Q

AND

b. A SIGNIFICANT TRANSIENT is in progress

. Control Room safety system indication b.

OR Control Room safety system indication AND AND C. Compensatory indications are unavailable

b. Either of the following:

A SIGNIFICANT TRANSIENT is in progress OR Compensatory indications are unavailable SU7 RCS leakage I I, h I !,In/

~

Emergenc1: Action Level{s}: (I or 2}

~ NtJte: A relief valve that operates and fail\' lo close

...,""g per design should he co1uidered applicable if the relief valve cannot be isolated

"'u ex:

I. Unidentified or pressure boundary leakage > I 0 gpm OR

2. Identified leakage> 35 gpm EIP-2-001 REV-028 PAGE 29 OF 144

REFERENCE USE ATIACHMENT5 PAGE 3 OF4 SYSTEM MALFUNCTION

,.. SUS

• 1,1,1,1,b Loss of all onsite or offsite communications capabilities Emergen!J: Action Level{s}: {I or 2)

I. Loss ofall of the following onsite communications methods affecting the ability to

.~ perform routine operations:

-~= Plant radio system Plant paging system E Sound powered phones

§ In-plant telephones

...u C

OR

2. Loss ofall of the following offsite

~ communications methods affecting the ability to perform offsite notifications:

All telephones NRC phones State of Louisiana Radio Offsite notification system and hotline Tati1e's1 I I, I, I, I lnl SU9

• '.FLOW

_Dose Rate Limit .. Fuel clad degradation (elm) (mR.IIJr) Emergencl'. Action Level{s}: ( I or 2)

I. Offgas pre-treatment radiation monitor reading>

<15 9000 the Table SI Dose Rate Limit for the actual

>15-17 8000 indicated offgas flow indicating fuel clad

-~=

-;; >17-20 7000 degradation > T.S. allowable limits OR E

A

. >20-25

>25-30 5000 4000

2. Reactor coolant sample activity value indicating fuel clad degradation > T.S. allowable limits

a. > 4.0 µCi/gm dose equivalent 1-131

=

>30-60 2000 OR a"" >60-140

>140-200 1000 700

. > 0.2 µCi/gm dose equivalent 1-131 for>

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> SUIO 11 bhlJ I ,Jnl

=~::: >, Inadvertent criticality a:; ~ Emergeng: Action Level(s}:

-6 :a

]U I. UNPLANNED sustained positive period observed on nuclear instrumentation EIP~2-001 REV-028 PAGE 30 OF 144

---

i REFERENCE USE ATIACHMENT 5 PAGE40F4 SYSTEM MALFUNCTION r___ ~ -~ ~

-- -GEN~R.e_\~~~!"lli:RG~~CY

-- SITE AREA E!\IERGENCY ALERT NOUE

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Inability to reach required operating mode within Technical Specification limits Emergena: Action Level{s}:

U .5 w...i I. Plant is not brought to required operating mode

"' within Technical Specifications LCO Action Statement time EIP-2-001 REV-028 PAGE 31 OF 144

REFERENCE USE ATTACHMENT 6 PAGE 1 OF3 COLD SHUTDOWN/ REFUELING "CG! CAI CUI Loss ofRCS/RPV inventory affecting fuel clad integrity - Loss of RCS/RPV inventory affecting core decay heat Loss of RCS/RPV inventory RCS leakage with containment challenged removal capability Emergency Action Level(s): {I or 2) Emergency Action Level(s)*

Emergency Action Level(s): (1 or 2) Emergency Action Level(s): (I or 2 or 3) NOTE: 11,e Emergency Direclor should not wail until lhe NOTE: The Emergency Director should not wait until the I. a. RPV level -162 inches (T AF) for 2: 3 0 minutes NOTE: The Emergency Director should not wait until the applicable time .,ws elapsed, but should declare the event a.,*

applicable time ha.,* elapsed, hut should declare the event a.,* :won a.,* ii is de/ermined that the condition will likely exceed the applicable time ha.,* elap.,*ed, but should declare lhe event as soon as ii is determined that the condition will likely exceed AND soon as if is determined thal the condition will likely exceed applicable lime. the applicable time.

the applicable lime.

b. Any containment challenge indication in Table Cl I. UNPLANNED loss of RCS inventory as indicated by RPV I. RCS leakage results in the inability to maintain or I. With CONTAINMENT CLOSURE not established, level < -43 inches (Level 2) restore RPV level> +9.7 inches (Level 3) for 2: 15 UNPLANNED RPV level < -49 inches minutes

2. a. RCS level cannot be monitored with core uncovery indicated by any of the following for

_ 30 minutes: 2. RCS level cannot be monitored for> 15 minutes with a loss

2. With CONTAINMENT CLOSURE established, RPV of RCS inventory as indicated by a;unexplained rise in floor

• RMS-REl6 reading> 100 R/hr level <-162 inches (TAF) or equipment sump level, Suppression Pool level, vessel

• Erratic Source Range Monitor indication make-up rate or observation ofleakage or inventory loss

• Unexplained rise in floor or equipment sump OR level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss 3. RCS level cannot be monitored for 2:. 30 minutes with a loss of RCS inventory as indicated by any of the following:

b. Any containment challenge indication in Table Cl RMS-REl6 reading> 100 R/hr I .. ... . Table Cl. ' .. :I, ...

Erratic Source Range Monitor indication Unexplained rise in floor or equipment sump level,

• ~:co11t8illinent C1ia11enee.IndicatiOits '.'.'

• CONTAINMENT CLOSURE not Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss established Explosive mixture inside containment UNPLANNED rise in containment

. pressure Secondary containmnent area radiation monitor above EOP Max Safe Operating Value below:

Area ORMS Max Safe Grid2 Operating Vaine RHREqnip 1213 9.SE+03 mR/hr Rm A RHREqnip 1214 9.SE+03 mR/hr RmB RHREqnip 1215 9.SE+03 mR/hr RmC Plant Modes (white boxes indicate applicable modes) I Power Operations 2 Startup 3 Hot Shutdown 4 Cold Shutdown Refuel D Dcfuclcd EIP-2-001 REV-028 PAGE 32 OF 144

I REFERENCE USE ATTACHMENT 6 PAGE 2 OF3 COLD SHUTDOWN/ REFUELING

' '~ '* ..

CU2 UNPLANNED loss ofRC~/RPV inventory lilljll~lslol Emergency Action Level(s} ( I or 2)

Note: 11,e Emergency Director should 1101 wait until the applicable time has elapsed, hut should declare the event as soon as ii is determined that lhe condition will likely exceed the applicable time.

t'

.s l. UNPLANNED RCS level drop as indicated by either of

~ the following:

~

..5

> a. RCS water level drop below the RPV flange for

~

u 2:. 15 minutes when the RCS level band is established above the RPV flange

~

0 0

OR

b. RCS water level drop below the RPV level band for:::.

15 minutes when the RCS level band is established below the RPV flange OR

2. RCS level cannot be monitored with a loss of RCS inventory as indicated by an unexplained rise in floor or equipment sump level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss CA3 CU3 Table C2 RCS Reheat Duration Thresholds lrl213Hsiol l1 li'314'slol

~ Inability to maintain plant in cold shutdown UNPLANNED loss of decay heat removal capability

~ RCS Containment Duration s Closure Emergency Action Level{s}: {1 or 2}

with irradiated fuel in the RPV o!

Intact NIA 60 minutes*

I. An UNPLANNED event results in RCS temperature Emergency Action Level{s): (I or 2)

Note: 111e Emergency Director should not wait until the

~

Not intact Established 20 minutes*

OR

> 200 °F > the specified duration in Table CZ applicable lime ha.\' elapsed, but should declare the event as

.rnon as iii,\' determined that the condilion will likely exceed the applicable time.

Not Established o minutes

'o *If an RCS heat removal system is in operation within

~ 2. An UNPLANNED event results in RCS pressure rise> I 0 I. An UNPLANNED event results in RCS temperature

...,0 this time frame and RCS temperature is being reduced, psig due to a loss of RCS cooling exceeding 200°F the EAL is not aoolicable.

OR

' 2. Loss of all RCS temperature and RCS/RPV level indication for 2: 15 minutes Plant Modes (white boxes indicate applicable modes) I Power Opemtions 2 Startup 3 Hot Shutdown 4 Cold Shutdown Refuel D Dcfuclcd EIP-2-001 REV-028 PAGE 33 OF 144

REFERENCE USE ATTACHMENT 6 PAGE30F3 COLD SHUTDOWN/ REFUELING Loss of all offsite and all onsite AC power to emergency busses AC power capability to emergency busses reduced to a for 2:, 15 minutes single power source for 2:, 15 minutes such that any additional single failure would result in station blackout Emergency Action Level(s): Emergency Action Level{s):

N(}fe: 17,e Emergency Director should no/ wail until the Note: 111e Emergency [Ji rector should not wail until the applicable time has elapsed, but should declare the event as :won applicable time has elapsed, hut should declare the event as as ii is determined that the condition will likely exceed the soon as ii is determined that the condition will likely exceed applicable time. the applicable time.

I. Loss of all offsite and all onsite AC power to Div I and Div II I. a. AC power capability to Div I and Div II ENS ENS busses for:::, 15 minutes busses reduced to a single power source for~ 15 minutes

b. Any additional single power source failure will result in station blackout CU6 Loss of required DC power for 2: 15 minutes

!1!*Hs!o!

Emergency Action Level(s):

Note: 17,e Emergency Director should not wait until the applicable lime has elapsed, hut should declare the event a.t

.won as ii is determined that the condition will likely exceed the applicable time 0

...l I. < I 05 VDC on required Vital DC busses for 2: 15 minutes CU7 Inadvertent criticality Emergency Action Level{s):

I. UNPLANNED sustained positive period observed on nuclear instrumentation CUB Loss of all onsite or offsite communications capabilities Emergency Action Level(s): (1 or 2)

I. Loss of all of the following onsite communication methods affecting the ability to perform routine operations:

Plant radio system Plant paging system Sound powered phones In-plant telephones OR

2. Loss of all of the following offsite communication methods affecting the ability to perform offsite notifications:

All telephones NRCphones State of Louisiana Radio Offsite notification system and hotline EIP-2-001 REV-028 PAGE 34 OF 144

• -*- - - * ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - : - - - - - - - - - - -

i REFERENCE USE ATTACHMEN T?

PAGE 1 OF 1 EVENTS RELATED TO ISFSI

-- ::-1[ --- 7 GENERAL EMERGENCY -

~~ .,i ** ~

SITE AREA EMERGENCY ALERT NOUE

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i:l Damage to a loaded cask CONFINEMENT BOUNDARY a"' Emergency Action Levels{s}:

1. Damage to a loaded cask CONFINEMENT BOUNDARY EIP-2-001 REV-028 PAGE 35 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 1 OF 107 EAL BASES BASES TABLE OF CONTENT RECOGNITION CATEGORY Introduction and Background Information ...................................................................................... 3 General Notes on Basis.Document Use ........................................................................................ 3 Emergency Classification Level Thresholds ................................................................................ ~

Emergency Action Levels (EALs) ............................................................................................... .4 Treatment of Multiple Events and Classification Level Upgrading ............................................. 6 Emergency Classification Level Downgrading ........ :................................................................... 6 Classifying Transient Events ......................................................................................................... 6 Operating Mode Applicability ...................................................................................................... 7 CATEGORY A -Abnormal Radiation Levels/ Radiological Effluent AUl ............................................................................................................................................... 43 AUl .......................... ;.................................................................................................................... 44 AUl ............................................................................................................................................... 45 AU2 ..................................... ;......................................................................................................... 46 AU2 ............................................................................................................................................... 47 AAl ....................................................... ,........................................................................................ 48 AAl ............................................................................................................................................... 49 AAl ............................................................................................................................................... 50 AA2 ............................................ :.................................................................................................. 51 AA2 ............................................................................................................................................... 52 AA3 ............................................................................................................................................... 53 ASl ............................................................................................................................................... 54 AS1 ............................................................................................................................................... 55 AGl ............................................................................................................................................... 56.

1 AGl ............................................................................................................................................... 57 FUl ............................................................................................................................................... 58 FAl ............................................................................................................................................... 59 FSl ************************************************************************************************************************************************ 60 FGl ............................................................................................................................................... 61 PCl ................................................................................................................................................ 62 PC2 ................................................................................................................................................ 64 PC3 ................................................................................................................................................ 65 PC4 ................................................................................................................................................ 67 PCS ................................................................................................................................................ 68 FCl ................................................................................................................................................ 69 FC2 ................................................................................................................................................ 70 FC3 ................................................................................................................................................ 71 FC4 ................................................................................................................................................ 72 RCl ............................................................................................................................................... 73

References:

........... ,..................................................................................................................... 73 RC2 ............................................................................................................................................... 74 RC3 ............................................................................................................................................... 75 RC4 ............................................................................................................................................... 78

'EIP-2-001 REV-028 PAGE 36 OF 144

REFERENCE USE ATTACHM ENTS PAGE2 OF 107 EAL BASES RC5 ................................................................................................................................................ 79 HUI ...............................: ............................................................................................................... 80 HU4 ............................................................................................................................................... 83 HU5 ............................................................................................................................................... 85 HU6 ................................................................................................................................................ 86 HAI ............................................................................................................................................... 89 HAI ............................................................................................................................................... 90 HA3 ............................................................................................................................................... 92 HA4 ............................................................................................................................................... 93 HA6 ............................................................................................................................................... 96 HSI ............................................................................................................................................. 100 HS3 ............................................................................................................................................. 102 HGI ............................................................................................................................................. 103 SUI ............................................................................................................................................. 105 SU7 ............................................................................................................................................. 108 SU8 ......................................................................................................................... : ................... 109 SU9 ............................................................................................................................................. 1IO SU9 ............................................................................................................................................. 111 SUIO ........................................................................................................................................... 112 SUII ........................................................................................................................................... 113 SA6 ............................................................................................................................................. I16 SA6 ............................................................................................................................................. 117 SSI .............................................................................................................................................. 118 SS3 .............................................................................................................................................. 119 SS4 .............................................................................................................................................. 120 SS6 .............................................................................................................................................. 121 SS6 .............................................................................................................................................. 122 SGI ............................................................................................................................................. 123 SGI ................................. ,........................................................................................................... 124 SG3 ............................................................................................................................................. 125 CUI ............................................................................................................................................. 126 CU2 ............................................................................................................................................. 127 CU3 ............................................................................................................................................. 129 CU5 ............................................................................................................................................. 130 CU6 ............................................................................................................................................. 131 CU7 ............................................................................................................................................. 132 CU8 ............................................................................................................................................. 133 CAI ............................................................................................................................................. 134 CA3 ............................................................................................................................................. 135 CA3 ............................................................................................................................................. 136 CA5 ..........................................................................................................*................................... 137 CGl ............................................................................................................................................. 140 CGl ............................................................................................................................................. 141 E-HUl ................................................................................................................................ :-........ 142 I

I EIP-2-001 REV - 028 PAGE 37 OF 144 I

REFERENCE USE ATTACHMENTS PAGE3 OF107 EAL BASES Introduction and Background Information General Notes on Basis Document Use This document provides an explanation and rationale for each Emergency Action Level (EAL) '

included in the RBS EAL scheme based on NEI 99-01 Revision 5. It should be used to facilitate review of the RBS EALs, provide historical documentation for future reference and serve as a resource for training. Decision makers responsible for implementation of EIP-2-001, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present.

Use of this document for assistance is not intended to delay the emergency classification.

Emergency Classification Level Thresholds The most common bases for establishing these boundaries are the technical specifications and setpoints that have been developed in the design basis calculations and the Updated Safety Analysis Report (USAR).

For those conditions that are easily measurable and instrumented, the boundary is likely to be the EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a particular emergency classification level.

In addition to the continuously measurable indicators, such as coolant temperature, coolant levels, leak rates, containment pressure, etc., the USAR provides indications of the consequences associated with design basis events. Examples include steam pipe breaks, MSIV malfunctions, and other anticipated events.that, upon occurrence, place the plant immediately into an emergency classification level.

Another approach for defining these boundaries is the use of a plant specific probabilistic safety assessment (PSA - also known as probabilistic risk analysis, PRA). PSAs can be used as a good first approximation of the relevant I_Cs and risk associated with emergency conditions. RBS has an .

Individual Plant Evaluation (IPE) and an Individual Plant Evaluation for External Events (IPEEE).

Another critical element of the analysis to arrive at these threshold (boundary) conditions is the time that the plant might stay in that condition before moving to a higher emergency classification level. In particular, station blackout coping analyses performed in response to 10 CFR 50.63 and Regulatory Guide 1.155, "Station Blackout," are used to determine whether RBS enters a Site Area Emergency or a General Emergency directly, and when escalation to General Emergency is indicated. The time dimension is critical to the EAL since the purpose of the emergency classification level for state and local officials is to notify them of the level of mobilization that may be necessary to handle the emergency. This is particularly true when a Site EIP-2-001 REV -028 PAGE 38 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 4 OF 107 EAL BASES Introduction and Background Information Emergen cy Classification Level Thresholds (Cont'd)

Area Emergen cy or General Emergen cy is IMMINEN T. Establish ing EALs for such condition s must take estimate d evacuation time into consider ation to minimize the potential for the plume to pass while evacuatio n is underway.

Regardle ss of whether or not containm ent integrity is challenged, it is possible for significan t

radioacti ve inventory within containm ent to result in EPA PAG plume exposure levels being exceede d even assuming containment is within technica l specification allowable leakage rates. With or without containm ent challenge, however, a major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. NUREG- 1228, "Source Estimations During Incident Respons e to Severe Nuclear Power Plant Accident s," indicates that such condition s do not exist when the amount of clad damage is less than 20%.

Emergen cy Action Levels (EALs)

Planned evolutions involve preplann ing to address the limitations imposed by the condition

, the performa nce of required surveillance testing, and the impleme ntation of specific controls prior to knowingl y entering the condition in accordan ce with the specific requirem ents of the RBS Technica l

Specifica tions. Activities which cause the site to operate beyond that allowed by the Technica l

Specifica tions, planned or unplanned, may result in an EAL threshold being met or exceeded r .

Planned evolutions to test, manipula te, repair, perform maintena nce or modificat ions to systems and equipme nt that result in an EAL value being met or exceede d are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operation al limitation s imposed by the operating license. However, these condition s may be subject to the reporting requirements of 10 CFR 50.72.

All classifications are to be based upon valid indications, reports or conditions. Indications, reports or condition s are considered valid when they are verified by (1) an instrume nt channel check, or (2) indicatio ns on related or redundant indications, or (3) by direct observat ion by plant personne l, such that doubt related to the indication's operability, the condition 's existence , or the report's accuracy is removed . Implicit in this definition is the need for timely assessm ent.

With the emergen cy classification levels defined, the threshold s that must be met for each EAL to be placed under the emergen cy classification level can be determin ed. There are two basic approach es to determin ing these EALs. EALs and emergen cy classification level boundari es coincide for those continuo usly measurable, instrumented ICs, such as radioactivity, core temperature, coolant levels, etc. For these,ICs, the EAL is the threshold reading that most closely correspo nds to the emergen cy classifica tion level description using the best available information.

EIP-2-001 REV-02 8 PAGE 39 OF 144

REFERENCE USE ATTACHMENTS PAGE 5 OF 107 EAL BASES Introduction and Background Information Emergency Action Levels (EALs) (Cont'd)

For discrete (discontinuous) events, the approach is somewhat different. Typically, in this category are internal and external hazards such as FIRE or earthquake. The purpose for including hazards in EALs is to assure that RBS personnel and offsite emergency response organizations are prepared to deal with consequential damage these hazards may cause. If, indeed, hazards have caused damage to safety functions or fission product barriers, this should be confirmed by symptoms or by observation of such failures. Therefore, it may be appropriate to enter an Alert status for events approaching or exceeding design basis limits such as Operating Basis Earthquake (OBE), design basis wind loads, FIRE within VITAL AREAS, etc. This would give the operating staff additional support and improved ability to determine the extent of plant damage. If damage to barriers or challenges to Critical Safety Functions (CSFs) have occurred or are identified, then the additional support can be used to escalate or terminate the emergency classification level based on what has been found. Of course, security events must reflect potential for rising security threat levels.

Emergency Operating Procedures (EOP) are designed to maintain and/or restore a set of CSFs which are listed in the order of priority for restoration efforts-during accident conditions.

There are diverse and redundant plant systems to support each CSF. By monitoring the CSFs instead of the individual system component status, the impact of multiple events is inherently addressed (e.g., the number of operable components available to maintain the critical safety function.).

The EOPs contain detailed instructions regarding the monitoring of these functions and provides a scheme for classifying the significance of the challenge to the functions. In providing EALs based on these schemes, the emergency classification level can flow from the EP assessment rather than being based on a separate EAL assessment. This is desirable as it reduces ambiguity and the time necessary to classify the event.

Although the majority of the EALs provide very specific thresholds, 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. While this is particularly prudent at the higher emergency classification levels (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classification levels.

EIP-2-001 REV - 028 PAGE 40 OF 144

REFERENCE USE ATTACHM ENTS PAGE 6 OF 107 EAL BASES Introduction and Background Information Treatmen t of Multiple Events and Classification Level Upgrading The above discussion deals primarily with simpler emergencies and events that may not escalate rapidly. However, usable EAL guidance must also consider rapidly evolving and complex events.

Hence, emergency classification level upgrading and consideration of multiple events must be addressed.

When multiple simultaneous events occur, the emergency classification level is based on the highest EAL reached. For example, two Alerts remain in the Alert category. or, an Alert and a Site Area Emergenc y is a Site Area Emergency. Further guidance is provided in RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.

Emergenc y Classification Level Downgrading Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly lowering. RBS uses a combination approach involving recovery (generally for higher classifications)and termination (for lower classifications. Downgrading to lower emergenc y classification levels is not used at RBS.

Classifying Transient Events For some events, the condition may be corrected before a declaration has been made. The key consideration in this situation is to determine whether or not further plant damage occurred while the corrective actions were being taken. In some situations, this can be readily determined, in other situations, further analyses (e.g., coolant radiochemistry sampling, may be necessary). Classify the event as indicated and termina~e the emergenc y once assessment shows that there were no consequences from the event and other termination criteria are met.

Existing guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate Operator actions.

There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no long~r exists. In these cases, an emergency should not be declared.

Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, should be applied.

EIP-2-001 REV-028 PAGE 41 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 70F 107 EAL BASES Introduction and Background Information Operating Mode Applicability The plant operating mode that existed at the time that the event occurred, prior to any protective system or operator action initiated in response to the condition, is compared to the mode applicability of the EALs. If an everit occurs, and a low~r or higher plant operating mode is reached before the emergency classification level can be declared, the emergency classification level shall be based on the mode that existed at the time the event occurred.

For events that occur in Cold Shutdown or Refueling, escalation is via EALs that have Cold Shutdown or Refueling for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

Pl an tO oerat*ma M od e Usaae f or RBS EALs:

1 MODE TITLE REACTOR MODE AVERAGE SWITH POSITION REACTOR COOLANT TEMPERATl URE (OF) 1 Power Operation Run N/A 2 Startup Refuel (a) or NIA Startup/Hot Standby 3 Hot ShutdownCaJ Shutdown >.200 4 Cold ShutdownCal Shutdown ~200 5 RefuelingCbl Shutdown or Refuel N/A (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned.

Defueled (D) - All reactor fuel removed from reactor pressure vessel (full core offload during refueling or extended outage). This is not an operating mode designation by Technical Specifications.

EIP-2-001 REV -028 PAGE 42 OF 144

REFERENCE USE ATTACH MENT 8 PAGE8 0Fl07 EAL BASES Initiating Condition - NOTIFICA TION OF UNUSUA L EVENT AUl Any release of gaseous or liquid radioactivity to the environment > 2 times the ODCM limit for~

60 minutes Operating Mode Applicability: All Emergen cy Action Level(s): (1 or 2 or 3)

Note: The Emergen cy 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.

1. VALID reading on any of the radiation monitors in Table R1 > the NOUE reading for~ 60 minutes

2. VALID reading on RMS-RE107 effluent monitor> 2 times the alarm setpoint established by a current radioactivity discharge permit for~ 60 minutes OR

3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates>

2 times the ODCM limit for~ 60 minutes 3.06E+05 µCi/sec 5.26E-03 µCi/ml Fuel Building Vent Primary 4GE005 2.19E+04 µCi/sec Secondary 5GE005 , 4.65E-03 Ci/ml Radwaste B'\,Jilding Vent Primary 4GE006 2.58E+04 µCi/sec Secondary 5GE006 6.84E-04 Ci/ml EIP-2-001 REV -028 PAGE 43 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 90F 107 EAL BASES AUl 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 IC addresses a potential reduction in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

RBS incorporates 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 (Offsite Dose Calculation Manual - ODCM). 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 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offslte 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.

Releases should not be prorated or averaged over 60 minutes. For example, a release exceeding 4 X the ODCM limit for 30 minutes does not meet the threshold for this IC.

This Initiating Condition 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.

EAL#1 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the Initiating Condition.

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.

Any release on the routine effluent monitors in excess of the TRM limit is considered a non-routine release. Table R1 provides the monitors' EAL setpoint values. Values are provided for a primary and secondary source for NOUE and Alert EAL determination. The Division I safety related monitors (ORMS 4GE125 and 4GE005) are the preferred source for main plant exhaust and fuel building EAL determination. Radwaste building preferred value is the effluent monitor (4GE006). The secondary monitors in Table R1 should be used to determine EALs if the preferred monitors are inoperable.

EAL#2 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in this Initiating Condition established by the radioactivity discharge permit. This value is associated with a planned batch release.

EIP-2-001 REV- 028 PAGE 44 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 10 OF 107 EAL BASES AUl EAL#3 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.

References:

T.R. 3.11 RSP-0008, Offsite Dose Calculation Manual (ODCM)

G.13.18:9.6*012 Rev 0, Effect of Core Uprate on the ORMS Process Safety Limit I Conversion Factors I PR-C-495 Rev 2 p 4 ESK-RMS05 ESK-RMS25 EIP-2-001 REV -028 PAGE 45 OF 144

REFERENCE USE ATTACHMENTS PAGE 11 OF 107 EAL BASES AU2 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED rise in plant radiation levels Operating Mode Applicability: All Emergency Action Level(s): (1 or 2)

1. a. UNPLANNED water level drop in a reactor refueling pathway as indicated by any of the following: *

• Water level drop in the reactor refueling cavity, spent fuel pool or fuel transfer canal indication on Control Room Panel 870

• . Personnel observation by visual or remote means

b. UNPLANNED VALID area radiation monitor alarm on any of the following:.

RMS-RE140 RMS-RE141 RMS-RE192 RMS-RE193 OR

2. UNPLANNED VALID area radiation monitor readings or survey results indicate a rise by a factor of 1000 over normal* levels NOTE: For area radiation monitors with ranges incapable of measuring 1000 times normal* levels, classification shall be based on VALID full scale indications unless surveys confirm that area radiation levels are below 1000 times normal* within 15 minutes of the area radiation monitor indications going full scale.

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

Basis:

This IC addresses elevated radiation levels as a result of a water level drop above irradiated .fuel or events that have resulted, or may result, in UNPLANNED rises in radiation dose rates within plant buildings. These radiation rises represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.

EIP-2-001 REV -028 PAGE 46 OF 144

REFEREN CE USE ATTACHM ENTS PAGE 12 OF 107 EAL BASES AU2 EAL#1 The locations of the EAL specific area radiation monitors are:

Containment RMS-RE140, North Refueling Floor RMS-RE141, South Refueling Floor Fuel Building RMS-RE192, South Operating Floor RMS-RE193, North Operating Floor The refueling pathway is a site specific combination of cavities, tubes, canals and pools. While a radiation monitor could detect a rise 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.

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.

This event escalates to an Alert per AA2 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 Matrix for events in operating modes 1-3.

EAL#2 This EAL addresses rises 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 rises 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.

References:

EIP-2-001 REV-028 PAGE 47 OF 144

REFERENCE USE ATTACHMENTS PAGE 13 OF 107 EAL BASES AAl Initiating Condition - ALERT Any release of gaseous or liquid rardioactivity to the environment > 200 times the ODCM limit for:::_ 15 minutes Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3)

Note: 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.

1. VALi D reading on any of the radiation monitors in Table R 1 > the ALERT reading for :::. 15 minutes

2. For RMS-RE107 effluent monitor:

EITHER VALID reading> 200 times the alarm setpoint established by a current radioactivity discharge permit for

~ 15 minutes VALi D reading > 1.27E-01 µCi/ml for~ 15 minutes OR

3. Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates >

200 times the ODCM limit for:::_ 15 minutes EIP-2-001 REV -028 PAGE 48 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 14 OF 107 EAL BASES AAl Tab'le.R1 ~;' <, . <

J:AL'TH~,l;:SfiOLD *

• Metl:16d Al;E~T

• [)RMS Grid 6

• fhreshol~l Main Plant Vent Primary 4GE125 3.06E+07 µCi/sec Seconda 1GE126 2.82E-01 µCi/ml Fuel Building Vent Primary 4GE005 2.19E+06 µCi/sec Secondary 5GE005 2.82E-01 Ci/ml Radwaste Building Vent Primary 4GE006 2.58E+06 µCi/sec Secondary 5GE006 6.84E-02 Ci/ml 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 IC addresses an actual or substantial potential reduction in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

RBS incorporates features intended to control the release of radioactive effluents to the environme nt.

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 and AA1 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.

Releases should not be prorated or averaged. For example, a release exceeding 600 times the ODCM limit for 5 minutes does not meet the threshold for this IC.

This Initiating Condition 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.

EAL#1 This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the Initiating Condition.

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.

EIP-2-001 REV-02 8 PAGE 49 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 15 OF 107 I EALBASES AAI Any release on the routine effluent monitors in excess of the TRM limit is considered a non-routine release. Table R1 provides the monitors' EAL setpoint values. Values are provided for a primary and secondary source for NOUE and Alert EAL determination. The Division I safety related monitors (ORMS 4GE125 and 4GE005) are the preferred source for main plant exhaust and fuel building EAL determination. Radwaste building preferred value is the effluent monitor (4GE006). The secondary monitors in Table R1 should be used to determine EALs if the preferred monitors are inoperable.

EAL#2 This EAL addresses radioactivity releases, that for whc:1tever reason, cause effluent radiation monitor readings to exceed the threshold identified in this Initiating Condition established by the radioactivity discharge permit. This value i~ associated with a planned batch release.

Historical release permits indicate that the Alert value of 200 times the radiation monitor setpoint established by the current permit may exceed the operating range of the RMS-RE107 effluent monitor in some instances. This potentially affected monitor is listed in EAL #2 with a corresponding value for the top of its indicating range.

EAL#3 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.

References:

T.R. 3.11 RSP-0008, Offsite Dose Calculation Manual (ODCM)

G.13.18.9.6*012 Rev 0, Effect of Core Uprate on the ORMS Process Safety Limit I Conversion Factors I PR-C-495 Rev 2 p 4 ESK-RMS05 ESK-RMS25 EIP-2-001 REV -028 PAGE 50 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 16 OF 107 EAL BASES AA2 Initiating Condition - ALERT Damage to irradiated fuel or loss of water level that has resulted or will result in the uncovering of irradiated fuel outside the reactor vessel Operating Mode Applicability: All Emergency ~ction Level(s): (1 or 2)

1. A water level drop in the reactor refueling cavity, spent fuel pool or fuel transfer canal that will result in irradiated fuel becoming uncovered OR

2. A VALID reading on any of the following radiation monitors due to damage to irradiated fuel or loss of water level:

RMS-RE140 2000 mR/hr RMS-RE141 2000 mR/hr RMS-RE192 2000 mR/hr RMS-RE193 2000 mR/hr RMS-RE5A 1.64E+03 µCi/sec RMS-RE5B (GE) 5.29E-04 µCi/ml Basis:

This IC addresses rises 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 in that fuel activity has been released, or is anticipated due to fuel heatup. This IC applies to spent fuel requiring water coverage and is not intended to address spent fuel which is licensed for dry storage.

The locations of the EAL specific area radiation monitors are:

Containment RMS-RE140 North Refueling Floor RMS-RE141 South Refueling Floor Fuel Building RMS-RE192 South Operating Floor RMS-RE193 North Operating Floor RMS-RE?A (B) Fuel Building Ventilation Exhaust EIP-2-001 REV- 028 PAGE 51 OF 144

REFERENCE USE ATTACHMENTS PAGE 17 OF 107 EAL BASES AA2 EAL#1 Indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. Depending on available level indication, the declaration may be based on indications of water makeup rate or decrease in Refueling Water Storage Pool level. Video cameras (Security or outage-related) may allow remote observation of level.

EAL#2 This EAL addresses radiation monitor indications of fuel uncovery and/or fuel damage.

Elevated ventilation monitor readings may be an indication of a radioactivity release from the fuel, confirming that damage has occurred. Elevated background at the ventilation monitor due to water level drop may mask elevated ventilation exhaust airborne activity and needs to be considered.

While a radiation monitor could detect a rise 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.

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 a:s removal of the reactor head. Generally, elevated radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss.

The Abnormal Operating Procedure (AOP) provides a table for guidance on pool level and of potential scenarios and the expected pool level assuming no operator action. The AOP is also entered for UNPLANNED lowering of refueling cavity or lower fuel pool wat~r level during refueling operations.

When control rod blades are stored in the Spent Fuel Pool, dose rate rise in the area may be attributed to the stored items instead of uncovered fuel assemblies.

Escalation of this emergency classification level, if appropriate, would be based on AS1 or AG1.

References:

TS Table 3.3.6.2-1 Calculation 813.18.9.4*10 EIP-2-001 REV -028 PAGE 52 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 18 OF 107 EAL BASES AA3 Initiating Condition - ALERT Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions Operating Mode Applicability: All Emergency Action Level(s):

Dose rate > 15 mR/hr in any of the following areas requiring continuous occupancy to maintain plant safety functions:

Main Control Room CAS Basis:

This IC addresses elevated 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 rise in radiation levels is not a concern of this IC. The Emergency Director must consider the source or cause of the elevated radiation levels and determine if any other IC may be involved.

This IC is not meant to apply to increases in the containment dome radiation monitors as these are events which are addressed in the fission product barrier matrix EALs.

RP surveys should be performed in the CAS area if radiation above the program limit is detected outside the RCA. The Control Room area radiation monitor should be observed for EAL conditions if rising radiation levels are detected outside the RCA.

The Main Control Room and CAS are the areas at RBS requiring continuous occupancy.

References:

EIP-2-001 REV- 028 PAGE 53 OF 144'

REFERENCE USE ATTACHMENT 8 PAGE 19 OF 107 EAL BASES AS1 Initiating Condition -- SITE AREA EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity

> 100 mR TEDE or 500 mR thyroid COE for the actual or projected duration of the release Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3)

Note: 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, the classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.

1. VALID reading on any of the radiation monitors in Table R1 > the SITE AREA EMERGENCY reading for~ 15 minutes

2. Dose assessment using actual meteorology indicates doses > 100 mR TEDE or 500 mR thyroid COE at or beyond the SITE BOUNDARY

3. Field survey results indicate closed window dose rates > 100 mR/hr expected to continue for:::. 60 minutes; or analyses of field survey samples indicate thyroid COE > 500 mR for one hour of inhalation, at or beyond the SITE BOUNDARY l

Main Plant Vent Primary 4GE125 4.50E+07 µCi/sec Secondary N/A Fuel Building Vent Primary 4GE005 1.00E+08 µCi/sec Secondary N/A Radwaste Building Vent N/A EIP-2-001 REV-028 PAGE 54 OF 144

REFERENCE USE ATTACHMENTS PAGE 20 OF 107 EAL BASES ASI Basis:

This IC 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.

EAL#1 The monitor list in EAL #1 includes monitors on all potential release pathways.

EAL#2 Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor readings in EAL #1 are 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 should 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 EALs.

References:

EIP-2-001 REV -028 PAGE 55 OF 144

REFERENCE USE ATTACHMENT 8

• PAGE 21 OF 107 EAL BASES AGl Initiating Condition -- GENERAL EMERGENCY Offsite dose resulting from an actual or IMMINENT release of gaseous radioactivity

> 1000 mR TEDE or 5000 mR thyroid COE for the actual or projected duration of the release using actual meteorology Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3)

Note: 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, the classification should be based on EAL #2 instead of EAL #1. Do not delay declaration awaiting dose assessment results.

1. VALID reading on any of the radiation monitors in Table R1 > the GENERAL EMERGENCY reading for~ 15 minutes

2. Dose assessment using actual meteorology indicates doses > 1000 mR TEDE or 5000 mR thyroid COE at or beyond the SITE BOUNDARY

3. Field survey results indicate closed window dose rates > 1000 mR/hr expected to continue for:: 60 minutes; or analyses offield survey samples indicate thyroid COE > 5000 mR for one hour of inhalation, at or beyond the SITE BOUNDARY Main Plant Vent Primary 4GE125 4.50E+08 µCi/sec Secondary N/A Fuel Building Vent Primary 4GE005 1.00E+09 µCi/sec Secondary NIA Radwaste Building Vent NIA EIP-2-001 REV-028 PAGE 56 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 22 OF 107 EAL BASES Basis:

AGI This IC 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.

EAL#1 The monitor list in EAL #1 includes monitors on all potential release pathways.

EAL#2 Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor readings in EAL #1 are not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures should 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 EALs.

References:

EIP-2-001 REV-028 PAGE 57 OF 144

REFERENCE USE ATTACHMENTS PAGE 23 OF 107 EAL BASES FUl INITIATING CONDITION- NOTIFICATION OF UNUSUAL EVENT ANY loss or ANY potential loss of containment Operating Mode Applicability: Mode 1 Power Operation Mode2 . Startup I Mode 3 Hot Shutdown Emergency Action Level(s):

1. Any loss or any potential loss of containment Bases:

Comparison of conditions / values with those listed in Fission Product Barrier Matrix indicates a loss or any potential loss of containment.

The Fuel Cladding (FC) and the Reactor Coolant System (RCS) are weighted more heavily than the Primary Containment (PC) barrier. NOUE ICs associated with RCS and FC barriers are addressed under System Malfunction ICs.

Loss of containment would be a potential degradation in the level of plant safety. The PC barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment barrier thresholds are used primarily as discriminators for escalation from an Alert to a *site Area Emergency or a General Emergency.

EIP-2-001 REV - 028 PAGE 58 OF 144

REFERENCE USE ATTACHMENT 8

/ PAGE 24 OF 107 EAL BASES FAl INITIATING CONDITION -ALERT Any loss or any potential loss of either fuel clad or RCS Operating Mode Applicability: Mode 1 Power Operation Mode 2 Startup Mode 3 Hot Shutdown Emergency Action Level(s):

1. Any loss or any potential loss of fuel clad Any loss or any potential loss of RCS J Bases:

Comparison of conditions / values with those listed in Fission Product Barrier Matrix indicates a loss or potential loss of a Fuel Clad barrier or a loss or potential loss of the RCS barrier.

The Fuel Cladding and the Reactor Coolant System are weighted more heavily than the Primary Containment barrier.

Loss of either the Fuel Cladding or the Reactor Coolant System would be a substantial degradation in the level of plant safety.

The Fuel Clad barrier is the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets.

The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

EIP-2-001 REV-028 PAGE 59 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 25 OF 107 EAL BASES FS1 INITIATING C9NDITION-S1TE AREA EMERGENCY Loss or potential loss of any two barriers Operating Mode Applicability: Mode 1 Power Operation Mode 2 Startup Mode 3 Hot Shutdown Emergency Action Level(s):

1. Loss or potential loss of any two barriers Bases:

Comparison of conditions / values with those listed in Fission Product Barrier Matrix indicates loss or potential loss of any two barriers.

Loss of 2 Fission Product Barriers would be a major failure of plant systems needed for protection of the public.

EIP-2-001 REV-028 PAGE 60 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 26 OF 107 EAL BASES FGl INITIATING CONDITION-GENERAL EMERGENCY Loss of any two barriers and loss or potential loss c;>f third barrier Operating Mode Applicability: Mode 1 Power Operation Mode 2 Startup Mode 3 Hot Shutdown Emergency Action Level(s):

1. Loss of any two barriers Loss or potential loss of the third barrier Bases:

Comparison of conditions / values with those listed in Fission Product Barrier Matrix indicates a loss of any two barriers and the loss or potential loss of the third barrier.

Conditions / events required to cause the loss of 2 Fission Product Barriers with the potential loss of the third could reasonably be expected to cause a release beyond the immediate site area exceeding EPA Protective Action Guidelines.

EIP-2-001 REV- 028 PAGE 61 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 27 OF 107 EAL BASES PCI PRIMARY CONTAINMENT Emergency Action Level:

Primary containment conditions EAL threshold:

LOSS: ...... ........................ 1. Rapid unexplained loss of PC pressure following initial pressure rise

2. PC pressure response not consistent with LOCA conditions POTENTIAL LOSS: .............. 1. PC pressure > 15 psig and rising OR

2. a. PC hydrogen in the unsafe zone of HOOL curve OR

a. OW hydrogen concentration > 9%

OR

3. RPV pressure and suppression pool temperature cannot be maintained below the HCTL EIP-2-001 REV-028 PAGE 62 OF 144

__J

REFERENCE USE ATTACHM ENT 8 PAGE 28 OF 107 EAL BASES PCI Bases:

LOSS - Rapid unexplained loss of pressure (i.e., not attributable to condensation effects or restoration of containment or drywell unit coolers) following an initial pressure rise from a high energy line break indicates a loss of containment integrity. Primary containment pressure should rise as a result of mass and energy released into containment from a LOCA. Thus, primary containment pressure not rising 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. Control room indicators may include ERIS data points, P808 CMS indication, or back-panel CMS pressure indication.

POTENTIAL LOSS - The site specific pressure is based on the primary containment design pressure.

Primary Containment pressure greater than 15 psig and rising is based on the design pressure of the Primary Containment. If the Containment pressure is exceeded, this represents a condition outside the analyzed conditions. This constitutes a potential loss of the Primary Containment barrier even if a failure to isolate has not occurred.

The Emergency Procedure Guidelines and Severe Accident Guidelines identify that deflagration could occur if containment hydrogen concentration reaches the HOOL or drywell hydrogen concentration reaches 9%. The deflagration of Hydrogen represents a potential loss of the primary containment.

Indication of actual hydrogen concentration fn the containment is affected by the environmental conditions (i.e., the presence of water vapor). The RBS hydrogen monitoring system removes water vapor from the sample before hydrogen concentration is measured and, thus, may provide readings that are higher_than the actual hydrogen concentration.

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise:

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, Suppression chamber pressure above PC pressure limit A, 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.

References:

1.-

EIP-2-001 REV - 028 PAGE 63 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 29 OF 107 EAL BASES PC2 PRIMARY CONTAINMENT Emergency Action Level:

Reactor vessel water level EAL Threshold:

LOSS: .................................. NONE POTENTIAL LOSS: .............. Entry into PC flooding procedures SAP-1 and SAP-2 Bases:

LOSS-NONE POTENTIAL LOSS - 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 SAP-1 and SAP-2 is a logical escalation in response to the inability to maintain adequate core cooling.

The condition in this potential loss threshold represents a potential core melt sequence which, if not corrected, could lead to vessel failure and higher 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.

References:

EIP-2-001 REV- 028 PAGE 64 OF 144

REFERENCE USE ATTACHMENT 8 PAGE.30 OF 107 EAL BASES PC3 PRIMARY CONTAINMENT Emergency Action Level:

Primary containment isolation failure or bypass EAL Threshold:

LOSS: ................................... .1. a. Failure of all valves in any one line to close

b. Direct downstream pathway to the environment exists after PC isolation signal

2. Intentional PC venting per EOPs or SAPs OR

3. UNISOLABLE RCS leakage outside PC as indicated by exceeding either of the following:

a. Max Safe Operating Temperijture (Table F1)

OR

b. Max Safe Area Radiation (Table F1)

POTENTIAL LOSS: .............. NONE

. *. / .

-~;

TABLE,F1

., '.' V

,, .. \,;. ., , * ,.. '

PC 3. Loss of Primary Containment .

'.*.<< ~ -- . * ,, - , - , *.

Parameter Area Temperature Area Radiation Level Max Safe 012erating Value ORMS Grid 2 Max Safe 012erating Value RHR A equipment area 200° F 1213 9.5E+03 mRlhr RHR B equipment area 200° F 1214 9.5E+03 mRlhr RHR C equipment area NIA 1215 9.5E+03 mRlhr RCIC room 200° F 1219 9.5E+03 mRlhr MSL Tunnel 200° F NIA RWCU pump room 1 (A) I 200° F NIA 2 (B)

EIP-2-001 REV -028 PAGE 65 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 31 OF 107 EAL BASES PC3 Bases:

These thresholds address incomplete containment isolation that allows direct release to the environment.

LOSS - Failure to isolate - Inability to isolate means the primary containment isolation valve(s) did not fully close after a VALID automatic or manual isolation signal and is not isolable from the Main Control Room, or an attempt for isolation from the Main Control Room has been made and was unsuccessful.

An attempt for isolation should be made upon identification and prior to the accident classification. If isolated from the Main Control Room upon identification, this INITIATING CONDITION is not applicable. Dispatch of Operators outside the Control Room for manual attempts to close the valve is not considered.

Primary Containment isolation valves are described in the Technical Specifications bases for Primary Containment, Primary Containment Airlock and Primary Containment Isolation Valves (T.S. 3.6.1.1 ).

The Cpntainment airlock is not considered in this EAL since airlock failure would be a potential failure mode to cause the EAL PC1 threshold.

The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems. 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.

Containment Venting - Site specific EOPs and SAPs 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 or SAPs 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.

Area temperature or radiation - The presence of area radiation or temperature Max Safe Operating setpoints indicating 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.

Leakage into a closed system is to be considered a loss of primary containment only if the closed system is breached and thereby 9reates a path to the environment.

POTENTIAL LOSS - None

References:

EIP-2-001 REV - 028 PAGE 66 OF 144

REFERENCE USE ATTACHMENTS PAGE 32 OF 107 EAL BASES PC4 PRIMARY CONTAINMENT Emergency Action Level:

Primary containment radiation monitors EAL Threshold:

LOSS: .................................. NONE POTENTIAL LOSS: ............. Containment radiation monitor RMS-RE16 reading> 10,000 R/hr BASIS LOSS-NONE POTENTIAL LOSS - The site specific reading 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.

References:

............ Calculation GI3.18.9.4-045 Rev. 0 EIP-2-001 REV -028 PAGE 67 OF 144

. REFERENCE USE ATTACHMENT 8 PAGE 33 OF 107 EAL BASES PC5 REACTOR COOLANT SYSTEM Emergency Action Level:

Emergency Director judgment EAL Threshold:

LOSS: .................................. Any condition in the opinion of the Emergency Director that indicates loss of the Primary Containment barrier POTENTIAL LOSS: .............. Any condition in the opinion of the Emergency Director that indicates potential loss of the Primary Containment barrier

• Bases:

LOSS or POTENTIAL LOSS - This EAL addresses any other factors that are to be used by the Emergency Di.rector in determining whether the primary containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be considered in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

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 Primary Containment barrier. When no event is in progress (loss or potential loss of either fuel clad and/or RCS) the Primary Containment barrier status is addressed by Technical Specifications.

References:

EIP-2-001 REV - 028 PAGE 68 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 34 OF 107 EAL BASES FCl FUEL CLAD Emergency Action Level:

Primary coolant activity level EAL Threshold:

LOSS: ................................... Coolant activity> 300 µCi/g dose equivalent 1-131 POTENTIAL LOSS: .............. NONE Bases:

LOSS-The site specific value is 300 µCi/gm dose equivalent 1-131. 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.

POTENTIAL LOSS-NONE

References:

EIP-2-001 REV -028 PAGE 69 OF 144 I

I_

REFERENCE USE ATTACHMENT 8 PAGE 35 OF 107 EAL BASES FC2 FUEL CLAD Emergency Action Level:

Reactor vessel water level EAL Threshold:

LOSS: .............................. RPV water level cannot be restored and maintained above

-187inches POTENTIAL LOSS: ......... .... RPV water level cannot be restored and maintained above -162 inches or cannot be determined Bases:

LOSS - This site specific 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.

Reactor vessel water level less than the minimum steam cooling RPV water level (-187") with injection is the lowest level with adequate core cooling to maintain peak clad temperature less than 1500°F where fuel clad damage (fuel rod perforation) may begin. Corrective actions as described in the Emergency Operating Procedures (EOPs) and Severe Accident Guidelines (SAGs) will be needed to mitigate fuel clad/core damage.

POTENTIAL LOSS -This threshold is the same as the RCS barrier loss threshold RC2 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 the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. With Reactor vessel water level less than the top of active fuel (-162"), adequate core cooling is still assured but is sufficiently low that any further drop in water level could result in the significant degradation of the cladding. Corrective actions as described in the Emergency Operating Procedures (EOPs) will be needed to mitigate fuel clad/core damage.

References:

EIP-2-001 REV -028 PAGE 70 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 36 OF 107 EAL BASES FUEL CLAD FC3 Emergency Action Level:

Primary containment radiation monitors EAL Threshold:

LOSS: .................................. Containment radiation monitor RMS-RE16 reading

> 3,000 R/hr POTENTIAL LOSS: ............. NONE Bases:

LOSS - Containment radiation monitors reading in excess of 3000 R/hr after Reactor Shutdown are indicative of both the loss of the reactor coolant system and 5% clad failure with the instantaneous release and dispersal of the reactor coolant noble gas and Iodine invento~y into the drywell and containment atmosphere.

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.

POTENTIAL LOSS-NONE

References:

Calculation GB.18.9.4-045 Rev. 0 EIP-2-001 REV -028 PAGE 71 OF 144

REFERENCE USE ATTACHMENTS PAGE 37 OF 107 EAL BASES FC4 FUEL CLAD Emergency Action Level:

Emergency Director judgment EAL Threshold:

LOSS: .................................. Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier

\ POTENTIAL LOSS: .............. Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Bases:

LOSS or POTENTIAL LOSS - This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be considered in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

References:

EIP-2-001 REV - 028 PAGE 72 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 38 OF 10}

EAL BASES RCl REACTOR COOLANT SYSTEM Emergency Action Level:

Drywell pressure EAL Threshold:

LOSS: ................................... Drywell pressure > 1.68 psid with indications of reactor coolant leak in drywell

• POTENTIAL LOSS: ............. NONE Bases:

LOSS - The site specific primary containment pressure is based on the drywell high pressure set point which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

Pressure rise due solely to loss of containment or drywell heat removal capability, testing, etc are not considered for this EAL threshold.

POTENTIAL LOSS - NONE.

References:

EIP-2-001 REV-028 PAGE 73 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 39 OF 107 EAL BASES RC2 REACTOR COOLANT SYSTEM Emergency Action Level:

Reactor vessel water level EAL Threshold:

LOSS: .................................... RPV water level cannot be restored and maintained above -162 inches or cannot be determined POTENTIAL LOSS: ............... NONE Bases:

LOSS - The loss EAL threshold of site specific RPV water level corresponds to the level that is used in EOPs to indicate challenge of core cooling.

This threshold is the same as the Fuel Clad barrier potential loss EAL threshold FC2 and corresponds to a challenge to core cooling. Thus, this threshold indicates a loss of the RCS barrier and potential loss of the Fuel Clad barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

POTENTIAL LOSS- NONE

References:

EIP-2-001 REV -028 PAGE 74 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 40 OF 107 EAL BASES RC3 REACTOR COOLANT SYSTEM Emergency Action Level:

RCS leak rate EAL Threshold:

LOSS: ........................ 1. UNISOLABLE main steam line break as indicated by the failure of both MSIVs in any one line to close High MSL flow annunciator (P601-19A-A2)

2. mdication of an UNISOLABLE HPCS, Feedwater, RWCU or RCIC break OR

3. Emergency RPV depressurization is required POTENTIAL LOSS: ...... 1. RCS leakage> 50 gpm inside the drywe/1 OR

2. UNISOLABLE RCS leakage outside PC as indicated by exceeding either of the following:

b. Max Normal Operating Temperature (Table F2)

OR

b. Max Normal Area Radiation (Table F2)

EIP-2-001 REV -028 PAGE 75 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 41 OF 107 EAL BASES RC3 TABLEF2 ,. *,

C "a .,

v~

• RC 3'.;t>Qtential Loss of RCS:.' ,,

.f Parameter Area Temperature Area Radiation Level (isolation temQerature DRMS Grid 2 Max Normal OQerating alarm) Value RHR A equipment area 117° F 1213 8.2E+Ol mR/hr (P601-20A-B4)

RHR B equipment area 117° F 1214 8.2E+Ol mR/hr (P601-20A-B4)

RHR C equipment area NIA 1215 8.2E+Ol mR/hr RCIC room 182° F 1219 l.20E+02 mR/hr (P601-21A-B6)

MSLTunnel 173°F NIA (P60l-l9A-All A3/Bl/B3)

RWCU pump room 1 (A)/ 2 (B) 165° F NIA (P680- l A-A2/B2)

Bases:

LOSS - An UNISOLABLE MSL break is a breach of the RCS barrier. Thus, this EAL threshold is included for consistency with the Alert emergency classification level.

Other large high-energy line breaks such as HPCS, Feedwater, RWCU, or RCIC that are UNISOLABLE also represent a significant loss of the RCS barrier and should be considered as MSL breaks for purposes of classification.

The leak is NOT isolable from the Main Control Room OR an attempt for isolation from the Main Control Room panels has been made and was not successful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification, this INITIATING CONDITION is not applicable.

Dispatch of operators outside the Control Room for manual attempts to close the valve is not considered.

• 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.

EIP-2-001 REV -028 PAGE 76 OF 144

REFEREN CE USE ATTACHM ENTS PAGE 42 OF 107 EAL BASES RC3 POTENTIAL LOSS - This threshold is based on leakage 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 the leak detection system leak rate information is unavailable (i.e., LOCA isolation, loss of power),

other indicators of RCS leakage should be used. Other indications include a rise in drywell temperature and pressure and a rise in the drywell radiation monitors. If the leakage computer is unavailable, sump level and pump status may help determine if the leakage is greater than 50 gpm.

If the DFR discharge line containment isolation valves have not isolated and a pump is running continuously without lowering sump level, the leakage may be assumed to exceed 50 gpm. The second pump can be started to verify that the first pump is not degraded. It is not intended to conclude a potential loss of the RCS barrier exists if both pumps are degraded and the observed leak rate as noted by rate of rise of level in the sump or calculated by the computer is such that it clearly confirms leakage below 50 gpm.

References:

EIP-2-001 REV -028 PAGE 77 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 43 OF 107 EAL BASES RC4 REACTOR COOLANT SYSTEM Emergency Action Level:

Drywell radiation EAL Threshold:

LOSS: .................................. Drywell radiation monitor RMS-RE20 reading > 100 R/hr POTENTIAL LOSS: ............. NONE Bases:

NOTE: Under post-LOCA conditions coaxial cables used on the drywell post accident monitors (RMS-RE20A/B) are susceptible to Thermally Induced Currents (TIC). These currents may cause the drywell PAMs to read falsely high (-469 R/hr) on a rapid temperature increase and read falsely low on a rapid temperature decrease. When accident temperature conditions stabilize indicated radiation dose rates would be more accurate. The duration of the spurious signal would last approximately 15 minutes.

During the period of false readings operators should rely on other indications of RCS leakage including a rise in drywell temperature and pressure.

LOSS - The site specific reading is a value which indicates the release of reactor coolant to the drywell.

This reading is less than that specified for Fuel Clad barrier Loss EAL threshold FC4. Thus, this threshold would be indicative of a RCS leak only. If the radiation monitor reading rose to that value specified by the Fuel Clad Barrier EAL threshold, fuel damage would also be indicated.

POTENTIAL LOSS-NONE

References:

813.18.9.4-051 NRC Information Notice IN 97-45 EIP-2-001 REV -028 PAGE 78 OF 144 J

REFERENCE USE ATTACHMENT 8 PAGE 44 OF 107 EAL BASES RCS REACTOR COOLANT SYSTEM Emergency Action Level:

Emergency Director judgment EAL Threshold:

LOSS: .................................. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier POTENTIAL LOSS: ........ ...... Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Bases:

LOSS or POTENTIAL LOSS - This EAL addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be considered in this EAL as a factor in Emergency Director judgment that the barrier may be considered lost or potentially lost.

References:

EIP-2-001 REV -028 PAGE 79 OF 144 l_

REFERENCE USE ATTACHMENTS PAGE 45 OF 107 EAL BASES

~

HUI Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed SECURITY CONDITION or threat which indicates a potential degradation in the level of safety of the plant Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3)

1. A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the RBS security shift supervision

2. A credible site specific security threat notification _

OR

3. A validated notification from NRC providing information of an aircraft threat 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 10 CFR 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under HA1, HS1 and HG1.

I A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. Consideration shall be given to upgrading the emergency response status and emergency classification in accordance with the Safeguards Contingency Plan and Emergency Plan.

EAL#1 The Security Shift Supervisor is the designated individual 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 EAL is based on the Safeguards Contingency Plan . The Safeguards Contingency Plan is based on guidance provided in NEI 03-12.

EAL#2 This EAL 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 Notification of Unusual Event.

EIP-2-001 REV - 028 PAGE 80 OF 144

REFERENCE USE ATTACHME NT 8 PAGE 46 OF 107 EAL BASES HUI The determination of "credible" is made through use of information found in the Safeguards Contingency Plan.

EAL#3 The intent of this EAL is to ensure that notifications for the aircra,ft 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 which the specific threat is made need declare the Unusual Event.

The NRC 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 NRC.

Escalation to Alert via HA 1 would be appropriate if the threat involves an airliner within 30 minutes of the plant. *

References:

NEI 03-12 EIP-2-001 REV-028 PAGE 81 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 47 OF 107 EAL BASES HU2 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other conditions exist which in the judgment of the Emergency Director warrant declaration of a NOUE Operating Mode Applicability: All Emergency Action Level(s):

1. Other conditions exist which 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 protectiofl has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

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 NOUE emergency classification level.

References:

EIP-2-001 REV -028 PAGE 82 OF 144 I

-J

REFEREN CE USE ATTACHM ENT 8 PAGE 48 OF 107 EAL BASES HU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT FIRE within the PROTECTED AREA not extinguished within 15 minutes of detection or EXPLOSION within the PROTECTED AREA Operating Mode Applicability: All Emergency Action Level(s): (1 or 2)

Note: 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 duration has exceeded, or will likely exceed, the applicable time.

1. FIRE not extinguished within 15 minutes of Control Room notification or verification of a Control Room FIRE alarm in any Table H2 structure or area.

OR

2. EXPLOSION within the PROTECTE D AREA

. ',;:,,,,~ ,}Cs'.'i:f:abl~ 01+2,,(,?) ;,*.,t\t\~!(f::;*

tj~i~;i~:,i:~iit!1Jlfl~~it!l;!t!t,:

Reactor Building Standby Cooling Tower Auxiliary Building Diesel Generator Building Control Building Tunnels (8, D, E, F, G)

Fuel Building

• Basis:

This IC addresses the magnitude and extent of FIRES or EXPLOSIO NS that may be potentially significant precursors of damage to safety systems. It addresses the FIRE/ EXPLOSION, 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.

EIP-2-001 REV-028 PAGE 83 OF 144

REFERENCE USE ATTACHMENTS PAGE 49 OF 107 EAL BASES HU4 EAL#1 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.

I 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).

EAL#2 This EAL addresses only those EXPLOSIONS of sufficient force to damage permanent structures or equipment within the PROTECTJ=D AREA.

No attempt is made to assess the actual magnitude of the damage. The occurrence of the EXPLOSION is sufficient for declaration.

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 HA4.

References:

EIP-2-001 REV - 028 PAGE 84 OF 144

REFERENCE USE ATTACHM ENTS PAGE 50 OF 107 EAL BASES HU5 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of toxic, corrosive, asphyxiant, or flammable gases deemed detrimental to NORMAL PLANT OPERATIONS Operating Mode Applicability: All Emergency Action Level(s): (1 or 2)

1. Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect NORMAL PLANT OPERATIONS

2. Report by West Feliciana Parish for evacuation or sheltering of site personnel based on an off-site event Basis:

I ,

This IC is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect NORMAL PLANT OPERATIONS.

The fact that SCBAs may be worn does not eliminate the need to declare the event.

This IC 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 HA5.

References:

EIP-2-001 REV-028 PAGE 85 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 51 OF 107 EAL BASES HU6 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Natural or destructive phenomena affecting the PROTECTED AREA Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3 or 4 or 5)

1. Seismic event identified by any 2 of the following:

• Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A-D06)

b. Event Indicator on ERS-NBR3D TRIGGER (RECORD START) is yellow

• Earthquake felt in plant

• National Earthquake Center OR

2. Tornado striking within PROTECTED AREA boundary OR

3. Internal flooding that has the potential to affect safety related equipment required by Technical Specifications for the current operating mode in any Table H1 area OR

4. Turbine failure resulting in casing penetration or damage to turbine or generator seals OR

5. Severe weather or hurricane conditions with indication of SUSTAINED high winds

74 mph within the PROTECTED AREA boundary EIP-2-001 REV-028 PAGE 86 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 52 OF 107 EAL BASES HU6 Affected Location/ Max Safe Operating Value/ Indicator Parameter Aux Bldg Crescent Area 6 inches above floor 70' EL (must be verified locally)

HPCS Room 70'EL 4 inches above floor (P870-51A-G4)

RHR A Room 70'EL 4 inches above floor (P870-51A-G4)

RHR B Room 70'EL 4 inches above floor (P870-51A-G4)

RHR C Room 70'EL 4 inches above floor (P870-51A-G4)

LPCS Room 70'EL 4 inches above floor (P870-51A-G4)

RCIC Room 70'EL 4 inches above floor (P870-51A-G4)

Basis:

These EALs are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

EAL#1 Damage may be caused to some portions of the site, but should not affect ability of safety functions to operate.

A "felt earthquake" is an earthquake of sufficient intensity such that 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.

The annunciators "SEISMIC SYS RECORDING / TROUBLE" and the "yellow event indicator are listed in the Alarm Response Procedure as verification of an earthquake event.

The National Earthquake Center can confirm if an earthquake has occurred in the area of the plant.

EAL#2 This EAL is based on a tornado striking (touching down) 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 HA6.

EIP-2-001 REV-028 PAGE 87 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 53 OF 107 EAL BASES HU6 EAL#3 This EAL addresse~ the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps.

The EAL is only applicable to areas in Table H1 areas that contain systems required for safe shutdown of the plant and that are not designed to be partially or fully submerged. The EAL is based on VALID indication that the area water level has reached the Maximum Safe Operating Values as identified in EOP-3. Exceeding the Maximum Safe Operating Value is interpreted as a potential degradation in the level of safety of the plant and is appropriately treated as an Unusual Event.

Escalation of this emergency classification level, if ~ppropriate, would be via HA6, or by other plant conditions.

EAL#4 This EAL addresses main turbine rotating compon~nt 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 HU4 and HU5.

This EAL is consistent with the definition of a NOUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment.

Escalation of this emergency classification level, if appropriate, would be to HA6 based on damage done by PROJECTILES generated by the failure or by the radiological releases. These latter events would be classified by the radiological (A) ICs or Fission Product Barrier (F) ICs.

EAL#5 This EAL is based on the assumption that high winds within the PROTECTED AREA may have potentially damaged plant structures, listed in Tab1e*H2, containing functions or systems required for safe shutdown of the plant. The high wind site specific value is based on the wind speed (74 mph) to classify severe weather conditions as a hurricane. FSAR design basis is that all Seismic Category I structures at RBS are designed to withstand 100 mph fastest mile of sustained wind 30 ft-above ground, based upon a 100-yr period of recurrence. Methods to measure wind speed in the PROTECTED AREA are not available; therefore, a sustained indication of 74 mph on the Meteorological Tower lower elevation average wind speed indication will be used to determine that this EAL is met. The upper scale for the lower elevation average meter wind speed on the MET Tower is 100 mph If the MET Tower lower average wind speed sensors are not operable, other tower sensors or sources may be considered for estimating-wind speed at RBS such as NOAA or Baton Rouge regional Airport.

If damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.

References:

EIP-2-001 REV -028 PAGE 88 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 54 OF 107 EAL BASES HAI Initiating Condition - ALERT HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat Operating Mode Applicability: All Emergency Action Level(s): (1 or 2)

1. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the RBS security shift supervision

2. A validated notification from NRC of an airliner attack threat within 30 minutes of the site Basis:

NOTE: Timely and accurate communication between Security Shiff Supervision and the Control Room is crucial for the implementation of effective Securit"t _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).

EAL#1 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 accideQ!:al events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the OWNER CONTROLLED AREA. 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 Independent Spent Fuel Storage Installations that may be outside the PROTECTED AREA but still in the OWNER CONTROLLED AREA.

EIP-2-001 REV -028 PAGE 89 OF 144

REFERENCE USE ATTACHMENTS PAGE 55 OF 107 EAL BASES HAI EAL#2 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.

The NRC 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 NRC.

References:

NEI 03-12 EIP-2-001 REV- 028 PAGE 90 OF 144

REFEREN CE USE ATTACHM ENT 8 PAGE 56 OF 107 EAL BASES HA2 Initiating Condition - ALERT Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert Operating Mode Applicability: All Emergency Action Level(s):

1. Other conditions exist which 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 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.

References:

EIP-2-001 REV -028 PAGE 91 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 57 OF 107 EAL BASES

)

HA3 Initiating Condition - ALERT Control

,- room evacuation has been initiated Operating Mode Applicability: All

-Emergency Action Level(s): *

1. AOP-0031, Shutdown from Outside the Main Control Room requires Control Room evacuation Basis:

With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facilities inay be necessary.

I '

Inability to establish plant control from outside the Control Room will escalate this event to c;1 Site Area Emergency.

References:

EIP-2-001 REV -028 PAGE 92 OF 144

REFEREN CE USE ATTACHM ENT 8 PAGE 58 OF 107 EAL BASES HA4 Initiating Condition - ALERT FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown Operating Mode Applicability: All Emergency Action Level(s):

1. FIRE or EXPLOSION resulting in VISIBLE DAMAGE to any of the structures or areas in Table H2 containing safety systems or component s or Control Room indication of degraded performance of those safety systems, -

Reactor Building Standby Cooling Tower Auxiliary Building Diesel Generator Building Control Building Tunnels (B, D, E, F, G)

Fuel Building Basis:

VISIBLE DAMAGE is used to identify the magnitude of the FIRE or EXPLOSIO N and to discriminate against minor FIRES and EXPLOSIO NS.

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 EXPLOSIO N was large enough to cause damage to these systems.

The use of VISIBLE DAMAGE should not be interpreted as mandating a lengthy damage assessmen t 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 Malfunction (S), Fission Product Barrier Degradation (F) or Abnormal Radiation Levels/ Radiological Effluent (A)

ICs.

References:

EIP-2-001 REV-028 PAGE 93 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 59 OF 107 EAL BASES HA5 Initiating Condition - ALERT 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

• Operating Mode Applicability: All Emergency Action Level(s):

Note: 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.

1. Access to Main Control Room, Auxiliary Building, or 95' Control Building 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 Basis:

Gases in a VITAL AREA can affect the ability to safely operate or safely shutdown the reactor. The Auxiliary Building and the 95'. Control Building are included with the Main Control Room due to required operator actions per the system operating procedure to place shutdown cooling in service.

The fact that SCBAs 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.

EIP-2-001 REV- 028 PAGE 94 OF 144

_J

REFERENCE USE ATTACHMENT 8 PAGE 60 OF 107 EAL BASES HAS 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 Malfunction (S), Fission Product Barrier Degradation (F) or Abnormal Radiation Levels/ Radioactive Effluent (A)

ICs.

References:

/

EIP-2-001 REV - 028 PAGE 95 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 61 OF 107 EAL BASES HA6 Initiating Condition -ALERT Natural or destructive phenomena affecting VITAL AREAS Operating Mode Applicability: All Emergency Action Level(s): (1 or 2 or 3 or 4 or 5 or 6)

1. a. Seismic event > Operating Basis Earthquake (OBE) as indicated by: .

Annunciator ,iSEISMIC SYS RECORDING/ TROUBLE" (P680-02A-D06)

AND ERS-NBR3D TRIGGER (RECORD START) is yellow AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A-B06) AND ERS-NBR3A OBE (HI) yellow light

b. Earthquake confirmed by any of the following:

• Earthquake felt in plant

• National Earthquake Center

• Control Room indication of degraded performance of systems 'required for the safe shutdown of the plant

2. Tornado striking resulting in VISIBLE DAMAGE to any of the Table H2 structures or areas containing safety systems or components or Control Room indication of degraded performance of those safety systems OR

• 3. Internal flooding in Auxiliary Building 70 ft elevation resulting in an electrical shock hazard that precludes access to operate or monitor safety equipment or Control Room indication of qegraded performance of those safety systems OR

4. Turbine failure-generated PROJECTILES resulting in VISIBLE DAMAGE to or penetration of any of the Table H2 structures or areas containing safety systems or components or Control Room indication of degraded performance of those safety systems.

OR

5. Vehicle crash resulting in VISIBLE DAMAGE to any of the Table H2 structures or areas containing safety systems or components or Control Room indication of degraded performance of those safety systems.

EIP-2-001 REV -028 PAGE 96 OF 144

REFEREN CE USE ATTACHM ENT 8 PAGE 62 OF 107 EAL BASES HA6 OR

6. Hurricane or high SUSTAINED wind conditions;;;: 74 mph within the PROTECTE D AREA boundary and resulting in VISIBLE DAMAGE to any of the Table H2 structures or areas containing safety systems or components or Control Room indication of degraded performance of those safety systems Reactor Building Standby Cooling Tower Auxiliary Building Diesel Generator Building Control Building Tunnels (8, D, E, F,G)

Fuel Building Basis:

These EALs escalate from HU6 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 structufes 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 these EALs 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 (S) ICs.

The Emergency Director may consider the Fuel Building as necessary to address the impact of the event on the loss of spent fuel cooling or spent fuel (e.g., freshly off-loaded reactor core in pool). At RBS, the term "freshly off-loaded reactor core" refers to fuel that has been discharged from the core and stored in the spent fuel pool for a period of LESS THAN one year.

EAL#1 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.

EIP-2-001 REV-028 PAGE 97 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 63 OF 107 EAL BASES HA6 EAL#2 This EAL is based on a tornado striking (touching down) that has caused VISIBLE DAMAGE to structures or areas containing functions or systems required for safe shutdown of the plant.

EAL#3 This EAL addresses the effect of internal flooding caused by events such 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.

The areas of concern are the Auxiliary Building 70 foot elevation cubicles and crescent area that contain systems required for safe shutdown of the plant that are not designed to be partially or fully submerged. Indication may be by local verification, control room indication, or in degraded performance of systems affected by the flooding.

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.

EAL#4 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. Some structures on the list may not be at risk for the turbine generated missile but are included for consistency in identifying structures or areas containing systems and functions required for safe shutdown of the plant.

EAL#5 This EAL addresses vehicle crashes within the PROTECTED AREA that result in VISIBLE DAMAGE to VITAL AREAS (as shown in Table H2) or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant.

EIP-2-001 REV - 028 PAGE 98 OF 144

REFERENCE USE ATTACHME NT 8 PAGE 64 OF 107 EAL BASES HA6 EAL#6 This EAL is based on high winds within the PROTECTED AREA that have caused VISIBLE DAMAGE to structures or areas containing functions or systems required for safe shutdown of the plant. The high wind site specific value is based on the wind speed (74 mph) to classify severe weather conditions as a hurricane. FSAR design basis is that all Seismic Category I structures at RBS are designed to withstand 100 mph fastest mile of sustained wind 30 ft above ground, based upon a 100-yr period of recurrence. Methods to measure wind speed in the PROTECTED AREA are not available; therefore, a sustained indication of 74 mph on the Meteorological Tower lower elevation average wind speed indication will be used to determine that this EAL is met. The upper scale for the lower elevation average wind speed on the MET Tower is 100 mph. If the MET Tower lower average wind speed sensors are not operable, other tower sensors or sources may be considered for estimating wind speed at RBS such as NOAA or Baton Rouge regional Airport.

References:

EIP-2-001 REV-028 PAGE 99 OF 144

i I

I REFERENCE USE ATTACHMENTS PAGE 65 OF 107 EAL BASES HS1 Initiating Condition - SITE AREA EMERGENCY HOSTILE ACTION within the PROTECTED AREA-Operating Mode Applicability: All Emergency Action Level(s):

1. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the RBS security shift supervision Basis:

This condition represents an escalated threat to plant safety above that contained iii 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 Organization readiness and preparation for the implementation of protective measures.

This EAL addresses the potential for a very rapid progression of events due to a HOSTILE ACTION. It i~ not intended to address incidents that are accidental events or acts of civil disobedience, such as

• small aircraft impact, hunters, I

or physical disputes between employees within the PROTECTED AREA.

Those events are adequately addressed by other EALs .

.Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack.

References:

NEI 03-12

/

EIP-2-001 REV- 028 PAGE 100 OF 144

REFERENCE USE ATTACHME NT 8 PAGE 66 OF 107 EAL BASES HS2 Initiating Condition - SITE AREA EMERGENCY Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency Operating Mode Applicability: All Emergency Action Level(s):

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve 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 beyond the SITE BOUNDARY 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.

References:

EIP-2-001 REV- 028 PAGE 101 OF 144

REFERENCE USE ATTACHMENTS PAGE 67 OF 107 EAL BASES HS3 Initiating Condition - SITE AREA EMERGENCY Control Room evacuation has been initiated and plant control cannot be established Operating Mode Applicability: All Emergency Action Level(s):

1. a. Control room evacuation has been initiated

b. Control of the plant cannot be established in accordance with AOP-0031, Shutdown from Outside the Main Control Room, within 15 minutes Basis:

The intent of this IC 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 such as 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) ..

The determination of whether or not control is established -at the remote shutdown panel is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes that the plant staff 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 (F) or Abnormal Radiation Levels/Radiological Effluent (A) EALs.

References:

EIP-2-001 REV - 028 PAGE 102 OF 144

REFERENCE USE ATTACHM ENT 8 PAGE 68 OF 107 EAL BASES HGl Initiating Condition - GENERAL EMERGEN CY HOSTILE ACTION resulting in loss of physical control of the facility Operating Mode Applicability: All Emergency Action Level(s): (1 or 2)

1. A HOSTILE ACTION has occurred such that plant personnel are unable to operate equipment required to maintain safety functions

2. 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 Basis:

EAL#1 This EAL encompasses conditions under which a HOSTILE ACTION has resulted in a loss of physical control of VITAL AREAS (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. These safety functions are reactivity control (ability to shut down the reactor and keep it shutdown), reactor water level (ability to cool the core), and decay* heat removal (ability to maintain a heat sink).

Loss of physical control of the Control Room or remote shutdown panel 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.

EAL#2 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. At RBS, the term "freshly off-loaded reactor core" refers to fuel that has been discharged from the core and stored in the spent fuel pool for a period of LESS THAN one year.

References:

NEI 03-12 EIP-2-001 REV-028 PAGE 103 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 69 OF 107 EAL BASES J

HG2 Initiating Condition - GENERAL EMERGENCY Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency Operating Mode Applicability: All r

Emergency Action Level(s):

1. Other conditions exist which 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 offsite for more than the immediate site area 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.

References:

EIP-2-001 REV -028 PAGE 104 OF 144

REFEREN CE USE ATTACHM ENT 8 PAGE 70 OF 107 EAL BASES SUI Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all offsite AC power to emergency busses for~ 15 minutes Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. Loss of all offsite AC power to Div I & 11 ENS busses for~ 15 minutes Basis:

Preferred station transformers are: 1RTX-XSR1C, 1RTX-XSR1D, 1RTX-XSR 1E and 1RTX-XSR1F.

Prolonged loss of offsite 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.

References:

(

EIP-2-001 REV -028 PAGE 105 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 71 OF 107 EAL BASES SU6 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of safety system annunciation or indication in the Control Room for::. 15 minutes Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. UNPLANNED Loss of> approximately 75% of the following for::. 15 minutes:

a. Control room safety system annunciation OR

b. Control Room safety system indication Basis:

This IC and its associated EAL are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Recognition of the availability of computer based indication equipment is considered e.g., SPDS, plant computer, etc ..

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

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 NOUE is based on SU11 "Inability to reach

• required operating mode within Technical Specification limits."

EIP-2-001 REV- 028 PAGE 106 OF 144

REFERENCE USE ATTACHME NT 8 PAGE 72 OF 107 EAL BASES

-SU6-Annunciators or indicators for this EAL include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures (EOPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.). Indicators associated with safety systems are those indicators for reactivity control, core cooling, RCS status and containment status. The panels to consider include:

H13-P601, H13-P680, H13-P808 (CMS and ORMS), H13-P863 (ORMS), P870 and P877 safety related annunciators and indicators.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

This NOUE will be escalated to-an Alert based on a concurrent loss of compensatory indications or if a SIGNIFICANT TRANSIENT is in progress during the loss of annunciation or indication.

References:

EIP-2-001 REV-028 PAGE 107 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 73 OF 107 EAL BASES SU7 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT RCS leakage Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s): (1 or 2)

Note: A relief valve that operates and fails to close per design should be considered applicable if the relief valve cannot be isolated.

1. Unidentified or pressure boundary leakage > 1O gpm OR

2. Identified leakage > 35 gpm Basis:

This IC is included as a NOUE 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 IC. However, a relief valve that operates and fails to close per design should be considered applicable to this IC if the relief valve cannot be isolated. The 15 minute EAL assessment period begins when the relief valve should have closed. An attempt for isolation from the Control Room should be made prior to classification. If operator actions from the Control Room are successful within the 15 minute EAL assessment period, this threshold is not applicable. Credit is not given for operator actions taken outside the Control Room.

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 IC to the Alert level is via Fission Product Barrier Degradation (F) ICs.

References:

RBS Technical Specification 3.4.5 EIP-2-001 REV - 028 PAGE 108 OF 144

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"'---------------------"-~""'"'---"------.......,__ l

REFERENCE USE ATTACHMENTS PAGE 74 OF 107 EAL BASES SU8 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s): (1 or 2)

1. Loss of all of the following onsite communications methods affecting the ability to perform routine operations:

Plant radio system Plant paging system Sound powered phones In-plant telephones

2. Loss of all of the following offsite communications methods affecting the ability to perform offsite notifications:

All telephones NRC phones State of Louisiana Radio Offsite notification system and hotline Basis:

The purpose of this IC and its associated EALs 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 offsite authorities.

The availability of one method of ordinary offsite 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.

References:

EIP-2-001 REV-028 PAGE 109 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 75 OF 107 EAL BASES SU9 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Fuel clad degradation Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s): (1 or 2)

3. Offgas pre-treatment radiation monitor reading > the Table S1 Dose Rate Limit for the actual indicated offgas flow indicating fuel clad degradation > T.S. allowable limits

~15 9000

>15-17 8000

>17-20 7000~

>20-25 5000

>25-30 4000

>30-60 2000

>60-140 1000

>140-200 700 OR

2. Reactor coolant sample activity value indicating fuel clad degradation > T.S. allowable limits

• >4.0 µCi/gm dose equivalent 1-131

• >0.2 µCi/gm dose equivalent 1-131 for> 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EIP-2-001 REV-028 PAGE 110 OF 144

REFERENCE USE ATTACHMENT S PAGE 76 OF 107 EAL BASES SU9 Basis:

This IC 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.

EAL#1 This EAL addresses site-specific radiation monitor readings that provide indication of a degradation of fuel clad integrity.

The Technical Specification limit of 290 mCi/sec Offgas pre-treatment release is equivalent to 11,21 O mR/hr (assumes flow of 17.875 cfm without adjustment for instrument accuracy). The Table S1 values account for instrument inaccuracy and changing offgas flow rate. The dose rate in the table corresponds to the adjusted TS limit for that associated indicated flow. The dose rates are rounded down conservatively to more accurately read the values on the available scale. The table dose rate values may not reflect the H13-P601/22NF03 alarm setpoint. To determine if EAL conditions are met when the pre-treatment high radiation alarm (H13-P601/22A/F03) is lit, the operator must read the actual indicated offgas flow rate on N64-R620 (Panel H13-P845) and indicated pre-treatment mR/h-r value on D17-R604 (Panel H13-P600). Compare the indicated mR/hr value with the Table S1 dose rate mR/hr for the indicated flow value. If the indicated mR/hr is greater than the Table S1 value, the EAL condition is met.

EAL#2 This EAL addresses coolant samples exceeding coolant technical specifications for transient iodine spiking limits and coolant samples exceeding coolant Technical Specifications for nominal operating iodine limits for the time period specified in the Technical Specifications.

Escalation of this IC to the Alert level is via the Fission Product Barriers (F).

References:

TS 3.4.8/B 3.4.8 TS 3.7.4/ B 3.7.4 G13.18.9.6.*012 Rev O

~~} ~ ~:~:;:~~

USAR 15.7.1

~:;~

EC-500004 7036 EIP-2-001 REV -028 PAGE 111 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 77 OF 107 EAL BASES SUlO Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality Operating Mode Applicability: Mode 3 ...... Hot Shutdown Emergency Action Level(s):

1. UNPLANNED sustained positive period observed on nuclear instrumentation Basis:

This IC addresses inadvertent criticality events. This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated).

This condition can be identified using period monitors. The term "sustained" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration. These short term positive periods are the result of the rise in neutron population due to subcritical multiplication.

Escalation would be by the Fission Product Barrier.Table (F), as appropriate to the operating mode at the time of the event.

References:

EIP-2-001 REV-028 PAGE 112 OF 144

REFERENCE USE ATTACHM ENT 8 PAGE 78 OF 107 EAL BASES SUll Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Inability to reach required operating mode within Technical Specification limits Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

1. Plant is not brought to required operating mode within Technical Specifications LCO Action Statement time 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

,i'mmediate NOUE 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 a NOUE is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed.

References:

EIP-2-001 REV -028 PAGE 113 OF 144

REFERENCE USE ATTACHMENTS PAGE 79 OF 107 EAL BASES SAl Initiating Condition - ALERT AC power capability to emergency busses-reduced to a single power source for

~15 minutes such that any additional single failure would result in station blackout Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. a. AC power capability to Div I and 11 ENS busses reduced to a single power source for :::. 15 minutes

b. Any additional single power source failure will result in station blackout Basis:

Preferred station transformers are: 1RTX-XSR1C, 1RTX-XSR1D, 1RTX-XSR1E and 1RTX-XSR1F.

The condition indicated by this IC is the degradation of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of all but one emergency diesel generator to supply power to its emergency busses. Another related condition could be the loss of all offsite power and loss of onsite emergency diesels generators with only one train of emergency busses being backfed from the unit main generator, or the loss of onsite emergency diesel generators with only one train of emergency busses being fed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Div Ill D/G and bus E22-S004 are not discussed explicitly in this IC. The loss of Div I and Div II are considered a station blackout. If Div Ill D/G or E22-S004 is available, entry into this IC is applicable.

References:

EIP-2-001 REV-028 PAGE 114 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 80 OF 107 EAL BASES SA3 Initiating Condition - ALERT Automatic scram fails to shutdown the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Emergency Action Level(s):

1. a. An automatic scram failed to shutdown the reactor

b. Manual actions taken at the reactor control console successfully shutdown the reactor as indicated by reactor power < 5%

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.

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.

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.

Reactor shutdown is considered to be when power is below 5%. The Emergency Operating Proceedure (EOP) definition of shutdown is not used.

If manual actions taken at the reactor control console fail to shutdown the reactor, the event would escalate to a Site Area Emergency.

References:

EIP-2-001 REV-028 PAGE 115 OF 144

./

REFERENCE USE ATTACHMENT 8 PAGE 81 OF 107 EAL BASES SA6 Initiating Condition - ALERT 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 Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. a. UNPLANNED loss of> approximately 75% of the following for~ 15 minutes:

• Control room safety system annunciation

• Control Room safety system indication

b. Either of the following:

• A SIGNIFICANT TRANSIENT is in progress OR.

• Compensatory indications are unavailable Basis:

This IC 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.

Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

EIP-2-001 REV-028 PAGE 116 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 82 OF 107 EAL BASES SA6 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 Manager 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 mqst 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 NOUE is based on SU11 "Inability to reach required operating mode within Technical Specification limits."

Annunciators or indicators for this EAL include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures (EOPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.). Indicators associated with safety systems are those indicators for reactivity control, core cooling, RCS status and containment status. The panels to consider include:

H13-P601, H13-P680, H13-P808 (CMS and ORMS), H13-P863 (ORMS), P870 and P877 safety related annunciators and indicators.

"Compensatory indications" in this context includes computer based information such as SPOS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required. Fifteen 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.

References:

EIP-2-001 REV - 028 PAGE 117 OF 144

REFERENCE USE ATTACHMENTS PAGE 83 OF 107 EAL BASES SS1 Initiating Condition - SITE AREA EMERGENCY Loss of all offsite and all onsite AC power to emergency busses for

~ 15 minutes Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. Loss of all offsite and all onsite AC power to Div I, II and Ill ENS busses for

~ 15 minutes Basis:

Preferred station transformers are: 1RTX-XSR1 C, 1RTX-XSR1 D, 1RTX-XSR1 E and 1RTX-XSR1 F.

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 this event can escalate to a General Emergency.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to emergency busses. Even though an emergency bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel 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 SAE per SS1 should be declared.

Escalation to General Emergency is via Fission Product Barrier Degradation (F) or IC SG1, "Prolonged loss of all offsite and all onsite AC power to emergency busses."

References:

EIP-2-001 REV - 028 PAGE 118 OF 144

REFERENCE USE ATTA CHME NTS

.PAGE 84 OF 107 EAL BASE S Initiating Condition - SITE AREA EMER GENC Y SS3 Automatic scram fails to shutdown the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor Operating Mode Applicability: Mode 1 ...... Power Operation Mode 2 ...... Startup Emergency Action Level(s):

1. a. An automatic scram failed to shutdown the reactor

b. Manual actions taken at the reactor control console do not shutdo wn the reactor as indicated by reactor power ~ 5%

Basis:

Automatic and manual scrams are not considered successful if action away from the reactor control console was required to scram the reactor.

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 the reactor. This EAL is still applicable even if actions taken away from the reactor control console are successful in shutting the reactor down becaus e the design limits of the fuel may have been exceeded or because of the gross failure of the Reacto r Protection System to shutdown the plant.

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 setpoin t is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiate d.

Reactor shutdown is considered to be when power is below 5%.

The Emergency Operating Proceedure (EOP) definition of shutdown is not used.

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.

References:

EIP-2-001 REV- 028 PAGE 119 OF 144

REFERENCE USE ATTACHMENTS PAGE 85 OF 107 EAL BASES SS4 Initiating Condition - SITE AREA EMERGENCY Loss of all vital DC power for ~ 15 minutes Operating Mode Applicability:* Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. < 105 VDC on all vital DC busses for .:::. 15 minutes Basis:

Loss of all DC power compromises abjlity 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.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation to a General Emergency would occur by Abnormal Radiation Levels/Radiological Effluent (A), Fission Product Barrier Degradation (F).

I

References:

EIP-2-001 REV-028 PAGE 120 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 86 OF 107 EAL BASES SS6 Initiating Condition - SITE AREA EMERGENCY Inability to monitor a SIGNIFICANT TRANSIENT in progress Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

Note: 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 has exceeded, or will likely exceed, the applicable time.

1. a. Loss of> approximately 75% of the following for:::. 15 minutes :

• Control Room safety system annunciation

• Control Room safety system indication

b. A SIGNIFICANT TRANSIENT is in progress

c. Compensatory indications are unavailable Basis:

This IC 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 Manager be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

EIP-2-001 REV-028 PAGE 121 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 87 OF 107 EAL BASES SS6 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 NOUE is based on SU11 "Inability to reach required operating mode within Technical Specification limits."

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.

Annunciators or indicators for this EAL include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures (EOPs and SAPs), and in other EALs (e.g., area process, and/or effluent rad monitors, etc.). Indicators associated with safety systems are those indicators for reactivity control, core cooling, RCS status and containment status. The panels to consider include:

H13-P601, H13-P680, H13-P808 (CMS and ORMS), H13-P863 (ORMS), P870 and P877 safety related annunciators and indicators.

"Compensatory indications" in this context includes computer based information such as SPOS. This should include all computer systems available for this use depending on specific plant design and subsequent retrofits.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

References:

EIP-2-001 REV-028 PAGE 122 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 88 OF 107 EAL BASES SGl Initiating Condition - GENERAL EMERGENCY Prolonged loss of all offsite and all onsite AC power to emergency busses Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Mode 3 ...... Hot Shutdown Emergency Action Level(s):

1. a. Loss of all offsite and all onsite AC power to Div I, II and Ill ENS busses.

b. Either of the following:

• Restoration of at least one emergency bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely

• RPV level can not be maintained> -162 inches Basis:

Preferred station transformers are: 1RTX-XSR 1C, 1RTX-XSR 1D, 1RTX-XSR 1E and 1RTX-XSR 1F.

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 IC 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.

EIP-2-001 REV-028 PAGE 123 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 89 OF 107 EAL BASES SGl Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of Fission Product Barriers is IMMINENT?

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

Thus, 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.

References:

EIP-2-001 REV-028 PAGE 124 OF 144

REFERENCE USE ATTACHMENTS PAGE 90 OF 107 EAL BASES SG3 Initiating Condition - GENERAL EMERGENCY 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 Operating Mode Applicability: Mode 1...... Power Operation Mode 2 ...... Startup Emergency Action Level(s):

1. a. An automatic scram failed to shutdown the reactor

b. All manual actions do not shutdown the reactor as indicated by reactor power~ 5%

c. Either of the following exist or have occurred due to continued power generation:

• Core cooling is extremely challenged as indicated by RPV level cannot be restored and maintained > -187 inches OR

• Heat removal is extremely challenged as indicated by RPV pressure and Suppression Pool temperature cannot be maintained in the EOP Heat Capacity Temperature Limit (HCTL)

Safe Zone 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.

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 offsite intervention time.

References:

EIP-2-001 REV -028 PAGE 125 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 91 OF 107 EAL BASES CUI Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT RCS leakage Operating Mode Applicability: Mode 4 ... .... Cold Shutdown Emergency Action Level(s):

Note: 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. RCS leakage results in the inability to maintain or restore RPV level > +9. 7 inches (Level 3) for :::. 15 minutes

• Basis:

This IC 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 IC. However, a relief valve that operates and fails to close per design- should be consid.ered applicable to this IC if the relief valve cannot be isolated.

Prolonged loss of RCS Inventory may result in escalation to the Alert emergency classification level via either CA 1 or CA3. -

References:

EIP-2-001 REV-028 PAGE 126 OF 144

REFERENCE USE ATTACHME NTS PAGE 92 OF 107 EAL BASES CU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of RCS/RPV inventory Operating Mode Applicability: Mode 5 ...... Refueling Emergency Action Level(s): (1 or 2)

Note: 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. UNPLANNED RCS level drop as.indicated by either of the following:

a. RCS water level drop below the RPV flange for?'._ 15 minutes when the RCS level band is established above the RPV flange

b. RCS water level drop below the RPV level band for?'._ 15 minutes when the RCS level band is established below the RPV flange OR

2. RCS level cannot be monitored with a loss of RCS inventory as indicated by an unexplained rise in floor or equipment sump level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss Basis:

This IC 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 lower RCS water level below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level lowering below the RPV flange, or below the planned RCS water level for the given evolution (if the planned RCS water level is already below the RPV flange), warrants declaration of a NOLIE 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 makeup 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 CA 1 or CA3. (

EIP-2-001 REV-028 PAGE 127 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 93 OF 107 EAL BASES CU2 EAL#1 This EAL involves a drop in RCS level below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to reductions in flooded reactor cavity level, which is addressed by AU2 EAL 1, until such time as the level drops to the level of the vessel flange.

If RPV level continues to drop and reaches the Low-Low ECCS Actuation Setpoint then escalation to CA 1 would be appropriate.

EAL#2 This EAL addresses conditions in the refueling mode when normal means of core temperature indication and RCS 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 rise 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 CA 1 or CA3.

References:

EIP-2-001 REV- 028 PAGE 128 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 94 OF 107 EAL BASES CU3 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the RPV Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Emergency Action Level(s): (1 or 2)

Note: 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. UNPLANNED event results in RCS temperature exceeding 200 °F.

2. Loss of all RCS temperature and RCS/RPV level indication for:::. 15 minutes.

Basis:

This IC 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 RCS inventory. Since the RCS usually remains intact in the cold shutdown mode a large inventory of water is available to keep the core covered.

During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions that lower 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 rises in RCS/RPV temperatures depending on the time since shutdown.

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, EAL 2 would result in declaration of a NOUE 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 CA 1 based on an inventory loss or CA3 based on exceeding its temperature criteria.

References:

EIP-2-001 REV -028 PAGE 129 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 95 OF 107 EAL BASES CU5 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT AC power capability to emergency busses reduced to a single power source for :::. 15 minutes such that any additional single failure would result in station blackout Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Emergency Action Level(s):

Note: 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. a. AC power capability to Div I and Div 11 ENS busses reduced to a single power source for :::. 15 minutes

b. Any additional single power source failure will result in station blackout Basis:

Preferred station transformers are: 1RTX-XSR1C, 1RTX-XSR1D, 1RTX-XSR1E and 1RTX-XSR1F.

The condition indicated by this IC is the degradation of the offsite and onsite AC power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of all but one emergency generator to supply power to its emergency busses. Another related condition could be the loss of onsite emergency diesels generators with only one train of emergency busses being fed from offsite power (or backfed from offsite power through the main transformer). The subsequent loss of this single power source would escalate the event to an Alert in accordance with CA5.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Div Ill D/G and bus E22-S004 are not discussed explicitly in this IC. The loss of Div I and Div II are considered a station blackout. If Div Ill D/G or E22-S004 is available, entry into this IC is applicable.

References:

EIP-2-001 REV-028 PAGE 130 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 96 OF 107 EAL BASES CU6 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of required DC power for~ 15 minutes Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Emergency Action Level(s):

Note: 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. < 105 VDC on required Vital DC busses for~ 15 minutes Basis:

The purpose of this IC and its associated 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.

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.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

References:

EIP-2-001 REV-028 PAGE 131 OF 144

REFERENCE USE ATTACHMENTS PAGE 97 OF 107 EAL BASES CU7 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent criticality Operating Mode Applicability: Mode 4 ...... Cold Shutdown.

Mode 5 ...... Refueling Emergency Action Level(s):

1. UNPLANNED sustained positive period observed on nuclear instrumentation Basis:

This IC addresses criticality events that occur in Cold Shutdown or Refueling modes such as fuel mis-loading events . This IC indicates a potential degradation of the level of safety of the plant, warranting a NOUE classification. '

This condition can be identified using period monitors. The term "sustained" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alteration. These short term positive periods are the result of the rise in neutron population due to subcritical multiplication.

Escalation would be by Emergency Director Judgment.

References:

EIP-2-001 REV- 028 PAGE 132 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 98 OF 107 EAL BASES CU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of all onsite or offsite communications capabilities Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Mode D ...... Defueled Emergency Action Level(s): (1 or 2)

1. Loss of all of the following onsite communication methods affecting the ability to perform routine operations:

Plant radio system Plant paging system Sound powered phones In-plant telephones

2. Loss of all of the following offsite communication methods affecting the ability to perform offsite notifications:

All telephones NRC phones State of Louisiana Radio Offsite notification system and hotline Basis:

The purpose of this IC and its associated EALs 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 offsite authorities. The loss of offsite 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 offsite 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 offsite locations, etc.)

are being utilized to make communications possible.

References:

EIP-2-001 REV -028 PAGE 133 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 99 OF 107 EAL BASES CAI Initiating Condition - ALERT Loss of RCS/RPV inventory Operating Mode Applicability: Mode 4 ... .... Cold Shutdown Mode 5 ....... Refueling Emergency Action Level(s): (1 or 2)

Note: 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. UNPLANNED loss of RCS inventory as indicated by RPV level < -43 inches (Level 2)

(

2. RCS level cannot be monitored for~ 15 minutes with a loss of RCS inventory as indicated by an unexplained rise in floor or equipment sump level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss Basis:

These EALs are not applicable when the RPV is defueled and serve as precursors 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 lowering and potential core uncovery. This condition will result in a minimum emergency classification level of an Alert.

EAL#1 The inability to restore and maintain level after reaching this setpoint would be indicative of a failure of the RCS barrier.

EAL#2 In the cold shutdown mode, normal RCS level and 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 reactor vessel 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 rise 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.

If RPV level continues to lower then escalation to Site Area Emergency will be via CS1.

References:

.EIP-2-001 REV-028 PAGE 134 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 100 OF 107 EAL BASES CA3 Initiating Condition - ALERT Inability to maintain plant in cold shutdown Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Emergency Action Level(s): (1 or 2)

1. An UNPLANNED event results in RCS temperature> 200 °F for> the specified duration in Table C2 OR

2. An UNPLANNED event results in RCS pressure rise > 1O psig due to a loss of RCS cooling Table C2: RCS Reheat Duration Thresholds RCS Containment Closure Duration Intact N/A 60 minutes*

Not intact Established 20 minutes*

Not Established O minutes

• If an RCS heat removal system is in operation within this time frame and RCS temperature is beinq reduced, then the EAL is not applicable.

Basis:

EAL#1 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 RCS integrity is established. RCS integrity should be considered to be in place 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 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 RCS integrity is not established.) As discussed above, RCS integrity should be assumed. to be in place 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, the EAL addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are established.

The (*) 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 specified time frame.

EIP-2-001 REV-028 PAGE 135 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 101 OF 107 E_ALBASES CA3 EAL#2 The 10 psig pressure rise 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 psig or the lowest pressure that the site can read on installed Control Board instrumentation that is equal to or greater than 10 psig.

Escalation to Site Area Emergency would be via CS1 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 .

References:

I I

EIP-2-001 REV-028 PAGE 136 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 102 OF 107 EAL BASES CA5 Initiating Condition - ALERT Loss of all offsite and all onsite AC power to emergency busses for~ 15 minutes Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5 ...... Refueling Mode D ...... Defueled Emergency Action Level(s):

Note: 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. Loss of all offsite and all on-site AC power to Div I & Div 11 ENS busses for~ 15 minutes Basis:

Preferred station transformers are: 1RTX-XSR1C, 1RTX-XSR1D, 1RTX-XSR1E and 1RTX-XSR1F.

Loss of all AC power to Div I & Div II 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, raising the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to emergency busses. Even though an emergency bus may be re-energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not functional on the energized bus, then the bus should not be considered restored for this EAL.

Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels / Radiological Effluent ICs.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

References:

EIP-2-001 REV-028 PAGE 137 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 103 OF 107 EAL BASES CS1 Initiating Condition - SITE AREA EMERGENCY Loss of RCS/RPV inventory affecting core decay heat removal capability Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5... . . . Refueling Emergency Action Level(s): (1 or 2 or 3)

Note: 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. With CONTAINMENT CLOSURE not established, UNPLANNED RPV level< -49 inches

2. With CONTAINMENT CLOSURE established, RPV level< -162 inches (TAF)

OR

3. RCS level cannot be monitored for~ 30 minutes with a loss of RCS inventory as indicated by any of the following:

• RMS-RE16 reading> 100 R/hr

• Erratic Source Range Monitor indication

• Unexplained rise in floor or equipment sump level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss Basis:

These EALs are not applicable when the RPV is defueled.

Under the conditions specified by this IC, continued reduction in RCS 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 or AG1.

EAL#3 In the cold shutdown mode, normal RCS level and 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 rise 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.

EIP-2-001 REV-028 PAGE 138 OF 144

REFERENCE USE ATTACHMENTS

• PAGE 104 OF 107 EAL BASES CSI The 30-minute duration allows sufficient time for actions to be performed to recover inventory control equipment.

As water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.

References:

COP-1050 NEDC-33045P Calculation 813.18.9.4-047 Rev. O EIP-2-001 REV-028 PAGE 139 OF 144

REFERENCE USE ATTACHMENTS PAGE 105 OF 107 EAL BASES CGl Initiating Condition 7 GENERAL EMERGENCY Loss of RCS/RPV inventory affecting fuel clad integrity with containment challenged Operating Mode Applicability: Mode 4 ...... Cold Shutdown Mode 5. . . . . . Refueling Emergency Action Level(s): (1 or 2)

Note: 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 ex.ceed the applicable time.

1. a. RPV level < - 162 inches (TAF) for:::. 30 minutes

b. Any containment challenge indication in. Table C1

2. a. RCS level cannot be monitored with core uncovery indicated by any of the following for:::. 30 minutes:

• RMS-RE16 reading > 100 R/hr

• Erratic Source Range Monitor indication

• Unexplained rise in floor or equipment sump level, Suppression Pool level, vessel make-up rate or observation of leakage or inventory loss AND

b. Any containment challenge indication in Table C1

• .. -:* >:. tab1~;G\. . *

. c6iitainm~nl.~h~ll~nge'. :

. . :indidad},n,s:, . .. : :

• CONTAINMENT CLOSURE not established

• Explosive mixture inside containment

• UNPLANNE_D rise in containment pressure

• Secondary containment area radiation monitor above EOP Max Safe 0 eratin Value below:

Area ORMS Max Safe Operating Value Grid 2 RHR Equip 1213 9.5E+03 mR/hr Rm A RHR Equip 1214 9.5E+03 mR/hr RmB RHR Equip 1215 9.5E+03 mR/hr RmC EIP-2-001 REV-028 PAGE 140 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 106 OF 107 EAL BASES CGI Basis:

These EALs are not applicable when the RPV is defueled.

This IC 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 and .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.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gasses in Containment. However, Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists.

• EAL#2 Sump and tank level rise 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.

In the cold shutdown mode, normal RCS level and 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 instaHed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupte.d:-However, if all level indication were to be lost dwing 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 rise must be evaluated against other potential sources of lec;1kage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

As water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in site specific monitor indication and possible alarm.

References:

COP-1050 NEDC-33045P EIP-2:.001 REV - 028 PAGE 141 OF 144

REFERENCE USE ATTACHMENT 8 PAGE 107 OF 107 EAL BASES E-HUI Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINE MENT BOUNDARY Operating Mode Applicability: All Emergency Action Level(s):

1. Damage to a loaded cask CONFINEMENT BO UNDARY.

Basis:

A NOUE in this IC 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.

This EAL addresses a dropped cask, a tipped over cask, EXPLOSION, PROJECTILE damage, FIRE damage or natural phenomena affecting a cask (e.g. , seismic event, tornado, etc.) .

References:

EIP-2-001 REV-028 PAGE 142 OF 144

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ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 1 of 8 I. OVERVIEW PAD Rev.#: _o_

Facility: River Bend Station Proposed Activity / Document: Procedure Revision/ EIP-2-001 Classification of Emergencies Change/Rev. #: 28 Description of Proposed Activity: Revise EIP-2-001 to update seismic Emergency Action Levels (EALs) changed due to Seismic Monitoring System Upgrade (that PAD is attached).

1) Attachment 4 Page 27, Attachment 9 User Aid 1 Page 143 and Attachment 9 User Aid 2 Page 144:

From: Seismic event identified by any 2 of the following:

Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Annuiciator "Seismic Tape Recording SYS Start" (P680-02A-D06)

b. Event Indicator on ERS-NBl-102 is white To: Seismic event identified by any 2 of the following:

Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Annuiciator "SEISMIC SYS RECORDING / TROUBLE" (P680-02A-D06)

b. Event Indicator on ERS-NBR3D TRIGGER (RECORD START) is yellow

2) Attachment 4 Page 27, Attachment 9 User Aid 1 Page 143 and Attachment 9 User Aid 2 Page 144:

From: a. Seismic event> Operating Basis Earthquake (OBE) as indicated by:

Annunciator "Seismic Tape Recording System Start" (P680-02A-D06)

AND Event Indicator on ERS-NBl-102 is white AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A-B06) AND amber light(s) on panel NBl-101 To: a. Seismic event> Operating Basis Earthquake (OBE) as indicated by:

Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A-D06)

AND ERS-NBR3D TRIGGER (RECORD START) is yellow AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A-B06) AND ERS-NBR3A QBE (HI) yellow light

3) Attachment 8 Page 86:

From: 1. Seismic event identified by any 2 of the following:

Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Annunciator "Seismic Tape Recording SYS Start" (P680-02A-D06)

b. Event Indicator on ERS-NBl-102 is white To: 1. Seismic event identified by any 2 of the following:

Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. AnmJ'lciator "SEISMIC SYS RECORDING / TROUBLE" (P680-02A-D06)

b. Event Indicator on ERS-NBR3D TRIGGER (RECORD START) is yellow

4) Attachment 8 Page 87:

From: The annunciators "Seismic Tape Recording SYS Starf' and the \Nhite" event indicator are listed in the Alarm Response Procedure as verification of an earthquake event To: The annunciators "SEISMIC SYS RECORDING/ TROUBLE" and the yellow" event indicator are listed in the Alarm Response Procedure as verification of an earthquake event.

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 2 of 8

5) Attachment 8 Page 96:

From: a. Seismic event > Operating Basis Earthquake (OBE) as indicated by: .

Annunciator "Seismic Tape Recording System Start" (P680-02A-D06)

AND Event Indicator on ERS-NBl-102 is white AND Receipt of EITHER 1 OR 2: ,

1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A-806) AND amber light(s) on panel NBl-101 To: a. Seismic event> Operating Basis Earthquake (OBE) as indicated by: .

Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A-D06)

AND ERS-NBR3D TRIGGER (RECORD START) is yellow AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A-806) AND ERS-NBR3A OBE (HI) yellow light II. DOCUMENT REVIEW METHOD Provide the requested information for each item below.

1. For documents available electronically:

a. List search engine or documents searched, and keywords used:

Manual electronic search of Technical Requirements Manual, Technical Specifications, Updated Safety Analysis Report done using keywords: 'Emergency Action Level'

'HU6' 'HA6'- no hits Search of the Emergency Plan using key words 'Emergency Action Level' 'HU6' 'HA6' had relevant hits on 'Emergency Action Level' in Sections 13.3.1.1 Definitions, 13.3.3.1 Classification System, HU6 and HA6 in Table 13.3-1

b. List relevant sections of controlled electronic documents reviewed:

Section 13.3.1.1 Definitions Section 13.3.3.1 Classification System Section 13.3.6.3.1 Onsite Assessment Facilities Table 13.3-1 Emergency Action Levels and Initiating Conditions

2. Documents reviewed manually (hardcopy):

The original Emergency Plan (1986) Sections 13.3.3.1, 13.3.6.3.1 and Table 13.3-1 were reviewed manually. No additional relevant or affected plan content was identified.

3. For those documents that are not reviewed either electronically or manually, use the specific questions provided in Sections Ill and IV of Attachment 9.2 of EN-Ll-100 as needed. Document, below, the extent to which the Attachment 9.2 questions were used.

The screening questions in Section Ill and IV of EN-Ll-100, Attachment 9.2 were ti reviewed during preparation of this Process Applicability Determination. No further documents were impacted beyond what is provided above.

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 3 of 8 Ill. PROCESS REVIEW Does the proposed activity affect, invalidate, or render incorrect, OR have the potential to affect, invalidate, or render incorrect, information contained in any of the following processes? Contact Program Owner if needed. Associated regulations and procedures are identified with each process below.

flROCE.SS (Regulations,/ Procedures)

• YES 'Nb REVIEW RES UL TS

. . ,' . * *. . . . . ' . ,t

• Chemistry/ Effluents D IB]

Radwaste / Process Control Program (PCP) D IB]

(EN-RW-105 or contact the Radiation Protection Dept.)

Radiation Protection / ALARA D IB]

(1 O CFR 20 / EN-RP-110 or contact the Radiation Protection Dept) lnservice Inspection Program (10 CFR 50.55a / EN-DC-333, -342, D IB]

-351, -352) lnservice Testing Program (10 CFR 50.55a / EN-DC-332) D IB]

Maintenance Rule Program (10 CFR 50.65 / EN-DC-203, -204, -205, -206, D IB]

-207)

Containment Leakage Rate Testing (Appendix J) Program (10 CFR 50 D IB]

Appendix J / EN-DC-334)

FLEX Program (NRC Order EA-12-49/NRC Order EA-12-051/FLEX D IB]

Program) (10 CFR 50.59 / EN-OP-201)

IF any box is checked "Yes," THEN contact the appropriate department to ensure that the proposed change is acceptable and document the results in the REVIEW RESULTS column.

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 4 of 8 IV. LICENSING BASIS DOCUMENT REVIEW Does the proposed activity affect, invalidate, or render incorrect, OR have the potential to affect, invalidate, or render incorrect, information contained in any of the following Licensing Basis Document(s)? Contact LBD Owner if needed. Associated regulations and procedures are identified with each Licensing Basis Document below.

LICENSING BASIS DOCUMENTS REVIEW RES ULTS OR SECTIONS YES NO (Regulations/ Procedures) AFFECTED OR LBDCR #

Quality Assurance Program Manual (QAPM) IB]

[10 CFR 50.54(a), 10 CFR 50 Appendix B / EN-QV-104] D Fire Protection Program (FPP) [includes the Fire Safety Analysis/Fire Hazards Analysis (FSA/FHA)] D IB]

OL Condition, 10 CFR 50.48 / EN-DC-128)

Emergency Plan (includes the On-Shift Staffing Analysis) Section 13.3.6.3.1 Onsite Assessment

[10 CFR 50.54(q) / 10 CFR 50.47 / EN-EP-305/EN-NS-220] Facilities 0 D Table 13.3-1 Emergency Action Level Initiating Conditions Environmental Protection Plan (Appendix B of the OL, Environmental Evaluation/ EN-EV-115, EN-EV-117, D IB]

EN-Ll-103)

Security Plan IB]

[10 CFR 50.54(p) / EN-NS-210 / EN-NS-220 or contact site Security Dept.] D Cyber Security Plan IB]

[10 CFR 50.54 (p) / EN-NS-210]

D Operating License (OL) / Technical Specifications (TS)

(10 CFR 50.90 I EN-Ll-103) o* IB]

TS Bases (10 CFR 50.59 / EN-Ll-100 / EN-Ll-101) D IB]

Technical Requirements Manual (TRM) (including TRM Bases) IB]

(10 CFR 50.59 / EN-Ll-100 / EN-Ll-101)

D Core Operating Limits Report (COLR), and Pressure and Temperature Limits Report (PTLR) (TS Administrative Controls, EN-Ll-113, EN-Ll-100, D IB]

EN-Ll-101)

Offsite Dose Calculation Manual (ODCM)

(TS Administrative Controls I EN-Ll-113, EN-Ll-100 )

D IB]

Updated Final Safety Analysis Report (UFSAR) IB]

(10 CFR 50.71 (e) / EN-Ll-113, EN-Ll-100, EN-Ll-101)

D Storage Cask Certificate of Compliance (10 CFR 72.244 / EN-Ll-113) D* IB]

Cask FSAR (CFSAR) (including the CTS Bases)

(10 CFR 72.70 or 72.248 / EN-Ll-113, EN-Ll-100,EN-Ll-112) D IB]

10 CFR 72.212 Evaluation Report (212 Report) IB]

(10 CFR 72.48 / EN-Ll-100, EN-Ll-112) D NRC Orders (10 CFR 50.90 I EN-Ll-103 or as directed by the Order) D* IB]

NRC Commitments and Obligations (EN-Ll-110) D* IB]

Site-Specific CFR Exemption (10 CFR 50.12, 10 CFR 55.11, 10 CFR 55.13, 10 CFR 72.7)

D* IB]

• contact the site Regulatory Assurance Department if needed.

IF any box is checked "Yes," THEN ensure that any required regulatory reviews are performed in accordance with the referenced procedures. Prepare an LBDCR per procedure EN-Ll-113, as required, if a LBD is to be changed, and document any affected sections or the LBDCR #. Briefly discuss how the LBD is affected in Section VII.A.

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 5 of 8 V. 10 CFR 50.59 / 10 CFR 72.48 APPLICABI LITY Can the proposed activity be dispositione d by one or more of the following criteria? Check the aooropriate box (if any).

D An approved, valid 50.59/72.48 Evaluation covering associated aspects of the proposed activity already exists. Reference 50.59/72.48 Evaluation# (if applicable) or attach documentat ion. Verify the previous 50.59/72.48 Evaluation remains valid.

D The NRC has approved the proposed activity or portions thereof in a license amendment or a safety evaluation, or is being reviewed by the NRC in a submittal that addresses the proposed activity. Implementa tion of change requires NRC approval. Reference the approval document or the amendment in review.:

D The proposed activity is administrat ively controlled by the Operating License (OL) or Technical Specificatio ns (TS).

Examples of programs and manuals controlled by the OL or TS are:

• Fire Protection Program (OL Condition) (EN-DC-128)

• Offsite Dose Calculation Manual (TS Administrat ive Controls)

• Surveillance Frequency Control Program (TS Administrat ive Controls) (EN-DC-355)

See NEI 96-07, Appendix E Section 2 for additional guidance on administrat ive controls.

Reference the administrat ive control(s): _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

[8] The proposed activity is controlled by one or more applicable regulations.

Examples of programs controlled by regulations that establish specific criteria are:

• Maintenanc e Rule (50.65) (EN-DC-203)

• Quality Assurance Program (10 CFR 50 Appendix B) o Security Plan [50.54(p)] (EN-NS-210)

• Cyber Security Plan [50.54(p)] (EN-NS-210)

• Emergency Plan [50.54(q)] (EN-EP-305)

• lnservice Inspection Program (50.55a) (EN-DC-351, -352)

• lnservice Testing Program (50.55a) (EN-DC-332)

See NEI 96-07 Section 4.1 for additional guidance on specific regulations.

Reference the controlling specific regulation(s ): 50.54(g)

IF the entire proposed activity can be dispositione d by one of the criteria in Section V, THEN 50.59 and 72.48 Screenings are not required. Proceed to Section VII and provide basis for conclusion in Section VII.A.

Otherwise, continue to Section VI to perform a 50.59 and/or 72.48 Screening, or perform a 50.59 and/or 72.48 Evaluation in accordance with EN-Ll-101 and/or EN-Ll-112.

Changes to the IPEC Unit 1 Decommiss ioning Plan are to be evaluated in accordance with the 50.59 process, as allowed by the NRC in a letter to IPEC dated January 31, 1996.

[Document ID: RA-96-014]

• EN-Ll-100 REV. 27

ATIAC HMENT9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 6 of 8 VI. 50.59 / 72.48 SCREENING REVIEW (All proposed activities must be evaluated to determine if 50.59, 72.48 or both apply. Check the applicable boxes)

VI.A 50.59 SCREENING D 50.59 applies to the proposed- activity, and all of the following 10 CFR 50.59 screening criteria are met; therefore, the proposed activity requires no further 50.59 review.

The proposed activity:

• Does not adversely affect the design function of an SSC as described in the UFSAR; AND

• Does not adversely affect a method of performing or controlling a design function of an SSC as described in the UFSAR; AND

• Does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC will be accomplished as described in the UFSAR; AND

• Does not involve a test or experiment not described in the UFSAR.

Document the basis for meeting the screening criteria in Section VI.C, then proceed to Section VII.

10 CFR 50.59 c 1 D The proposed activity does not meet the above criteria. Perform a 50.59 Evaluation in accordance with EN-Ll-101. Attach a co of the Evaluation to this form and roceed to Section VII.

VI.B 72.48 SCREENING D 72.48 applies to the proposed activity, and all of the following 10 CFR 72.48 screening criteria are met; therefore, the proposed activity requires no further 72.48 review.

The proposed activity:

• Does not adversely affect the design function of an SSC as described in the CFSAR; AND

• Does not adversely affect a method of performing or controlling a design function of an SSC as described in the CFSAR; AND

• Does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC will be accomplished as described in the CFSAR; AND

• Does not involve a test or experiment not described in the CFSAR.

Document the basis for meeting the screening criteria in Section VI.C, then proceed to Section VII.

10 CFR 72.48 c 1 D The proposed activity does not meet the above criteria. Perform a 72.48 Evaluation in accordance with EN-Ll-112. Attach a co of the Evaluation to this form and roceed to Section VII.

EN-Ll-100 REV. 27

ATTACHMENT9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 7 of 8 VI.C BASIS Provide a clear, concise basis for determining the proposed activity may be screened out such that a third-party reviewer can reach the same conclusions. Identify the relevant design function, as appropriate. Refer to NEI 96-07 Section 4.2 for guidance. Refer to NEI 12-06 Section 11.4 for guidance regarding FLEX. Provide supporting documentation or references as appropriate.

VII. REGULATORY REVIEW

SUMMARY

VII.A GENERAL REVIEW COMMENTS (Provide pertinent review details and basis for conclusions if not addressed elsewhere in form.)

These changes are governed by the Emergency Plan and have been reviewed under the 10CFR50.54( q) process per EN-EP-305 and have been found to have no adverse impact and no reduction in effectiveness of the Emergency Plan. The changes being implemented affect EIP 001, Classification of Emergencies only.

VII.B CONCLUSIONS

1. Is a change to an LBD being initiated? D Yes IF "Yes," THEN enter the appropriate change control process and include 00 No this form with the change package.

2. Is a 10 CFR 50.59 Evaluation required? D Yes IF "Yes," THEN complete a 50.59 Evaluation in accordance with EN-LI-101 00 No and attach a copy to the change activity.

3. Is a 10 CFR 72.48 Evaluation required? D Yes IF "Yes," THEN complete a 72.48 Evaluation in accordance with EN-LI-112 00 No and attach a copy to the change activity.

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 8 of 8 VIII. SIGNATURES 1 Preparer:

Reviewer:

pany / Department / Date Process Applicability Exclusion Site Procedure NA Champion or Name (print) / Signature / Company/ Department/ Date Owner:

Upon completion, forward this PAD form to the appropriate organization for record storage. If the PAD form is part of a process that requires transmittal of documentation, including PAD forms, for record storage, then the PAD form need not be forwarded separately.

• 1 The printed name should be included on the form when using electronic means for signature or if the handwritten signature is illegible. Signatures may be obtained via electronic authentication, manual methods (e.g., ink signature), e-mail, or telecommunication. Signing documents with indication to look at another system for signatures is not acceptable such as "See EC" or "See Asset Suite." Electronic signatures from other systems are only allowed if they are included with the documentation being submitted for capture in eB (e.g., if using an e-mail, attach It to this form; if using Asset Suite, attach a screenshot of the electronic signalure(s); if_ using PCRS, attach a copy of the completed corrective action).

EN-Ll-100 REV. 27

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 1 of 11 I. OVERVIEW PAD Rev.#: .Q Facility: Entergy - River Bend Station Proposed Activity / Document: ECs 75344. 75345 and 75346 - Seismic Monitorin g System Upgrade Change/Rev.#: Q Description of Proposed Activity:

Parent EC-75344, c1,long with Child ECs 75345 (online scope) and 75346 (outage scope) will upgrade the River Bend Seismic Monitoring System (System 557). The existing analog system is aging and replacement components are obsolete. To alleviate this issue, parts of this system will be replaced with a Syscom Instruments Seismic Monitoring System.

The Seismic Monitoring System can be separated into two divisions: an active portion and passive portion. The active portion currently consists primarily of Kinemetrics, Inc. components with one Engdahl Response Spectrum Recorder. The active portion is in place to provide signals back to the Main Control Room (MCR) Seismic Panel 1H13-P869 based on information obtained from field sensors. The passive portion, consisting of exclusively Engdahl recorders, is in place to provide recording functions only of any seismic activity.

This EC will upgrade the active portion only of the Seismic Monitoring System.

II. DOCUMENT REVIEW METHOD Provide the requested information for each item below.

1. For documents available electronically:

a. List search engine or documents searched, and keywords used:

Search Engine:

Autonomy was used to perform the LRS RBS search with the "50.59-SEARCHES",

"72.48 SEARCHES" options selected at 20% quality.

Keywords:

"Seismic Monitoring" (8 hits/ 1 relevant), "System 557" (19 hits), "Kinemetrics" (0 hits), "Engdahl" (0 hits), "seismic instrument" (360 hits/ 0 relevant), "accelerometer (7 hits/ 0 relevant)

"spectrum recorder (64 hits / O relevant)

NOTE: Correspondence from Licensing was sent January 3, 2019, after initial Autonom y search for this PAD was used, with the instructions to no longer use Autonomy for documen t searches.

As a result the keywords provided above were searched manually and the results are the same.

Additionally, "earthquake" was added as a key word search and relevant sections are updated below.

LRS Commitments were searched using the following keywords: Earthquake, Seismic.

b. List relevant sections of controlled electronic documents reviewed:

UFSART able 1.8-1 UFSAR Section 2.5.2.6, 2.5.2.7 & 2.5.4.9 UFSAR Section 3.7.4A Emergency Plan Section 13.3.6.3.1 Technical Requirements Manual Section 3.3.7.5 No relevant commitments were found.

EN-Ll-10 0, Rev. 25

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 2 of 11

2. Documents reviewed manually (hardcopy):

The River Bend Station Updated Safety Analysis Report, obtained from eB Reflib, was searched manually based on correspondence to no longer use Autonomy as a Licensing document search engine.

3. For those documents that are not reviewed either electronically or manually, use the specific questions provided in Sections Ill and IV of Attachment 9.2 of EN-Ll-100 as needed. Document, below, the extent to which the Attachment 9.2 questions were used.

The screening questions in Sections Ill and IV of EN-Ll-100, Attachment 9.2, were reviewed during the preparation of this Process Applicability Determination. No further documents were impacted beyond what is provided above.

Ill. PROCESS REVIEW Does the proposed activity affect, invalidate, or render incorrect, OR have the potential to affect, invalidate, or render incorrect, information contained in any of the following processes? Contact Program Owner if needed. Associated regulations and procedures are identified with each process below.

PROCESS (Regulations/ Procedures) YES NO REVIEW RES ULTS Chemistry/ Effluents D 121 Radwaste I Process Control Program (PCP) D 121 (EN-RW-105 or contact the Radiation Protection Dept.)

Radiation Protection/ ALARA D 121 (1 O CFR 20 / EN-RP-11 O or contact the Radiation Protection Dept.)

lnservice Inspection Program (1 O CFR 50.55a / EN-DC-333, -342, D 181

-351, -352) lnservice Testing Program (1 O CFR 50.55a / EN-DC-332) D 121 Maintenance Rule Program (1 O CFR 50.65 / EN-DC-203, -204, -205, -206, D 121

-207)

Containment Leakage Rate Testing (Appendix J) Program (1 O CFR 50 D 121 Appendix J / EN-DC-334)

FLEX Program (NRC Order EA-12-049/NRC Order EA-12-051/FLEX D [81 Program) (10 CFR 50.59 / EN-OP-201)

IF any box is checked "Yes," THEN contact the appropriate department to ensure that the proposed change is acceptable and document the results in the REVIEW RESULTS column.

EN-Ll-100, Rev. 25

ATIACHME NT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 3 of 11 IV. LICENSING BASIS DOCUMENT REVIEW Does the proposed activity affect, invalidate , or render incorrect, OR have the potential to affect, invalidate , or render incorrect, informatio n contained in any of the following Licensing Basis Documen t(s)? Contact LBD Owner if needed. Associate d regulation s and procedure s are identified with each Licensing Basis Documen t below.

LICENSING BASIS DOCUMENTS REVIEW RESULTS OR SECTIONS (Regulation s/ Procedures) YES NO AFFECTED OR LBDCR #

Quality Assurance Program Manual (OAPM)

[10 CFR 50.54(a), 10 CFR 50 Appendix B / EN-QV-104] D 181 Fire Protection Program (FPP) [includes the Fire Safety Analysis/Fire Hazards Analysis (FSA/FHA)]

OL Condition, 10 CFR 50.48 I EN-DC-128)

D 181 Emergency Plan (Includes the On-Shift Staffing Analysis)

Section 13.3.6.3.1 will require an

[10 CFR 50.54(q) / 10 CFR 50.47 / EN-EP-305/ EN-NS-220]

update to replace statements 181 D regarding the existing system.

See LBDCR # 2018-09 Environmental Protection Plan (Appendix B of the OL, Environmental Evaluation/ EN-EV-115, EN-EV-117, EN-Ll-103)

D 181 Security Plan

[1 O CFR 50.54(p) / EN-NS-210/ EN-NS-220 or contact site Security Dept.] D ~

Cyber Security Plan

[10 CFR 50.54 (p) /10 CFR 73.54 / EN-IT-103 or EN-IT-103-01] D 181 Operating License (OL) / Technical Specifications (TS)

(10 CFR 50.90 / EN-Ll-103) o* 181 TS Bases (10 CFR 50.59 / EN-Ll-100 / EN-Ll-101)

D ~

Technical Requirements Manual (TAM) (including TRM Bases)

Implementation of this activity will (10 CFR 50.59 / EN-Ll-100 / EN-Ll-101) require entry Into TLCO 3.0.3 for compensatory actions due to portions 181 D being installed online and others during an outage.

,Table 3.3.7.5-1 requires an update due to this activity. See LBDCR # 2018-08.

Core Operating Limits Report (COLA), and Pressure and Temperature Limits Report (PTLR) (TS Administrative Controls, EN-Ll-113, EN-Ll-100, EN-Ll-101)

D 181 Offsite Dose Calculation Manual (ODCM)

(TS Administrative Controls/ EN-Ll-113, EN-Ll-100) D 181 Updated Final Safety Analysis Report (UFSAR)

Various portions of Section 3.7.4A (10 CFR 50.71 (e) I EN-Ll-113, EN-Ll-100, EN-Ll-101) require an update to replace 181 D statements regarding the existing system.

See LBDCR # 03.07A-009 Storage Cask Certificate of Compliance (10 CFR 72.244 / EN-Ll-113) o* 181 Cask FSAR (CFSAR) (including the CTS Bases)

(10 CFR 72.70 or 72.248 / EN-Ll-113, EN-Ll-100,EN-Ll-112) D 181 10 CFR 72.212 Evaluation Report (212 Report)

(10 CFR 72.48 / EN-Ll-100, EN-Ll-112) D 181 NRC Orders (1 O CFR 50.90 / EN-Ll-103 or as directed by the Order) o* 181 NRC Commitments and Obligations (EN-Ll-110) o* 181 Site-Specific CFR Exemption (10 CFR 50.12, 10 CFR 55.11, 10 CFR 55.13, 10 CFR 72.7) o* 181

• contact the site Regulatory Assurance Department if needed.

IF any box is checked "Yes," THEN ensure that any required regulatory reviews are performed in accordance with the referenced procedures. Prepare an LBDCR per procedure EN-Ll-113

, as required, if a LBD is to be changed, and document any affected sections or the LBDCR #.

Briefly discuss how the LBD is affected in Section VII.A.

EN-U-*100, Rev. 25

ATTACHMENT9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 4 of 11 V. 10 CFR 50.59/ 10 CFR 72.48 APPLICABILITY Can the proposed activity be dispositioned by one or more of the following criteria? Check the appropriate box (if any).

D An approved, valid 50.59n2.48 Evaluation covering associated aspects of the proposed activity already exists. Reference 50.59ll2.48 Evaluation # (if applicable) or attach documentation. Verify the previous 50.59n2.48 Evaluation remains valid.

D The NRC has approved the proposed activity or portions thereof in a license amendment or a safety evaluation, or is being reviewed by the NRC in a submittal that addresses the proposed activity. Implementation of change requires NRC approval. Reference the approval document or the amendment in review.: _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

D The proposed activity is administratively controlled by the Operating License (OL) or Technical Specifications (TS).

Examples of programs and manuals controlled by the OL or TS are:

• Fire Protection Program (OL Condition) (EN-DC-128)

• Offsite Dose Calculation Manual (TS Administrative Controls)

• Surveillance Frequency Control Program (TS Administrative Controls) (EN-DC-355)

See NEI 96-07, Appendix E Section 2 for additional guidance on administrative controls.

Reference the administrative control(s): _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

D The proposed activity is controlled by one or more applicable regulations.

Examples of programs controlled by regulations that establish specific criteria are:

• Maintenance Rule (50.65) (EN-DC-203)

• Quality Assurance Program (1 O CFR 50 Appendix 8)

• Security Plan [50.54(p)] (EN-NS-210)

• Cyber Security Plan [73.54] (EN-IT-103)

• Emergency Plan [50.54(q)] (EN-EP-305)

• lnservice Inspection Program (50.55a) (EN-DC-351, -352)

• lnservice Testing Program (50.55a) (EN-DC-332)

See NEI 96-07 Section 4.1 for additional guidance on specific regulations.

Reference the controlling specific regulation(s): - - - - - - - - - - - - - - - - -

IF the entire proposed activity can be dispositioned by one of the criteria in Section V, THEN 50.59 and 72.48 Screenings are not required. Proceed to Section VII and provide basis for conclusion in Section VII.A.

Otherwise, continue to Section VI to perform a 50.59 and a 72.48 Screening, or perform a 50.59 and/or 72.48 Evaluation in accordance with EN-Ll-101 andfor EN-Ll-112.

Changes to the IPEC Unit 1 Decommissioning Plan are to be evaluated in accordance with the 50.59 process, as allowed by the NRC in a letter to IPEC dated January 31, 1996. [Merlin Document ID: RA-96-014]

EN-Ll-100, Rev. 25

ATTACHMEN T9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 5 of 11 VI. 50.59 / 72.48 SCREENING REVIEW (All proposed activities must be evaluated to determine if 50.59, 72.48 or both apply. Check the applicable boxes)

VI.A 50.59 SCREENING l'8I 50.59 applies to the proposed activity, and all of the following 10 CFR 50.59 screening criteria are met; therefore, the proposed activity requires no further 50.59 review.

The proposed activity:

• Does not adversely affect the design function of an SSC as described in the UFSAR; AND

• Does not adversely affect a method of performing or controlling a design function of an SSC as described In the UFSAR; ANQ

• Does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC will be ac*complished as described in the UFSAR; AND

• Does not Involve a test or experiment not described in the UFSAR .

Document the basis for meeting the screening criteria in Section VI.C, then proceed to Section VII.

f10 CFR 50.59(c){1 )1 D The proposed activity does not meet the above criteria. Perform a 50.59 Evaluation in accordance with EN*Ll-101. Attach a coov of the Evaluation to this form and oroceed to Section VII.

VI.B 72.48 SCREENING D 72.48 applies to the proposed activity, and all of the following 10 CFR 72.48 screening criteria are met; therefore, the proposed activity requires no further 72.48 review.

The proposed activity:

• Does not adversely affect the design function of an SSC as described in the CFSAR; AND

• Does not adversely affect a method of performing or controlling a design function of an SSC as described in the CFSAR; AND

• Does not adversely affect a method of evaluation that demonstrate s intended design function(s) of an SSC will be accomplishe d as described in the CFSAR; AND

• Does not involve a test or experiment not described in the CFSAR .

Document the basis for meeting the screening criteria in Section VI.C, then proceed to Section VII.

f10 CFR 72.48(c)(1 ll D The proposed activity does not meet the above criteria. Perform a 72.48 Evaluation in accordance with EN-Ll-112. Attach a coov of the Evaluation to this form and oroceed to Section VII.

EN-Ll-100, Rev ..25

. ATTACHMENT9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 6 of 11 VI.C BASIS Provide a clear, concise basis for determining the proposed activity may be screened out such that a third-party reviewer can reach the same conclusions. Identify the relevant design function, as appropriate. Refer to NEI 96-07 Section 4.2 for guidance. Refer to NEI 12-06 Section 11.4 for guidance regarding FLEX. Provide supporting documentation or references as appropriate.

This activity, Parent EC-75344 and Child ECs 75345 & 75346, will upgrade the Seismic Monitoring System (System 557). As stated previously, this EC will only upgrade the active portion of the system with a Syscom Instruments Seismic Monitoring System. The design function of this system is to detect, measure and provide indication to plant personnel in the event of seismic activity. This system does not have any Safety-Related design functions and *is not credited in the Accident analysis. This activity is an analog to digital upgrade, as the existing Kinemetrics and Engdahl components are analog, while the replacement equipment provided by Syscom is digital. This is a standard, proven system supplied by Syscom with previous nuclear site installations, including at Arkansas Nuclear One and Grand Gulf.

As shown on drawing 0210.860-220-013, the active portion currently consists of field instruments with connections back to MCR seismic monitoring panel 1H13-P869. The field instruments are made up of four (4) triaxial accelerometers (1 ERS-NBE1 A, 1 B, 1C, 1D), a seismic trigger (1 ERS-NBS4A) and a seismic switch (1 ERS-NBS4B), all of which are manufactured by Kinemetrics. The triaxial accelerometers detect acceleration. The seismic trigger and switch packages are made up of three accelerometers to sense accelerations in three directions and differ primarily by their sensitivities. The trigger is set to function when 0'°1g acceleration or greater is reached and will provide a SEISMIC TAPE RECORDING SYSTEM START alarm in the Control Room (see UFSAR Section 3.7.4.3A.c). The switch will function when 0.083g is reached on the vertical axis or 0.082g is reached on the horizontal axes and will provide a SEISMIC EVENT HIGH alarm in the Control Room (see UFSAR Section 3.7.4.3A.a).

There is also a Response Spectrum Recorder (1 ERS-NBR2D), manufactured by Engdahl, which provides indication at the seismic panel annunciator (1ERS-NB1101 ). This annunciator has predetermined acceleration limits, making up the response spectrum of 1 to 32Hz. The annunciator has three banks of indicator lamps, one for the vertical axis and two for the horizontal axes. Each bank features two sets of lights, amber and red.

The amber lights indicate that accelerations are approaching design limits for Operating Basis Earthquake (OBE) while the red lights indicate accelerations are exceeding design limits for Safe Shutdown Earthquake (SSE) in a given frequency (see UFSAR Section 3.7.4.3A.b). Connections between the response spectrum recorder and annunciator are made at the seismic panel junction box, (1ERS-PNL103).

Also included in seismic panel 1H13-P869 is two recording panels (1 ERS-PNL3A and 38), each with two magnetic tape recording drive units. The recording drive units record on four separate channels (based on input from the four accelerometeri;;) and are only actuated when activity is sensed by the seismic trigger.

There is a control unit (1 ERS-NBl102) in place for the recording panels which provides event status and other control functions for the recorder units. A playback panel (1 ERS-PNL3C) is included in P869 to provide play-back of information recorded on the magnetic tape recorders. Finally, there is also a seismic switch power supply (1 ERS-PNL3D) which not only serves as the power supply, but also functions as a test panel for the seismic switch. All of this equipment and their functionalities are being replaced and upgraded by the new Syscom Seismic System.

The new seismic monitoring system is based on Syscom's Marmot system. Equipment included in this system include: tour (4) MR2002-SM24K recorders with internal MS2008+ triaxial accelerometers, one (1) NCC2002 Network Control Center with an indicator panel (annunciator), two (2) power supply units, and a rack-mount computer with screen and keyboard. The new MR2002 recorders are designed to record seismic activity similar to the existing recorders; however, they have the added ability to calculate Cumulative Absolute Velocity (GAV), or seismic intensity. These devices, along with the new accelerometers, will replace these existing accelerometers at their exact current locations. The new recorder/accelerometer packages will be connected back to the network control center. This control center is designed to provide the interconnections to the recorders and coordinate all of the recorders' activities. The network control center will be configured such that the alarm levels based on OBE and SSE design limits of 0.05g and 0.1 g, respectively, which are obtained from the recorders and will replace the existing seismic trigger and switch. The computer will be in place to provide the interface to the network control center and recorders/accelerometers.

EN-Ll-100, Rev. 25

ATIACHM ENT9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 7 of 11 As noted on drawing 0210.860-220-013, there are currently four locations where the field devices are installed:

1. Reactor Building Mat, EL. 70'-0" - contains sensor (1 ERS-NBE1 A), recorder (1 ERS-NBR 2D) and seismic switch (1 ERS-NBS4B).

2. Drywell, EL. 151 '-0" - contains sensor (1 ERS-NBE1 C)

3. Reactor Building, EL. 233'-0" - contains sensor (1 ERS-NBE1 B)

4. Free Field - contains sensor (1 ERS-NBE1 D) and seismic trigger (1 ERS-NBS4A)

Each of these locations will have their components replaced by a Syscom recorder and sensor package with no individual, separate triggers/switches. Note that the 1 ERS-NBR2D will remain in place as a passive device (i.e. indications to Control Room will be removed, detail provided below).

Currently, UFSAR Table 1.8-1 provides the commitment to USNRC Regulato ry Guide 1.12, Rev. 1, "Instrumentation for Earthquakes." Included in this commitment are clarifications which determine that it is acceptable to use the existing response spectrum recorders despite having a frequenc y range of 1 to 32Hz.

The NRC has requested a range of 1 to 33Hz. It is explained that this is due to the design response spectra for seismic design of nuclear power plants (Regulatory Guide 1.60) provides for a response spectra covering 0.1 to 33 Hz. This commitment further explains that the difference between 32Hz and 33Hz is indistinguishable and that the existing sensors and peak accelerographs cover a range of Oto 50Hz. As stated previously, only the active portion of the existing Seismic Monitoring System is being replaced. The passive portion of the system (explicitly described in RG 1.12, Rev. 1) also contains the response spectrum recorders described in this commitment. Although the recorder in the active system is being replaced, those in the passive system will remain, as will the peak accelerographs. The new sensors of the Syscom system are capable of covering a range of Oto 600Hz. Therefore, this commitment is not impacted by this design activity.

Further, the Syscom Seismic Monitoring system meets the requirements of Regulato ry Guide 1.12, Rev. 1, and this commitment is not required to change.

Section 3.7.4A of the UFSAR provides a description of the existing instrumentation of the Seismic Monitoring System. Generally speaking, there are portions of this section that will require a change based on the components of the new system; however, the functions and setpoints will either be enhanced or remain unchanged. Items of note that will change in this section are described below.

Section 3.7.4.2A of the UFSAR describes the existing instrumentation as follows:

"The strong motion triaxial accelerographs to be installed have the following physical characteristics:.

1. Accelerometers are the transducer-type with the capability of recording a maximum scale. of 1.0 g at full

2. Accelerometers are sensitive to frequencies in the range of 0.1 to 50 Hz.

3. The seismic instrumentation and recording system is in a quiescent state until activated by seismic triggers which are set at 0.01 g. These seismic triggers (both horizontal and vertical) activate the recording system in less than 100 ms. Recording continues until the level of motion drops below 0.01 g.

4. The recording system is powered by internal batteries with trickle charge from 110 VAC capable of recording up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> upon loss of normal power..

5. Each sensor package contains three mutually orthogonal accelerometers. All four sensor packages are oriented to the same azimuths.

6. Recording of the electrical signals from the accelerometers is by magnetic tape with the acceleration signal and the time signal occupying separate tracks on the tape."

The replacement accelerometers address each of these characteristics as follows:

1. The new accelerometers are capable of recording up to 4.0g, an increase in overall range.

2. As noted earlier, the new accelerometers cover'O to 600Hz, an increas13 in frequency range which bounds the existing conditions.

3. The seismic trigger function will remain as is. The trigger is located in the free field and the new equipment will be set to the same setpoint of 0.01 g to start system recording.

This information is transmitted real-time to the network control center. The new recorders continuou sly record to a ring EN-Ll-100, Rev. 25

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 8 of 11 buffer with pre- and post-event time history recordings, which is user selected from 1-1 ODs. This data is available to be recorded to permanent memory.

4. All new components are supplied by normal AC power from an uninterruptible source. In addition, the new components are capable of having battery backed power in the event of loss of normal power and will continue operating up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

5. The new system has a single sensor unit at each location; however this is a triaxial accelerometer capable of measuring in the three directions and will be oriented to the same azimuths.

6. The magnetic tapes will be removed and the new system will provide real-time output that shows clear delineation when QBE and SSE limits are exceeded. Additionally, the new system is capable of calculating GAV.

Further in Section 3.7.4.2A, the seismic switch is described as having a frequency range of 0.1 to 33Hz and that the trigger levels are set to 0.082g horizontally and 0.083g vertically for QBE and also signals the SEISMIC EVENT HIGH annunciator on main control board H13-P680. Similar to the seismic trigger, the function of this device will be replaced by the individual recorder at the same location with these limits user selectable. It should be noted, Sections 2.5.2. 7 and 2.5.4.9 of the UFSAR provide the definition for earthquake design basis as 0.05g QBE. Therefore, the trigger levels of 0.083g and 0.082g will be changed to 0.05g and section 3.7.4.2A will be updated accordingly.

Section 3.7.4.3A of the UFSAR provides a description of the MCR alarms of the Seismic Monitoring System.

The annunciator panel (1ERS-NBl101) on the Seismic Monitoring cabinet, which takes inputs from the response spectrum recorder, will be completely removed. Recall that this annunciator panel has amber (QBE) and red (SSE) indication lights that cover a 1 to 32Hz frequency spectrum. This is not a requirement provided in Regulatory Guide 1.12, Rev. 1 and is acceptable to remove. However, ERS-NBl101 directly provides the signals for the SEISMIC EVENT HIGH/HIGH annunciator on main control board H13-P680. In other words, the HIGH/HIGH alarm receives its input based on either QBE or SSE response spectra exceeded limits. Due to the OBE spectra providing both HIGH/HIGH and HIGH.annunciations, and both setpoints being different, the HIGH/HIGH indication will be updated based on 0.1 g SSE (UFSAR Section 2.5.2.6 and 2.5.4.9 defined limit) response spectra.

As noted previously, the network control center coordinates the activities of the field recorders. This device can be customized to alarm and indicate, based on selected seismic levels, of each recorder. Recall that there are also indication lamps located on Control Panel P680 that contain 'SEISMIC EVENT HIGH' and

'SEISMIC EVENT HIGH/HIGH' indications as well as 'TAPE RECORDING START'. The new network control center has up to four (4) relay outputs to provide signals for alarms and is capable of maintaining these indications. It should be noted that the "TAPE RECORDING START' indication will be changed to 'SEISMIC SYS RECORDING / TROUBLE' due to the removal of the magnetic tapes as well as the addition of the

'Warning/Error indicators on the NCC which track loss of communication or loss of power with the field recorders.

Section 3.7.4.4A of the UFSAR describes the comparison of measured responses to predicted responses.

This section will mostly require an update to change the types of instrumentation involved in making these comparisons. As stated, the magnetic tapes are removed, so these items will no longer be used in making these comparisons. Real-time outputs of seismic activity from the network control center are clear, user-friendly reports which show clear OBE and SSE exceedance during the response spectra. Further, as discussed previously, the new system is capable of calculating GAV, or seismic intensity, in which the existing system is not capable.

Due to the Seismic Monitoring System being a standalone system (it provides no direct input to other systems), highest level failure modes are limited to faulty recordings at the network control center. The existing setpoints used to trigger recording will also be used for the new system and all new equipment will be installed at the same locations as the existing equipment. Therefore, failure modes associated with the new system will be similar to those of the existing system. The recorders and network control center are equipped with batteries that, in the event of loss of AC power, these devices are capable of running for an extended period up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

EN-Ll-100, Rev. 25

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 9 of 11 CR-RBS-2012-01283 was written in response to INPO Event Report IER L2-12-12. In this report, three lessons learned were identified:

1. Ensure seismic recording equipment is available to perform its design and licensing functions (recording s

and indications) and to support timely processing and evaluation of data to determine the intensity of a seismic event. Actions taken to determine if the seismic event exceeded OBE and to address the loss of power should include the following:

a. Identify minimum equipment required, including the need for free-field seismic instrumentation.

b. Ensure that seismic monitoring and alarms in the main control room function on a loss of offsite power.

2. Develop a long-term asset management plan for obsolete seismic monitoring equipment. Consider that the equipment should be capable of calculating a CAV from free-field measurements and that the equipment should be fully supported and serviced.

3. Determine if seismic abnormal operating procedures and engineering evaluation procedure s are adequate for operators and engineers to respond to a seismic event, including specific equipment and structures to walk down post-event, based on the guidance contained in AG 1.166, AG 1.167 and EPRI NP-6695.

It was determined for IER lesson learned #1 that River Bend is not susceptible to the issues described by this lesson. The new system will also not be susceptible to this issue. As stated, the new system is in compliance with Regulatory Guide 1.12, Rev. 1. Also, the new system will be powered by the same uninterrup tible power source that is currently feeding the existing system and the new components all have battery backup capabilities. For IER lessons learned #2 and 3, it was determined that River Bend is currently susceptibl e to these issues. However, as stated, the new system is capable of calculating CAV, therefore this item will no longer be deficient. In regards to IER lesson learned #3, River Bend has already taken actions to update Abnormal Operating Procedure AOP-0028, and Engineering Procedures STP-557-3700 and STP-704-3 30 to resolve the gaps with the Regulatory Guides and EPRI report. These procedures are impacted by this activity only to ensure that the new system is incorporated.

The proposed change will not require a change to the RBS Technical Specifications, Technical Specification Basis. Changes to the UFSAR, Technical Requirements Manual and Emergency Plan will be limited to the removal of existing components that will no longer be installed and addition of new components.

EN-Ll-100, Rev. 25

ATTACHMENT 9.1 PROCESS APPLICABILITY DETERMINATION FORM Page 10 of 11 This activity does not adversely affect the design function of an SSC as described in the UFSAR.

- The implementation of ECs 75344, 75345 and 75346 for the upgrade of the Seismic Monitoring System will not adversely affect operation or function of any component, system, or structure of the Seismic Monitoring System as described in the UFSAR or any other Licensing Basis Document. The existing setpoints will be used to trigger the same annunciations to determine an SSE or OBE. Although the 1-32 Hz response spectrum annunciator panel will be removed, the new system is capable of providing a real-time, user friendly report which shows clear and concise delineation when SSE and OBE levels are exceeded. The new system will perform the same functions as the existing and will add the ability to calculate Cumulative Absolute Velocity.

This activity does not adversely affect a method of performing or controlling a design function of an SSC as described in the UFSAR.

As noted previously, the existing setpoints used to trigger annunciations will change, however, the setpoints will be aligned with the UFSAR design basis for SSE or OBE limits. Additionally, inclusion of the ability to calculate Cumulative Absolute Velocity further enhances the system output. There are no automatic functions of this system as outputs are limited to MCR indications with which Operators make decisions to shut down the plant.

This activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC will be accomplished as described in the UFSAR.

A review of the UFSAR sections in this PAD (see Section II) identified no method of evaluation relevant to demonstrating the design functions of the affected SSCs that would be adversely affected by this activity. No safety analyses are impacted by this activity.

• This activity does not involve a test or experiment not described in the UFSAR.

The proposed activity installs permanent equipment which does not alter the basis of operation for the affected SSCs or the manner of use of an SSC. The installation, testing, operation, and maintenance of the new Seismic Monitoring System will be conducted in accordance with approved plant procedures.

EN-Ll-100, Rev. 25

ATTACHMENT 9.1 PROCESS APPLICABILIT Y DETERMINATION FORM Page 11 of 11 VII. REGULATORY REVIEW

SUMMARY

VII.A GENERAL REVIEW COMMENTS (Provide pertinent review details and basis for conclusions if not addressed elsewhere in form.)

See Section VI.C for all pertinent details. The proposed activity is scheduled for implementation with the plant at power (Child EC-75345) and during an outage (Child EC-75346). LBDCR #'s 03.07A-009, 2018-09 and 201 s~os are issued to identify changes associated with the UFSAR, Emergency Plan, and Technical Requirements Manual, respectively. Implementation of this activity requires online and outage scopes, therefore rendering parts of the system inoperable. The Technical Requirements Manual, Section 3.3.7.5, provides the action to enter TLCO 3.0.3 which requires compensatory actions as described in AOP-0028 and duty manager approval.

This activity does not affect any structures, systems, or components controlled by 10 CFR 72.48.

VII.B CONCLUSIONS

1. Is a change to an LBD being initiated? [81 Yes IF "Yes," THEN enter the appropriate change control process and include this form with the change package.

D No

2. Is a 10 CFR 50.59 Evaluation required?

D Yes IF "Yes," THEN complete a 50.59 Evaluation in accordance with EN-Ll-101 [81 No and attach a copy to the change activity.

3. Is a 10 CFR 72.48 Evaluation required?

D Yes IF "Yes," I!:!.§:! complete a 72.48 Evaluation in accordance with EN-Ll-112 [81 No and attach a copy to the change activity.

VIII. SIGNATURES 1 Preparer: Shahid Ali/ See AS/ DP Engineering LTD. CO./ Des. Elec-l&C / 1O Jan 2019 Name (print)/ Signature I Company I Department I Date Reviewer: Gary Yezefski / See AS/ DP Engineering LTD. CO./ Des. Elec-l&C / 10 Jan 2019 Name (print)/ Signature/ Company/ Department I Date Process Applicability Exclusion Site Procedure _ N - - ' / _ A ' - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

Champion or Name (print)/ Signature I Company I Department I Date Owner:

Upon completion, forward this PAD form to the appropriate organization for record storage. If the PAD form is part of a process that requires transmittal of documentation, including PAD forms, for record storage, then the PAD form need not be forwarded separately.

1 The printed name, company, department, and date must be included on the form. Signatures may be obtained via electronic processes (e.g., PCRS, ER processes, Asset Suite signature), manual methods (e.g., ink signature), e-mail, or telecommunication. If using an e-mail, attach it to this form.

EN-Ll-100, Rev. 25

ATTACHMENT 9.2 10 CFR 50.54(a)(3) SCREENING SHEET1 OF4 Procedure/Document Number: EIP-2-001 Revision: 2B Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies Part I. Description of Activity Being Reviewed (This is generally changes to the emergency plan, EAL.s, EAL bases, etc. - refer to step 3.0(61):

Revise EIP-2*001, Classlficalion of Emergencies as listed below:

1) Attachment 4 Page 27, Attachment 9 User Aid 1 Page 143 and Attachment 9 User Aid 2 Page 144:

From: Seismic event identilied by any 2 of the following;

• Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Annurciato< "Selsm:c Tape Rocorcfing SYS Start" (P680-02A-DCl6)

b. Event Indicator on ERS*NBl-102 ls white To: Seismic event idenHfied by any 2 of the following:

Seismic event confirmed by activated seismic switch as Indicated by receipt of EITHER a OR b:

a Annl.flciator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A*D06)

b. Event Indicator on EAS*NBR3D TRIGGER (RECORD START) Is yellow

2) Attachment 4 Page 27, Attachment 9 User Ald 1 Page 143 and Attachment 9 User Aid 2 Page 144:

From: a. Seismic event> Operating Basis Earthquake (OBE) as Indicated by:

Annllnclator "Seismic Tape Recording System Start" (P680*02A-D06}

AND Event Indicator on EAS-NBl-102 is white AND Receipt of EITHER 1 OR 2;

1. Annunciator "Seismic Event High" (P6B0-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-02A*B06) ANO amber light(s) on panel NBl-101 as To: a Seismic event> Operating Basis Earthquake (OBE} Indicated by; Annunciator "SEISMIC .SYS RECORDING I TROUBLE" (P680-02A-D06)

ANO .

ERS-NBR3D TRIGGER (RECORD START} is yellow AND Receipt ol ~ 1 OR 2;

1. Annunciator "Seismic Event High" (P6B0-02A*C06)

2. Annunciator "Seismic Event High-High" (P6B0*02A*B06} ~ ERS*NBR3A OBE (HI) yellow light

3) Attachment B Page 86:

From: 1. Seismic event ldentilied by any 2 of the following:

' Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b

a Annunciator "Seismic Tape Recordng SYS Start" (P680-02A*D06)

b. Event Indicator on ERS*NBl-102 is white To: 1. Seismic event Identified by any 2 of the following:

0 Seismic event conflnned by activated seismic switch as Indicated by receipt of EITHER a OR b:

a Annl61Cialor "SEISMIC SYS RECORDING /TROUBLE" (P680-02A*DCl6)

b. Event Indicator on EAS-NBR3D TRIGGER (RECORD START) Is yellow

4) Attachment B Page 87:

From: The annunciators "Seismic Tape Recording SYS Start" and the '\>mite" event indicator are listed in the Alarm Response Procedure as verification of an earthquake event

• To: The annunciators "SEISMIC SYS RECORDING/TROUBLE" and the "yellow' event lrdicator are listed in the Alann Response Procedure as verification of an earthquake event 5} Attachment 8 Page 96:

~ a. Seismic event > Operating Basis Earthquake (OBE} as Indicated by:

• Annunciator "Seismic Tape Recording System Start" (P680-02A-DD6}

AND Event Indicator on EAS*NBl-102 is while AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (PS8D*02A*C06)

2. Annunciator "Seismic Event Hioh-Hioh" /P680*02A-B06) AND amber liohl/s) on oanel NB1*101 EN-EP-305 REV 8 I_

ATTACHMENT9.2 10 CFR 50.54(0)(3) SCREENING SHEET20F4 Procedure/Document Number: EIP-2-001 I Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies To: a. Seismic event > Operating Basis Earthquake (QBE) as Indicated by:

• Annunciator "SEISMIC SYS RECORDING /TROUBLE" (P680-02A*D06)

AND EAS*NBR30 TRIGGER (RECORD START) Is yellow AND Receipt of EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680*02A*C06)

2. Annunciator "Seismic Everit High-High" (P680*02A*B06) A!!1Q ERS-NBR3A OBE (HI) yellow light Part II. Activity Previously Reviewed? DYES @NO 50.54(q)(3) Continue to Evaluation Is next part NOT required.

Is this activity fully bounded by an NAC approved 10 CFR 50.90 submittal or Enter Alert and Notification System Design Report? Justification below and If YES, identify bounding source document number/approval reference and complete Part VJ.

ensure the basis for concluding the source document fully bounds the proposed change is documented below:

Justification:

D Bounding document attached (optional}

Part Ill. Applicability of Other Regulatory Change Control Processes Check if any other regulatory change processes control the proposed activily.(Refer to EN-Ll-100)

APPLICABILITY CONCLUSION 2J If there are no other controlling change processes, continue the 50.54(q)(3) Screening.

D One or more controlling change processes are selected, however, some portion of the activity involves the emergency plan or affects the implementation of the emergency plan; continue the 50.54(q)(3) Screening for that portion of the activity. Identify the applicable controlling change processes l:>elow.

D One or more controlling change processes are selected and fully bounds all aspects of the activity. S0.54(q)(3)

Evaluation Is NOT reouired. Identify controllino chance processes below and comolete Part VI.

CONTROLLING CHANGE PROCESSES 10CFR50.54(q)(3)

Part IV. Editorial Change DYES @NO 50.54(q}(3}

Is this activity an editorial or typographical change such as formatting, paragraph Continue lo next Evaluation is pan numbering, spelling, or punctuation that does !JOI change intent? NOT required.

Justification: Enter justification and c:onlinuc lo next part or complete Pan Vias applicable.

EN-EP-305 REV 8

ATIACHMENT9.2 10 CFA 50.54(a)(3) SCREENING SHEET30F4 Procedure/Document Number: EIP-2-001 Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies Part V. Emergency Planning Element/Function Screen (Associated 10 CFR 50.47(b) planning standard function identified in brackets) Does this activity affect any of the following, including program elements from NUREG*

0654/FEMA REP-1 Section II?

1. Responsibility for emergency response is assigned. [1 J D

2. The response organization has the staff to respond and to augment staff on a continuing basis (24/7 D staffing) in accordance with the emergency plan. [1 J

3. The process ensures that on shift emergency response responsibilities are staffed and assigned. [2] D

4. The process for timely augmentation of onshift staff is established and maintained. [2] D

5. Arrangements for requesting and using off site assistance have been made. [3J D

6. State and local staff can be accommodated at the EOF in accordance with the emergency plan. [3] D

7. A standard scheme of emergency classification and action levels is in use. (4) ~

8. Procedures for notification of State and local governmental agencies are capable of alerting them of D the declared emergency within 15 minutes after declaration of an emergency and providing follow-up notifications. [5]

9. Administrative and physical means have been established for alerting and providing prompt D instructions to the public within the plume exposure pathway. [5]

10. The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and D Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. [5)

11. Systems are established for prompt communication among principal emergency response D organizations. [6]

12. Systems are established for prompt communication to emergency response personnel. [6] D

13. Emergency preparedness information is made available to the public on a periodic basis within the D plume exposure pathway emergency,planning zone (EPZ). [7]

14. Coordinated dissemination of public information during emergencies is established. [7] D

15. Adequate facilities are maintained to support emergency response. [8] D

16. Adequate equipment is maintained to support emergency response. [8] u

17. Methods, systems, and equipment for assessment of radioactive releases are in use. [9] 0

18. A range of public PARs is available for implementation during emergencies. [10] D

19. Evacuation time estimates for the population located ln the plume exposure pathway EPZ are D available to support the formulation of PARs and have been provided to State and local governmental authorities. [1 O]

20. A range of protective actions is available for plant emergency workers during emergencies, including D those for hostile action events.[10)

21. The resources for controlling radiological exposures for emergency workers are established. [11] D

22. Arrangements are made for medical services for contaminated, injured individuals. [12] D

23. Plans for recovery and reentry are developed. (13] D

24. A drill and exercise program (including radiological, medical, health physics and other program D areas) fs established. (14]

25. Drills, exercises, and training evolutions that provide performance opportunities to develop, D maintain, and demonstrate key skills are assessed via a formal critiaue process in order to identify EN-EP-305 REV 8

ATIACHMENT 9.2 10 CFR 50.54(0) (3) SCREENING SHEET4 0F4 Procedure/Document Number: EIP-2-001 I Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies weaknesses. (14]

26. Identified weaknesses are corrected. (14)

D

27. Training is provided to emergency responders. (15)

D

28. Responsibility for emergency plan development and review is establish ed. (16] D

29. Planners responsible for emergency plan development and maintenance are properly trained. (16)

APPLICABILITY CONCLUSION

J II no Part V criteria are checked, a 50.54(q)(3) Evaluation is NOT required:

document the basis for conclusion below and complete Part VI.

@ II any Part V criteria are checked, complete Part VI and perlorm a 50.54(q)(3)

Evaluation.

BASIS FOR CONCLUSION Emergency planning element 7 (10 CFR 50.47(b) planning standard function 4) in Part V of this form is affected by changes to seismic instrumentation which resulted in changes 1 - 5 in Part 1. A 10CFRS0.54 (q)(3) evaluation will be performed to delermine whether or not the effectiveness of the emergency plan is reduced and prior NRG approval is required.

Part VI. Signatures:

Preparer Name (Print) Preparer Signature Date:

Norman E Tison 1/20/20 (Optional) Reviewer Name (Print)_,.,.,* Date:

James J Lewis (

Reviewer Name (Print)

/-'Jt>--/

Date:

Aaron Magee Nuclear EP Project Manager /-2{} ..,/9 Approver Name (Print)

Oats:

T. W.Gates Manager, Emergency Planning or designee EN*EP-305 REV 8

Attachment 3 Page 1 of 4 Procedure/Document Number: EIP-2-001 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies Part I. Description of Proposed Change: Revise EALs HU6 and HAS to reflect outputs of new seismic monitoring instrumentation. Specific changes !isled below:

Revise EIP-2-001, Classification of Emergencies as listed below:

1) Attachment 4 Page 27, Attachment 9 User Aid 1 Page 143 and Attachment 9 User Aid 2 Page 144:

~ Seismic event Identified by any 2 of the following:

• Seismic event confirmed by activated seismic switch as indicated by receipt of EITHER a OR b:

a. Anmn:iatOI' "Seismic Tape Reconfng SYS Slart" (P680-02A-D06)

b. Event Indicator on ERS-NBl*102 ls white To: Seismic event Identified by any 2 of the following:

Seismic event confirmed by activated seismic switch as Indicated by receipt of EITHER a OR b:

a Anm.roator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A*D06)

b. Event Indicator on ERS*NBR30 TRIGGER (RECORD START) is yellow

2) Attachment 4 Page 27, Attachment 9 User Aid 1 Page 143 and Attachment 9 User Aid 2 Page 144:

From: a. Seismic event > Operating Basis Earthquake (OBE) as indicated by:

Annunciator "Seismic Tape Recording System Start* (P680-02A-006)

AND Event Indicator on ERS-NBl-102 is white aMQ Receipt ol fil!:ilIB 1 OR 2:

1. Annunciator "Seismic Event High" (P680*02A-C06)

2. Annunciator "Seismic Event High*High" (P680*02A*B06) A!:iQ amber llght(s) on panel NB1*101 To: a Seismic event> Operating Basis Earthquake (OBE) as indicated by:

Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P660-02A-D06)

Alm ERS-NBR30 TRIGGER (RECORD START) Is yellow AND Receipt or EITHER 1 OR 2:

1. Annunciator "Seismic Event High" (P680-02A*C06)

2. Annunciator "Seismic Event High-High" (P6B0*02A*B06) AND ERS-NBR3A OBE (HI) yellow light

3) Attachment B Page 86:

From: 1. Seismic event Identified by any 2 of the following:

• Seismic event confirmed by activated seismic switch as Indicated by receipt of EITHER a OR b:

a Amlllciator "Seismic Tape Recording SYS Start" (P680-02A*D06)

b. Event Indicator on ERS-NB1*102 ls white To: 1. Seismic event Identified by any 2 of the lollowing:

• Seismic event confirmed by activated seismic switch as Indicated by reoeipt of EITHER a QB b:

a. Anm.nciator "SEISMIC SYS RECORDING /TROUBLE" (P680-02A-D06)

b. Event Indicator on ERS*NBR30 TRIGGER (RECORD START) ls yellow

4) Attachment B Page 87:

From: The annunciators "Seismk:: Tape Recorcing SYS Slart" and the 'Wlile" event Indicator are rlSled in the Alarm Response Procedure as veriflcalion of an earthquake evenl

~ The annunciators "SEISMIC SYS RECORDING/ TROUBLE" and the "yellow" event lncftealor are lis1ed In lhe Alam, Respcnse Procedure as verification of an earthquake event

5) Attachment 8 Page 96:

From: a. Seismic event > Operating Basis Earthquake (OBE) as Indicated by; .

Annunciator "Seismic Tape Recording System Start* (P680*02A-D06)

AND Event Indicator on ERS-N81*102 ls white AND Receipt of EITHER 1 OR 2:

• 1. Annunciator "Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (PB80*02A-806) AND amber light(s) on panel NBl-101 EN-EP-305 ROOS

Attachment 3 Page 2 of 4 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: EIP-2-001 I Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies To: a. Seismic event > Operating Basis Earthquake {OBE} as indicated by: .

Annunciator "SEISMIC SYS RECORDING/ TROUBLE" (P680-02A-D06)

AND ERS-NBR30 TRIGGER (RECORD START) Is yellow AND Receipt of EITHER 1 Q!! 2:

1. Annunclator"Seismic Event High" (P680-02A-C06)

2. Annunciator "Seismic Even! High-High" (P680-02A*B06) Mil! ERS-NBR3A OBE (HI) yellow light Part II. Description and Review of Licensing Basis Affected by the Proposed Change:

In accordance with EN-Ll-100, a Process Applicability Determination (PAD) was performed to review the proposed changes against the licensing Basis Documents (LBDs). As part of the PAD Manual electronic search of Technical Requirements Manual, Technical Specifications, Updated Safety Analysis Report was done using keywords:

'Emergency Action Level' 'HUS' 'HAS'- no hits resulted.

Search of the Emergency Plan using key words 'Emergency Action Level' 'HU6' 'HA6' had relevant hits on 'Emergency Action Level' in Sections 13.3.1.1 Definitions, 13.3.3.1 Classification System, HU6 and HA6 in Table 13.3-1. Key word search using seismic had relevant hits in Section 13.3.6.3.1 Onsite Assessment Facilities.

The following sections of the Emergency Plan were reviewed:

Section 13.3.1.1 Definitions Section 13.3.3.1 Classification System Section 13.3.6.3.1 Onsite Assessment Facilities Table 13.3-1 Emergency Action Levels and Initiating Conditions The original Emergency Plan (1986) Sections 13.3.3.1, 13.3.6.3.1 and Table 13.3-1 were reviewed manually. No additional relevant or affected plan content was identified.

Part Ill. Describe How the Proposed Change Complies with Relevant Emergency Preparedness Regulation(s) and Previous Commitment(s) Made to the NRC:

10 CFA 50.47(b)(4)-A standard emergency'classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures Site Compliance - A standard emergency classification and action level scheme based on NEI 99-01 l;levision 5 remains in effect with this change. Proposed changes (1, 2, 3, 4 and 5) revise bases and attachments in Ef P-2-001 Classification of Emergencies for indication of seismic events impacting EALs HU6 and HA6 (natural or destructive phenomena) to the outputs provided by the new seismic monitoring instrumentation. The ability to monitor, detect and provide alarm indications of seismic events for emergency classification is maintained.

NRC Commitments - The Licensing Research System and NRG commitment sections of Emergency Implementing Procedure EIP-2-001 were reviewed for potential NAC commitment changes as a result of this procedure revision.

There were no identified conflicts with this procedure revision and the current listing of NRG commitments associated with the EIPs or Emergency Plan. All current NRC commitments that relate to emergency action levels continue to be maintained and fulfilled under this procedure revision.

EN-EP-305 ROOS

Attachment 3 Page 3 of 4 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: EIP-2*001 / Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change:

10 CFR 50.47(b)(4) - Emergency Classification System

• A standard scheme of emergency classification and action levels is in use.

Sections IV.B and IV.C of Appendix E to 10 CFR 50 provide supporting requirements. Informing criteria appear in Section 11.D of NUREG-0654 and the licensee's emergency plan.

Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:

The changes in EIP-2-001 result from an upgrade of the active portion of the Seismic Monitoring System from a_nalog components to digital. The analog Kinemetrics and Engdahl components are replaced with a digital Syscom Instruments Seismic Monitoring System;

1) The analog accelerometers had capability of recording a maximum of 1.0 g at full scale. The new digital accelerometers are capable of recording up to 4.0 g, an increase in overall range.

2) The analog accelerometers are sensitive to frequencies in the range of 0.1 to 50 Hz. The new digital accelerometers cover Oto 600Hz, an increase in frequency range which bounds the existing condition.

3) The seismic trigger of the analog system is set at 0.01 g. The seismic trigger function will remain as is with the new equipment set to the same setpoint of 0.01 g to start recording. The new recorders continuously record to a ring buffer with pre- and post-event time history recordings which is user selected from 1-100s. This data is available to be recorded to permanent memory.

4) The analog system is powered by internal batteries with trickle charge from 110 VAC capable of recording up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> upon loss of power. New components are supplied by normal AC power from an uninlerruplible power source. In addition, the new components are capable of having battery backed power in event of loss of normal power and will continue operating up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

5) The analog seismic monitoring system sensor packages contain three mutually orthogonal accelerometers. All four sensor packages are oriented to the same azimuths. The new digital system has a single sensor unit al each location; however this is a triaxial accelerometer capable of measuring in three directions and will be oriented to the same azimuths. *

6) The analog system records electrical signals from the accelerometers by magnetic tape with the acceleration signal and the time signal occupying separate tracks on the tape. With the digital system the magnetic tapes are removed and the new system wm provide real time output that shows clear delineation when Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) limits are exceeded. Additionally, the new system is capable of calculating Cumulative Absolute Velocity (CAV) or seismic intensity.

As shown above the new digital seismic monitoring system either maintains or exceeds the capabilities of the analog seismic monitoring system. The ability to detect seismic events to perform emergency classification actions is maintained.

This proposed change does not reduce the effectiveness of the emergency plan because:

1. While the new digital Seismic Monitoring System has updated indications the change does not modify the meaning of or intent of the EAL. The seismic trigger setpoint is maintained.

2. The change does not result in any additional emergency classifications or any less emergency classifications than those that should be made using the EAL.

The proposed changes to EIP-2-001, Classification of Emergencies continue to meet the planning standard outlined in 10 CFR 50.47(b)(4). These proposed changes do not represent a reduction in the effectiveness to the emergency plan and can be incorporated without prior NRC approval.

EN-EP-305 ROOS

Attachment 3 Page4 of 4 10CFR50.54(Q)(3) Evaluation Procedure/Document Number: EIP-2-001 I Revision: 28 Equipment/Facility/Other: River Bend Station

Title:

Classification of Emergencies Part VI. Evaluation Conclusion Answer the following questions about the proposed change.

I. Does the proposed change comply with 10CFR50.47(b) and 10CFR50 Appendix E? @YES0NO

2. Does the proposed change maintain the effectiveness of the emergency plan (i.e., no reduction in effectiveness)? li'.JYES0NO

3. Does the proposed change constitute an emergency action level scheme change? 0YES@'NO If questions 1 or 2 are answered NO, or question 3 answered YES, reject the proposed change, modify the proposed change and perform a new evaluation or obtain prior NRC approval under provisions of 1OCFA50.90. If questions 1 and 2 are answered YES, and question 3 answered NO, implement applicable change process(es). Refer to Section 6.7 Step 8.

Part VII. Signatures Preparer Name (Print) Preparer Signature Date:

Norman E Tison 1/20/20

{Optional) Reviewer Name (Print) Reviewer Signature Date:

James Lewis I...

Reviewer Name (Print) Dale:

Aaron Magee Nuclear EP Project Manager Approver Name (Print) Date:

T.W. Gates Emergency Planning Manager or designee EN-EP-305 ROOB

INFORMATION USE ATTACHMENT 1 PAGE10F2 (T yp1ca I)

--===-

---ENTERGY PROCEDURE NO.

PAR Procedure Action Request CURRENT REV. PROCEDURE TITLE EIP-2-001 27 Classification of Emerf!encies TYPE OF ACTION:

[&I PROCEDURE REVISION (PR)

Cl NEW PROCEDURE (NP) 0 CANCEL PROCEDURE (CX)

Part I. Description of Activity Being Reviewed (This is generally changes to the emergency plan. EALs, EAL bases, etc. - refer to step 3.0[6)):

Revise EIP-2-001, Classification of Emergencies as listed below:

I) Anachmcnl 4 Page 27, Am1chmcnt 9 User Aid I Page 143 nnd Anachmcnt 9 User Aid 2 Page 144:

From: Seismic event idcniificd by any 2 of 1he following:

Seismic C\'Cnt conlinned by activated seismic switch as indicated by receipt of EITHER a QB. b;

a. Annuncintor"Scismic Tope RocoruingSYS Sllllt" (P68()-0"-A-D<Xi)

b. Event lmlicatoron ERS-NBl-102 is white To: Seismic event identified by any 2 of the following:

• Seismic event confim1cd by activated seismic switch as intlica1etl by receipt of EITHER a OR b,

a. Annuncintor"SEISMIC SYS RECORDING /TROUBLE" (P680-02A-DCX>)

b. Even! Indicator on ERS-NBR3D TRIGGER (RECORD STARn is yellow

2) Annchmenl 4 Page 27, Anachmcnl 9 User Aid l Page 143 and Auachmenl 9 User Aid 2 Page 144:

From; a. Seismic c\*ent > Operating Basis Earthqu:ikc (OBE) as im.licatcd by:

Annuncia1or "Seismic Tape Recording System S1art" (P680-02A-D06J AND Event Indicator on ERS-NB1-I02 is white AND Receipt of EITHER I OR 2:

l. Annunciator "Seismic Event High" {P680,02A-CD6)

2. Annuncimor "Seismic Event High-High" (P680-02A-B06) AND amber ligh1(s) on panel NBI-IOJ To: a. Seismic evcm > Opi.-r.uing Basis Earthquake (QBE) as indicated by; Annuncia1or "SEISMIC SYS RECORDING/ TROUBLE' (P680-02A-D06)

AND ERS-NBR3D TRIGGER (RECORD STARn is yellow AND Rcceip1 of.fil.!!IBR I OR 2:

I. Annunciator "Seismic Event High" (P680,02A*CD6)

2. Annunciator "Seismic Event High-High (P68D-D2A-B06) ,:Yill ERS-NBR3A OBE (H[) yellow light

3) Auachmcnt 8 Page 86:

From: I. Seismic ewnt idcnlifictl by any 2 of the following:

Seismic event con finned by ac1iva1cd seismic switch as indicalcd by receipt of EITHER a OR b:

a. Annuncialor"Seismic Tope Roconling SYS Start" (P680-0'>..A,D06)

b. Even! Indicator on ERS,NBI-I02 is while To: I. Seismic even! itlcn1ificd by any 2 of !he following;

• Seismic event confinncd by activated seismic switch as indicated by receipt of EITHER a OR b; a Annwx:iator"SEL<;MIC SYS RF.CORDING I TROUBLE' (P68(}.()')-A-D06)

b. Even1 Indicator on ERS-NBR3D TRIGGER (RECORD STARn is yellow

INFORMATION USE ATTACHMENT 1 PAGE20F2 (T yp1ca l)

---

PAR ENTERGY Procedure Action Request PROCEDURE NO. CURRENT REV. PROCEDURE TITLE EIP-2-001 27 Classification of Emergencies TYPE OF ACTION:

[BJ PROCEDURE REVISION (PR)

D NEW PROCEDURE (NP) 0 CANCEL PROCEDURE (CX)

4) Attachment BPage 87:

From: The annunciators "Seismic Tape Rcconling SYS Start" nnd the "white" C\'Cnt indic.uorarc listed in the Alann Response Procwun:11.> \'ailication of an e.uthquake eVl:lll To: The annunciators "SEISMIC SYS RECORDING /TROUBLE' nn<l lhe ')\!llow" cwnt indicator arc listed in the Al:mn R ~ Procedure as \'ailication ofilll eanhjuakecvent

5) Att.Jchm:nt8 Pagl: 96; From: a. Seismic event> Operating Basis Eanhquake (OBE) as indicated by:.

Annunciator "Seismic Tape Rcconling System Stan" {P680-02A-D06)

Mm faent lndicntoron ERS-NBl-102 is white Mm Receipt or ElffiER I QR 2:

I. Annunciator "Seismic Event High (P680-02A-C06)

2. Annunciator "Seismic Event High-High" (P680-D2A-B06) ~ amber light(s) on panel NBl*l OJ Tu a. Seismic event > Operating Basis Earthquake (OBE) as indicated by. ,

Annunciator "SEISMIC SYS RECORDING/ TROUBLE (P680,02A*D06)

AND ERS-NBR3D TRIGGER (RECORD START) is yellow AND Receipt of EITIIER I QR 2:

I. Annunciator"Seismic Event High" (P680.02A-C06)

2. Annunciator "Seismic facnt High-High" (P680-02A*B06) AND ERS,NBR3A OBE (HI) yellow lighl

[BJ PAD COMPLETED (EN-Ll-100-ATI-9.1) [BJ 50.54Q REVIEW COMPLETED, (EN-EP-305)

[BJ LICENSING COMMITMENTS VERIFIED D CROSS DISCIPLINE REVIEW (if a licable)

REVIEW AND APPROVAL:

SIGNATURE/ KCN / DATE PREPARER . 7 / .tJ TECHNICAL REVIEWER EPMANAGER EFFECTIVE DA TE: 01/22/2020