ML18194A587

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Attachment 1 to 2.18.045, EP-AD-301, Emergency Action Level Technical Bases Document, Revision 9 and Associated 10 CFR50.54(q) Review
ML18194A587
Person / Time
Site: Pilgrim
Issue date: 07/11/2018
From:
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
Shared Package
ML18194A622 List:
References
2.18.045 EP-AD-601, Rev. 9
Download: ML18194A587 (356)
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Text

Attachment 1 Letter Number 2.18.045 EP-AD-601, Emergency Action Level Technical Bases Document, Revision 9 and associated 10 CFR 50.54(q) Review

a-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES

- REFERENCE USE Page 1 of 351 Emergency Action Level Technical Bases Document Emergency Action Level Technical Bases Document RTYPE H8.26 Change Statement

  • Correct wording from "well below the flood level of +13'6" MSL" to "well above the flood level of +13'6" MSL". Pages 78 and 96
  • Add "(+21.5')" to enhance conversion from Figure H-1 to chart use of "+21'6" ". Pages 78 and 96
  • Correct reference from "recorder 40-RR-1705-19" to "recorder 40-RR-1705-24". Page 21 .

-=--Entergy. ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 2 of 351 Emergency Action Level Technical Bases Document TABLE OF CONTENTS Section 1.0 PURPOSE ...................................................................................................... 4

2.0 REFERENCES

............................................................................................... 4 2.1 DEVELOPMENTAL ........................................................................................ 4 2.2 IMPLEMENTING ............................................................................................ 5 3.0 DEFINITIONS ................................................................................................ 5 4.0 RESPONSIBILITIES ...................................................................................... 8 5.0 DETAILS ....................................................................................-.................... 9 5.1 PRECAUTIONS AND LIMITATIONS ............................................................. 9 5.2 PROCEDURE ................................................................................................ 9 6.0 INTERFACES ................................................................................................ 9 7.0 RECORDS ............................... .-: .................................................................. 14 8.0 REQUIREMENTS AND COMMITMENTS .................................................... 14 9.0 ATTACHMENTS .......................................................................................... 15 ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT ...................... 17 ATTACHMENT 9.2 CATEGORY H, HAZARDS ..................................................................... 68 ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION ................................................ 138 ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION ................... 190 ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION ... 198 ATTACHMENT 9.6 FISSION PRODUCT BARRIER Loss/POTENTIAL Loss MATRIX AND BASES.'..................................................................................... 262 ATTACHMENT 9.7 Category E, ISFSI .......................................................................... 327 ATTACHMENT

9.8 BACKGROUND

AND DISCUSSION ........................................................ 329 9.8.1 Background .................................................................................... 329 9.8.2 Fission Product Barriers ................................................................. 329

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

~Entergy. ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 3 of 351 Emergency Action Level Technical Bases Document TABLE OF CONTENTS Section Title 9.8.3 Emergency Classification Based on Fission Product Barrier Degradation ........................................................................ 330 9.8.4 EAL Relationship to EOPs ............................................................. 331 9.8.5 Symptom-Based vs. Event-Based Approach .................................. 331 9.8.6 EAL Organization ........................................................................... 332 9.8.7 Technical Bases Information ......*.................................................... 334 9.8.8 Operating Mode Applicability ......................................................... 336 9.8.9 Validation of Indications, Reports, and Conditions ......................... 337 9.8.10 Planned vs. Unplanned Events ...................................................... 337 9.8.11 Classifying Transient Events .......................................................... 338

  • 9.8.12 9.8.13 ATTACHMENT 9.9 Imminent EAL Thresholds .............................................................. 338 Treatment of Multiple Events ......................................................... 338 ABBREVIATIONS/ACRONYMS .............................................................. 339 ATTACHMENT 9.10 PNPS-To-NEI 99-01 EAL CROSS-REFERENCE ................................ 345 ATTACHMENT 9.11 EAL Page Listing ............................................................................ 350 ATTACHMENT9.12 DOCUMENT CROSS-REFERENCES ...................................................... 351

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 4 of 351 Emergency Action Level Technical Bases Document 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Pilgrim Nuclear Power Station (PNPS). It should be used to facilitate review of the PNPS EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EP-IP-100.1, "Emergency Action Levels (EALs) ", may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The bases information may also be useful in training and for explaining event classifications to offsite officials.

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.

Attachments 9.9, Abbreviations/Acronyms, and 9.10, PNPS-To-NEI 99-01 EAL Cross-Reference, are to be used to clarify abbreviations and acronyms and as a cross-reference of PNPS EALs to the NEI 99-01 IC/EAL identification scheme.

2.0 REFERENCES

2.1 DEVELOPMENTAL In addition to the general references listed below, see the EAL basis discussions of Attachments 9.1 through 9. 7 for EAL-specific developmental references.

[1] EN-AD-101-01, "NMM Procedure Writer Manual"

[2] EP-PP-01, "PNPS Emergency Plan"

[3] NEI 99-01 Revision 5 Final, "Methodology for Development of Emergency Action Levels," February 2008

[4] NRG Regulatory Issue Summary (RIS) 2003-18, "Supplement 2, Use of Nuclear Energy Institute (NEI) 99-01, Methodology for Development of Emergency Action Levels,"

Revision 4, dated January 2003 (December 12, 2005)

[5] NRG Regulatory Issue Summary (RIS) 2007-01, "Clarification of NRC Guidance for Maintaining a Standard Emergency Action Level Scheme," dated January 10, 2007

[6] PNPS 1.3.4-10, "Writers' Guide for Emergency Operating Procedures"

a Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=:- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 5 of 351 Emergency Action Level Technical Bases Document 2.2 IMPLEMENTING

[1] EAL wall chart

[2] EP-IP-100, "Emergency Classification and Notification"

[3] EP-IP-100.1, "Emergency Action Levels (EALs)"

3.0 DEFINITIONS

[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 Specifications LCOs in effect.

(a) 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" .

  • [2]

(b) 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".

Bomb - Refers to an explosive device suspected of having sufficient force to damage plant systems or structures.

[3] Civil Disturbance - A group of people violently protesting station operations or activities at the site.

[4] Confinement Boundary - The barrier(s) between areas containing radioactive substances and the environment.

[5] Containment Closure - The action taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment closure is established when Primary or Secondary Containment integrity is established in accordance with Section 3.7 of the Technical Specifications .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

  • ~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 6 of 351 Emergency Action Level Technical Bases Document

[6] Explosion - A rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

[7] Extortion - An attempt to cause an action at the station by threat of force.

[8] 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.

[9] Hostage - Person(s) held as leverage against the station to ensure that demands will be met by the station.

[1 O] Hostile Action - An act toward PNPS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate plant personnel 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.

(a) Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on PNPS.

Nonterrorism-based EALs should be used to address such activities (e.g., violent acts between individuals in the Owner Controlled Area).

[11] 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.

[12] 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 time frames are specified, they shall apply.

[13] Intrusion - The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force.

[14] Normal Plant Operations - Activities at the plant site associated with routine testing, maintenance, or equipment operations in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from Normal Plant Operations.

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 7 of 351 Emergency Action Level Technical Bases Document

[15] Owner Controlled Area - For purpose of EAL classification, the "Owner Controlled Area" is the Entergy owned property on the north side of Rocky Hill Road.

[16] Projectile - An object directed toward an NPP that could cause concern for its continued operability, reliability, or personnel safety.

[17] Protected Area - An area which normally encompasses all controlled areas within the security protected area fence as delineated in PNPS 1 .3 .131 , "Owner Controlled Area (OCA) Access".

[18] Sabotage - Deliberate damage, misalignment, or misoperation 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.

[19] 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 .

[20] Significant Transient - An unplanned event involving any of the following:

(a) Runback > 25% thermal power (b) Electrical load rejection > 25% full electrical load (c) Reactor Scram (d) ECCS injection (e) Thermal power oscillations > 10%

[21] Strike Action - Work stoppage within the Protected Area by a body of workers to enforce compliance with demands made on PNPS. The strike action must threaten to interrupt Normal Plant Operations.

[22] Unisolable - A breach or leak that cannot be promptly isolated.

[23] Unplanned - A parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions .

NON-QUALITY

& PNPS RELATED EP-AD-601 Revision 9

~Entergy.

~*

EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 8 of 351 Emergency Action Level Technical Bases Document

[24] Valid - An indication, report, or condition is considered to be valid when it is verified by:

(a) an instrument channel check; or (b) indications on related or redundant indicators; or (c) 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.

[25] Visible Damage - Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

[26] Vital Areas - 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.

  • 4.0 RESPONSIBILITIES The Emergency Planning Manager shall be responsible for:

[1] Ensuring that each EAL listed in EP-IP-100.1, "Emergency Action Levels (EALs)" has a technically sound basis described within Attachments 9.1 through 9. 7 of this Procedure.

[2] Ensuring that, for any proposed revision to an EAL listed in EP-IP-100.1, "Emergency Action Levels (EALs) ":

(a) A technical basis has been developed to support the proposed revision.

(b) The technical basis is included as part of the revision review and approval processes.

(c) Changes are. performed and documented in accordance with the plant administrative procedures which include evaluating changes in accordance with 10CFR50.54(q) requirements *

[3] Ensuring that any approved changes to an EAL listed in EP-IP-100.1, "Emergency Action Levels (EALs)" are reflected on the controlled copies of the EAL wall charts and EAL Technical Bases Document.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 9 of 351 Emergency Action Level Technical Bases Document 5.0 DETAILS 5.1 PRECAUTIONS AND LIMITATIONS None 5.2 PROCEDURE

[1] The Emergency Planning Manager shall, upon review of a proposed revision to one or more EALs listed in EP-IP-100.1, "Emergency Action Levels (EALs)", verify that:

(a) A technical basis has been developed to support the proposed revision.

(b) The technical basis is included as part of the revision* review and approval processes.

(c) Changes are performed and documented in accordance with the plant administrative procedures which include evaluating changes in accordance with 10CFR50.54(q) requirements .

(d) If the proposed change affects the wording of the EAL wall charts, a revision to the wall charts is made prior to implementation of the revision.

[2] The Emergency Planning Manager shall, upon review of any proposed revision to the EALs listed in EP-IP-100.1, "Emergency Action Levels (EALs) ", ensure that changes are performed and documented in accordance with plant administrative procedures.

This includes evaluating changes in accordance with 10CFR50.54(q) requirements for the applicable source documents referenced and listed in this procedure and for any potential impact based on those EAL technical bases which reference them.

6.0 INTERFACES Bechtel Drawing Electrical single line diagram S-E-155 Plant Specific Technical Guidelines for EOPs and SAGs, Section PC/G EN-EP-313, "Offsite Dose Assessment using the Unified RASCAL Interface" EOP-01, "RPV Control" EOP-01, "RPV Control", Entry Condition EOP-02, "RPV Control, Failure-To-Scram" EOP-03, "Primary Containment Control" EOP-03, "Primary Containment Control", Entry Condition

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 10 of 351 Emergency Action Level Technical Bases Document EOP-04, "Secondary Containment Control" EOP-05, "Radioactive Release Control" EOP-18, "Steam Cooling" EOP-16, "RPV Flooding" EOP-17, "Emergency RPV Oepressurization" EOP-26, "RPV Flooding, Failure-to-Scram" EOP-27, "Emergency RPV Oepressurization, Failure-to-Scram" EP-AD-413, "Emergency Communications Test" FSAR Figures 5.2-1 through 5.2-6 FSAR Figure 8.6-1 FSAR Sections 4.4, 4.5, 4.6, 4.11 (MSL)

FSAR Section 4.7 (RCIC)

FSAR Section 4.9 (RWCU)

FSAR Section 4.10 - Nuclear System Leakage Rate Limits FSAR Section 5.2 - Primary Containment System FSAR Section 5.3 - Secondary Containment System FSAR Section 6.3 (HPCI)

FSAR Section 8.1 FSAR Section 8.3 FSAR Section 8.5 FSAR Section 9.2 - Liquid Radwaste System FSAR Section 10.15 FSAR Section 11.9 (FW)

FSAR Table 5.2-1 Modification ER03114340 NEI 99-01 Definitions and Appendix C NEI 99-01 Revision 5 NEI/NRC EAL FAQ #2006-014

=-* PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

~Entergy .

id.

EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 11 of 351 Emergency Action Level Technical Bases Document NEI White Paper, "Enhancements to Emergency Preparedness Programs for Hostile Action",

November 18, 2005 NRG Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security Based Events", July 18, 2005 PNPS 1.3.40, "Specifications For Vital Area Barrier Openings, Degradation, And Repair' PNPS 1.4.12, "Primary Containment Entry" PNPS 2.1.6, "Reactor Scram" PNPS 2.1.7, "Vessel Heatup and Coo/down" PNPS 2.2.120, "Postaccident Monitoring Panel" PNPS 2.2.126, "Anticipated Transient Without Scram (A TWS)"

PNPS 2.2.133, "H2!02 Analyzer and C19 Systems" PNPS 2.2.14, "12~V DC Battery Systems" PNPS 2.2.146, "Station Blackout Diesel Generator' PNPS 2.2.17, "Communications Systems"

PNPS 2.2.22, "Reactor Core Isolation Cooling System (RCIC)"

PNPS 2.2.32, "Salt Service Water System (SSW)"

PNPS 2.2.62, "Area Radiation Monitoring System" PNPS 2.2.64, "Source Range Monitoring System" PNPS 2.2.77, "Drywe/1 Leak Detection Systems" PNPS 2.2.79, "Reactor Protection System" PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"

PNPS 2.2.80, "Reactor Vessel Level, Temperature, And Internal Pressure Instrumentation" PNPS 2.2.83, "Reactor Cleanup System" PNPS 2.2.85, "Fuel Pool Cooling and Filtering System" PNPS 2.2.92, "Main Steam Line Isolation and Turbine Bypass Valves" PNPS 2.2.94, "Seawater System" PNPS 2.3.1, "General Action for Alarm Response and Annunciator Control" PNPS 2.4.143, "Shutdown from Outside Control Room" PNPS 2.4.144, "Degraded Voltage" PNPS 2.4.154, "Intake Structure Fouling"

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 12 of 351 Emergency Action Level Technical Bases Document PNPS 2.4.155, "Loss of Annunciator System" PNPS 2.4.25, "Loss of Shutdown Cooling" PNPS 2.4.31, "Reactor Basin and/or Spent Fuel Pool Drain-Down" PNPS 2.4.44, "Loss of Drywe/1 Area Coolers" PNPS 2.4.57, "Loss of Public-Address System" PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line" PNPS 2.4.A.5, "Loss of Electrical Bus AS" PNPS 2.4.A.6, "Loss of Electrical Bus A6" PNPS 2.5.2.71, "Radwaste Collection System" PNPS 3.M.2-5.1, "Source Range Monitor Calibration Instruction" PNPS 3.M.2-40, "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication" PNPS 5.2.1, "Earthquake" PNPS 5.3.11, "Loss of Essential DC Bus 016 or 04 and 036" PNPS 5.3.12, "Loss of Essential DC Bus 017 or 05 and 037" PNPS 5.3.31, "Station Blackout" PNPS 5.4.3, "Refueling Floor High Radiation" PNPS 5.5.1, "General Fire Procedure" PNPS 5.5.2, "Special Fire Procedure" PNPS 5.5.4, "Response to Hazardous Material Incidents" PNPS 6.3-064, "Routine Radiological Surveillance Program" PNPS 6.5-160, "Calibration of the Area Radiation Monitoring System" PNPS 8.A.13, "Plant Emergency Alarms and Radio Test" PNPS 8.E.29, "Salt Service Water System Instrumentation Calibration" PNPS 8.M.2-6.1, "Reactor Pressure Readout" PNPS ARP-C3RC-A7 PNPS ARP-C3RC-B7 PNPS ARP-C904LC-A4 PNPS ARP-C904LC-B4 PNPS ARP-C904LC-B7

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 13 of Revision 9 351 Emergency Action Level Technical Bases Document PNPS ARP-C904LC-C7 PNPS Emergency Plan Section B-3 PNPS NE-07-00006 Rev. 0 PNPS ODCM Section 7.1.1, Liquid Radioactive Waste Effluent Release PNPS ODCM Section 7.2.1, Liquid Radioactive Waste Effluent Monitoring System PNPS ODCM Section 7.2.2, Main Stack Gas Monitoring System PNPS ODCM Section 7.2.3, Reactor Building Exhaust Vent Monitor System PNPS ODCM Section 8.1, Liquid Effluent Monitor PNPS ODCM Section 8.3, Steam Jet Air Ejector Monitor PNPS ODCM Table 4.2-1, Radioactive Liquid Waste Sampling and Analysis Program PNPS ODCM Table 4.3-1, Radioactive Gaseous Waste Sampling and Analysis Program PNPS Plant-Specific and Severe Accident Management Guidelines PNPS Technical Specifications PNPS Technical Specifications, 1.0 Definitions

Revision 0, Prepared By: Scott McCain (05/5/09), Reviewed By: Ed Salomon (05/15/09) and Approved By: Kevin Wolf (05/15/09)

SAG-02, "Containment and Radioactivity Release Control" SUDDSRF96-35, Wind and Tornado Evaluation for PNPS

  • ~Entergy. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 14 of 351 Emergency Action Level Technical Bases Document 7.0 RECORDS This document does not generate any records.

8.0 REQUIREMENTS AND COMMITMENTS This section lists those external commitments (NRC commitments, QA audit findings, and INPO inspection items) implemented in this Procedure.

Reference Document Commitment Affected Section(s)/Step(s)

NRC Inspection Finding Develop and implement a Attachments 9.1 through 9.7.

81-15-34 system for use by the (See also EAL wall chart)

Control Room staff to aid in promptly classifying events.

NRC Inspection Finding Provide EALs which Attachments 9.1 through 9.7.

81-15-35 include specific and (See also EAL wall chart) observable Control Room instrument readings for each EAL corresponding to the respective initiating condition NRC Inspection Finding Provide EALs based on Attachment 9.1 84-05-04 field monitoring results (EALS AG1 .3 and AS1 .3, AG1 .2 and on the methods used and AG1 .3) if the effluent and containment monitors are inoperable or off-scale.

~Entergy EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 15 of 351 Emergency Action Level Technical Bases Document Reference Document Commitment Affected Section(s)/Step(s)

NRG Inspection Report Provide further Attachment 9.2 50-293/88-28 Item 2.4 clarification and ( EAL HA 1 .1 , HU 1.1 )

quantification for earthquake EALs.

NRG Bulletin 2005-02, Revise EGL criteria Attachment 9.2 (EALs HG4.1, "Emergency Preparedness definitions and HS4.1, HA4.1, HU4.1) and Response Actions for subcategory security Security Based Events" threat EALs to include expanded security based events.

9.0 ATTACHMENTS 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT 9.2 CATEGORY H, HAZARDS 9.3 CATEGORY S, SYSTEM MALFUNCTION 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL LOSS MATRIX AND BASES 9.7 CATEGORY E, ISFSI

9.8 BACKGROUND

AND DISCUSSION 9.8.1 Background 9.8.2 Fission Product Barriers 9.8.3 Emergency Classification Based on Fission Product Barrier Degradation 9.8.4 EAL Relationship to EOPs 9.8.5 Symptom-Based vs. Event-Based Approach 9.8.6 EAL Organization

  • 9.8.7 Technical Bases Information

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 16 of 351 Emergency Action Level Technical Bases Document 9.8.8 Operating Mode Applicability 9.8.9 Validation of Indications, Reports and Conditions 9.8.10 Planned vs. Unplanned Events 9.8.11 Classifying Transient Events 9.8.12 Imminent EAL Thresholds 9.8.13 Treatment of Multiple Events 9.9 ABBREVIATIONS/ACRONYMS 9.10 PNPS-TO-NEI 99-01 EAL CROSS-REFERENCE9.11 EAL PAGE LISTING 9.12 DOCUMENT CROSS-REFERENCES

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy: EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 17 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 1 of 51 Category A - Abnormal Rad Release/Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via nonradiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Offsite Rad Conditions Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements, or measured release rates via sampling indicate doses or dose rates above classifiable limits.
2. Onsite Rad Conditions & Spent Fuel Pool Events Sustained general area radiation levels in excess of those indicating loss of control of radioactive materials also warrant emergency classification.
3. MGR/GAS Radiation Sustained general area radiation levels in excess of those levels which may preclude access to vital plant areas also warrant emergency classification

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 18 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 2 of 51 Unusual Event - AU 1.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the Radiological Effluent Technical Specifications/ODCM for 60 minutes or longer EAL:

AU1.1 Unusual Event Any valid gaseous monitor reading> Table A-1 column "UE" for~ 60 min. (Note 2)

Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

I

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 19 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 3 of 51 Unusual Event - AU 1.1 TableA-1 Efflµent Monitor Clai;sification Thresholds .

GE SAE ALERT UE Release Point Monitor for;::,: 15 min. for;::,: 15 min. for;::,: 15 min. for;::,: 60 min.

RM-1705-18A/B (Panel C910 -

Stack Gas units of cps)

----- 4E+5 cps 2E+4 cps Low Range

(/)

=> Rl-1001-608 0 (Panel C170 -

w Stack Gas 20 R/hr 2 R/hr ----- -----

(/) units of R/hr)

<t

(.!) High Range RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C91 O - ----- 1E+5 cps 1E+3 cps units of cps)

C RM-1705-30 200 X hi-hi 5 Radwaste Discharge alarm*

0 Effluent (Panel C91 O - ----- ----- not to exceed 2 x hi-hi alarm*

J units of cps) 8E+5 cps
  • with Radwaste discharsie not isolated Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 20 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 4 of 51 Unusual Event - AU 1.1 Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

PNPS has found it advantageous to address gaseous and liquid releases* with separate EALs.

The ODCM multiples are specified in EALs AU1 .1/AU1 .2 and AA1 .1/AA1 .2 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared or a

  • release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

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

This EAL is intended for sites that have established effluent monitoring on nonroutine release pathways for which a discharge permit would not normally be prepared.

EALs AU1 .1 and AU1 .2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the setpoints.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

===== EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 21 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 5 of 51 Unusual Event - AU 1.1 PNPS Basis:

Gaseous releases in excess of two times the PNPS Offsite Dose Calculation Manual (ODCM) instantaneous limits that continue for greater than 60 minutes represent an uncontrolled situation and, hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.

The radiation monitors that detect gaseous radioactivity effluent release to the environment are the Main Stack process radiation monitors RM-1705-18A/B and the Reactor Building Ventilation Exhaust (RBVE) monitors RM-1705-32A/B (ref. 1 and 2).

  • The Main Stack process radiation monitor system initiates alarms whenever stack radioactivity release levels approach unacceptable limits. The Main Stack directs the effluents from one or more of the following systems to an elevated filtered release point:
  • Augmented Offgas System
  • Condenser Air Removal System Mechanical Vacuum Pump
  • Gland Seal Exhauster Indication of Main Stack effluent is provided by process radiation monitors RM-1705-1 BA/B on Panel C910 and recorder 40-RR-1705-19 on Panel C902 (ref. 1). The RBVE monitor alarms whenever effluent radioactivity release levels approach unacceptable limits. Air from areas containing potential sources of radioactive contamination such as the Reactor Building, Radwaste Building basement, and Turbine Building basement are discharged through the Reactor Building exhaust vent.

Indication of RBV effluent is provided by process radiation monitors RM-1705-32A/B on Panel C910 and recorder 40-RR-1705-24 on Panel C902 (ref. 2) .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 22 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 6 of 51 Unusual Event - AU1 .1 Complete assumptions and inputs for these EAL threshold values are documented from the calculation in Radiological Gaseous Effluent EAL Values (ref. 3). The calculated Main Stack and RBV values were rounded down to the most readable division marking on the limiting monitor for the EAL thresholds.

PNPS Basis Reference(s):

1. PNPS ODCM Section 7.2.2, Main Stack Gas Monitoring System
2. PNPS ODCM Section 7.2.3, Reactor Building Exhaust Vent Monitor System
3. Radiological Gaseous Effluent EAL Values (EALs AG 1.1, AS 1.1, AA 1.1 , and AU 1.1)

Revision 0, Prepared By: Scott McCain (05/5/09), Reviewed By: Ed Salomon (05/15/09) and Approved By: Kevin Wolf (05/15/09)

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 23 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 7 of 51 Unusual Event - AU1 .2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the Radiological Effluent Technical Specifications/ODCM for 60 minutes or longer EAL:

AU1.2 Unusual Event Valid radwaste effluent radiation monitor RM-1705-30 (Panel C910) reading > Table A-1

  • column "UE" for~ 60 min. (Note 2)

AND Radwaste discharge is not isolated Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown .

..,.,_ - PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

-===- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 24 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 8 of 51 Unusual Event - AU1 .2

  • , '<*Table A-1 Effluent IVlonitor Clas.sification Tl1re~ljol~s. ..

GE SAE ALERT UE Release Point Monitor for 2 15 min. for215min. for 2 15 min. for 2 60 min.

RM-1705-1 BA/B (Panel C910 -

Stack Gas units of cps)

----- 4E+5 cps 2E+4 cps Low Range Cl)

> Rl-1001-608 0 (Panel C170 -

w Stack Gas 20 R/hr 2 R/hr ----- -----

Cl) units of R/hr) c:i::

(!) High Range RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C910 - ----- ----- 1E+5 cps 1E+3 cps units of cps)

Cl RM-1705-30 200 X hi-hi Radwaste Discharge alarm*

0 Effluent (Panel C910 - ----- ----- 2 x hi-hi alarm*

units of cps) not to exceed

J 8E+5 cps
  • with Radwaste discharge not isolated Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time .

  • --:-.::=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 25 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 9 of 51 Unusual Event -AU1 .2 Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

PNPS has found it advantageous to address gaseous and liquid releases with separate EALs.

The ODCM multiples are specified in EALs AU1 .1/AU1 .2 and AA1 .1/AA1 .2 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate .

  • This EAL addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the IC established by the radioactivity discharge permit. This value may be associated with a planned batch release, or a continuous release path.

EAL AU1 .1 and EAL AU1 .2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the setpoints.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

PNPS Basis:

Liquid releases in excess of two times the Hi-Hi alarm that continue for greater than 60 minutes represent an uncontrolled situation and, hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes .

a* PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 26 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 10 of 51 Unusual Event- AU1 .2 Indication of radwaste effluent is provided by process radiation monitor RM-1705-30 on Panel C910 and recorder RR-1792 (red pen) on Panel C20 (Radwaste Control Room). The Radwaste Effluent Radiation Monitoring System monitors radwaste discharges to the discharge canal and provides alarm and automatic isolation functions if radioactivity levels exceed predetermined setpoints. The alarm and isolation setpoints are calculated for each discharge to ensure ODCM liquid effluent limits are not exceeded (ref. 1, 2).

At low classification levels, the concern for classification is the continuing, uncontrolled release of radioactivity and not the magnitude of the release. When the liquid release is isolated, the release is no longer continuing nor is it uncontrolled. Therefore, the classification is not appropriate when the liquid release is isolated. Radwaste effluent discharge isolation valves FV-7214A and B close if radwaste effluent radiation levels exceed the Hi-Hi alarm setpoint (ref. 3).

PNPS Basis Reference(s):

1.

2.

3.

PNPS ODCM Section 7.1.1, Liquid Radioactive Waste Effluent Release PNPS ODCM Section 8.1, Liquid Effluent Monitor PNPS ODCM Section 7.2.1, Liquid Radioactive Waste Effluent Monitoring System

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 27 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 11 of 51 Unusual Event - AU1 .3 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the Radiological Effluent Technical Specifications/ODCM for 60 minutes or longer EAL:

AU1.3 Unusual Event Confirmed sample analyses for gaseous or liquid releases indicate concentrations or

  • release rates > 2 x ODCM limits for 2:: 60 min. (Note 2)

Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable

. time.

This EAL addresses a potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 28 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 12 of 51 Unusual Event -AU1 .3 Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

PNPS has found it advantageous to address gaseous and liquid releases with separate EALs.

The ODCM multiples are specified in EALs AU1 .3 and AA1 .3 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared, or

  • a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

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.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

PNPS Basis Confirmed sample analyses in excess of two times the PNPS Offsite Dose Calculation Manual (ODCM) instantaneous limits that continue for greater than 60 minutes represent an uncontrolled situation and, hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the Unusual Event emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes. This EAL addresses collecting and analyzing both gaseous and liquid effluent samples to ensure that release conditions above nominal steady state conditions are detected and reported (ref. 1, 2).

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 29 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 13 of 51 Unusual Event - AU 1.3 PNPS Basis Reference(s):

1. PNPS ODCM Table 4.2-1, Radioactive Liquid Waste Sampling and Analysis Program
2. PNPS ODCM Table 4.3-1, Radioactive Gaseous Waste Sampling and Analysis Program
  • Entergy . PNPS NON-QUALITY

= ==- EMERGENCY PLAN RELATED EP-AD-601 Revision 9

-===- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 30 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 14 of 51 Alert - AA 1.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the Radiological Effluent Technical Specifications/ODCM for 15 minutes or longer EAL:

AA1.1 Alert Any valid gaseous monitor reading> Table A-1 column "Alert for~ 15 min." (Note 2)

Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 31 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 15 of 51 Alert - AA 1.1 GE SAE ALERT UE Release Point Monitor for~ 15 min. for~ 15 min. for~ 15 min. for~ 60 min.

RM-1705-1 BA/B Stack Gas (Panel C910 - 4E+5 cps 2E+4 cps units of cps)

Low Range u,t---------t-------+------------------------1

, Rl-1001-608

@ Stack Gas (Panel C170 - 20 R/hr 2 R/hr

~ units of R/hr)

CJ High Range RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C910 - 1E+5 cps 1E+3 cps

  • Radwaste Discharge Effluent units of cps)

RM-1705-30 (Panel C910 -

units of cps) 200 X hi-hi alarm*

not to exceed 8E+5 cps 2 x hi-hi alarm*

  • with Radwaste discharge not isolated Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 32

. Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 16 of 51

.'Alert - AA 1.1 PNPS has found it advantageous to address gaseous and liquid releases with separate EALs.

The ODCM multiples are specified in EALs AU1 .1/AU1 .2 and AA1 .1/AA1 .2 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

  • This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

This EAL is intended for sites that have established effluent monitoring on nonroutine release pathways for which a discharge permit would not normally be prepared.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

PNPS Basis:

This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100 for the RBV. A multiple of 200 times the ODCM limit is not used for the Main Stack EAL as it would result in the Alert value being higher than the Site Area Emergency value. Therefore, to provide a numeric progression for the EAL, the AA 1.1 value is set logarithmically half way between the AS1 .1 value and the AU1 .1 value. The radiation monitors that detect gaseous radioactivity effluent release to the environment are the Main Stack process radiation monitors RM-1705-1 BA/Band the Reactor Building Ventilation Exhaust (RBVE) monitors RM-1705-32A/B on Panel C910 (ref .1, 2).

Complete assumptions and inputs for these EAL threshold values are documented from the calculation in Radiological Gaseous Effluent EAL Values (ref. 3). The calculated Main Stack and RBV values were rounded down to the most readable division marking on the limiting monitor for the EAL thresholds.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 33 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 17 of 51 Alert - AA 1.1 PNPS Basis Reference(s):

1. PNPS ODCM Section 7.2.2, Main Stack Gas Monitoring System
2. PNPS ODCM Section 7.2.3, Reactor Building Exhaust Vent Monitor System
3. Radiological Gaseous Effluent EAL Values (EALs AG1 .1, AS1 .1, AA 1.1 and AU1 .1)

Revision 0, Prepared By: Scott McCain (05/5/09), Reviewed By: Ed Salomon (05/15/09) and Approved By: Kevin Wolf (05/15/09)

-::::::-Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 34 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 18 of 51 Alert - AA 1.2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the Radiological Effluent Technical Specifications/ODCM for 15 minutes or longer EAL:

AA1.2 Alert Valid radwaste effluent radiation monitor RM-1705-30 (Panel C910) reading> Table A-1 column "Alert for~ 15 min." (Note 2)

Radwaste discharge is not isolated Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

NON-QUALITY A

PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 35 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 19 of 51 Alert - AA1 .2 GE SAE ALERT UE Release Point Monitor for;:: 15 min. for;:: 15 min. for;:: 15 min. for;:: 60 min.

RM-1705-1 BA/B (Panel C910 - 4E+5 cps 2E+4 cps Stack Gas units of cps)

Low Range u,t---------t-------+-------1.-------+------+---------t

> Rl-1001-608

@ Stack Gas (Panel C170 - 20 R/hr 2 R/hr

~ units of R/hr)

G H~hRa~e 1---------1-------1-----~1-------1------+---------1 RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C910 - 1E+5 cps 1E+3 cps units of cps)

C RM-1705-30 200 X hi-hi 5 Radwaste Discharge alarm*

(Panel C910 - 2 x hi-hi alarm*

0 Effluent not to exceed

J units of cps) 8E+5 cps
  • with Radwaste discharge not isolated Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceed.s regulatory commitments for an extended period of time .

NON-QUALITY A

PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 36 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 20 of 51 Alert -AA1 .2 Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

PNPS has found it advantageous to address gaseous and liquid releases with separate EALs.

The ODCM multiples are specified in EALs AU1 .1/AU1 .2 and AA1 .1/AA1 .2 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared, or

  • a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

This EAL addresses radioactivity releases that, for whatever reason, cause effluent radiation monitor readings to exceed the threshold identified in the IC established by the radioactivity discharge permit. This value may be associated with a planned batch release or a continuous release path.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

PNPS Basis:

This event escalates from the Unusual Event by escalating the magnitude of the release by a factor of 100.

The threshold of> 200 times the Hi-Hi alarm setpoint is limited to a maximum value of 8E+5 cps to assure an on-scale readable value.

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 37 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 21 of 51 Alert - AA 1.2 PNPS Basis Reference(s):

1. PNPS ODCM Section 7.1.1, Liquid Radioactive Waste Effluent Release
2. PNPS ODCM Section 8.1, Liquid Effluent Monitor
3. PNPS ODCM Section 7.2.1, Liquid Radioactive Waste Effluent Monitoring System

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-==-- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 38 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 22 of 51 Alert - AA1 .3 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 200 times the Radiological Effluent Technical Specifications/ODCM for 15 minutes or longer EAL:

AA1.3 Alert Confirmed sample analyses for gaseous or liquid releases indicate concentrations or release rates > 200 x ODCM limits for 2 15 min. (Note 2)

Note 2: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the release duration has exceeded, or will likely exceed, the applicable time. In the absence of data to the contrary, assume that the release duration has exceeded the applicable time if an ongoing release is detected and the release start time is unknown.

Mode Applicability:

All NEI 99-01 Basis:

The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

This EAL addresses an actual or substantial potential decrease in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time.

a~Entergy- NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 39 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 23 of 51 Alert- AA1 .3 Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of a degradation in these features and/or controls.

The ODCM multiples are specified in EALs AU1 .3 and AA1 .3 only to distinguish between nonemergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, not the magnitude of the associated dose or dose rate.

This EAL includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow,

  • alarm setpoints, etc.) on the applicable permit.

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.

The underlying basis of this EAL involves the degradation in the level of safety of the plant implied by the uncontrolled release.

PNPS Basis:

Confirmed sample analyses in excess of 200 times the PNPS Offsite Dose Calculation Manual (ODCM) limits that continue for greater than 15 minutes represent an uncontrolled situation and, hence, a potential degradation in the level of safety. This event escalates from the Unusual Event by raising the magnitude of the release by a factor of 100 over the Unusual Event level (i.e., 200 times ODCM).

The required release duration was reduced to 15 minutes in recognition of the raised severity.

PNPS Basis Reference(s):

1. PNPS ODCM Table 4.2-1 Radioactive Liquid Waste Sampling and Analysis Program
  • 2. PNPS ODCM Table 4.3-1 Radioactive Gaseous Waste Sampling and Analysis Program

-===- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 40 of Revision 9 351

  • Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 24 of 51 Site Area Emergency - AS 1.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mrem TEDE or 500 mrem thyroid COE for the actual or projected duration of the release EAL:

AS1.1 Site Area Emergency Valid Main Stack High Range Effluent Monitor (Rl-1001-608) reading> Table A-1 column "SAE for 2 15 min." (Note 1)

Note 1: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. (See EAL AS1 .2.) Do not delay declaration awaiting dose assessment results.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 41 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 25 of 51 Site Area Emergency - AS 1.1

- * , " * * - ' , , -',: ,.. '* ,_ > ' * < ~ ~ /-' , ..,

Tab!~;A-1 * .Effluent Mo~jt9r Glas~mcatjon.'fhf:~Ji_ol<is '.:

GE SAE ALERT UE Release Point Monitor for~ 15 min. for~ 15 min. for~ 15 min. for~ 60 min.

RM-1705-1 BA/B Stack Gas (Panel C910 - 4E+5 cps 2E+4 cps units of cps)

Low Range U)

, Rl-1001-608 0 (Panel C170 -

w Stack Gas 20 R/hr 2 R/hr

~

C) units of R/hr)

High Range RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C910 - 1E+5 cps 1E+3 cps units of cps)

C RM-1705-30 200 X hi-hi

, Radwaste Discharge alarm* 2 x hi-hi alarm*

(Panel C910 -

0 Effluent not to exceed

i units of cps) 8E+5 cps
  • with Radwaste discharge not isolated Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The site specific monitor list in Table A-1 includes effluent monitors on all potential release pathways .

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 42 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 26 of 51 Site Area Emergency - AS1 .1 PNPS Basis:

The radiation monitors that detect gaseous radioactivity effluent release to the environment include both the Main Stack process radiation monitors and the Reactor Building Ventilation Exhaust (RBVE) process radiation monitors. In an accident condition or when high radiation condition exists in the Reactor Building, the Reactor Building Isolation System (RBIS) automatically trips and isolates the normal Reactor Building ventilation exhaust system and also starts the Standby Gas Treatment System which then exhausts air from the Reactor Building directly to the Main Stack. The Main Stack release point is therefore the only gaseous release pathway suitable for assessing this EAL (ref. 1).

The Main Stack process radiation monitor system initiates alarms whenever stack radioactivity release levels approach unacceptable limits. The Main Stack directs the effluents from one or more of the following systems to an elevated filtered release point:

Standby Gas Treatment System Augmented Offgas System

  • Condenser Air Removal System Mechanical Vacuum Pump
  • Gland Seal Exhauster Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If dose assessment results are available, declaration should be based on dose assessment instead of the radiation monitor EAL threshold values.

(See EAL AS1 .2). However, the emergency declaration should not be delayed awaiting dose assessment results.

Complete assumptions and inputs for these EAL threshold values are documented from the calculation in Radiological Gaseous Effluent EAL Values (ref. 2). The calculated Main Stack value was rounded down to the most readable division marking on the limiting monitor for the EAL threshold.

e-=:- Entergy . PNPS NON-QUALITY ,

RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 43 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 27 of 51 Site Area Emergency - AS1 .1 PNPS Basis Reference(s):

1. PNPS FSAR Section 5.3, Secondary Containment System
2. Radiological Gaseous Effluent EAL Values (EALs AG1 .1, AS1 .1, AA 1.1 and AU1 .1)

Revision 0, Prepared By: Scott McCain (05/5/09), Reviewed By: Ed Salomon (05/15/09) and Approved By: Kevin Wolf (05/15/09)

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

- Entergy. ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 44 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 28 of 51 Site Area Emergency - AS 1.2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mrem TEDE or 500 mrem thyroid CDE for the actual or projected duration of the release EAL:

AS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or

> 500 mrem thyroid CDE at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures 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 EAL.

NON-QUAL ITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 45 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT

======

Sheet 29 of 51 I Site Area Emergency - AS 1.2 I PNPS Basis:

The Site Area values are based on doses at or beyond the site boundary that result from an actual or imminent release of gaseous radioactivity that exceeds 100 mrem TEDE or 500 mrem COE thyroid for the actual or projected duration of the release.

When dose assessment results (using actual meteorology and release information) become available, they will be used to determine whether the integrated dose exceeds the threshold values of> 100 mrem TEDE or> 500 mrem thyroid COE at or beyond the site boundary.

Actual meteorology is specifically identified since it gives the most accurate dose assessment.

If the dose assessment results are available at the time that the classification is made, the results will be used in conjunction with this EAL for classifying the event rather than effluent radiation monitor EAL threshold values.

  • PNPS Basis Reference(s):
1. EN-EP-313, "Offsite Dose Assessment using the Unified RASCAL Interface"

~--------------------

~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 46 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 30 of 51 Site Area Emergency -AS1 .3 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mrem TEDE or 500 mrem thyroid COE for the actual or projected duration of the release EAL:

AS1.3 Site Area Emergency Field survey indicates closed window dose rate > 100 mR/hr expected to continue for 2 1 hr at or beyond the site boundary OR Field survey sample analysis indicates thyroid COE > 500 mrem for 1 hr of inhalation at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated 'with the failure of plant systems needed for the protection of the public.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 47 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 31 of 51 Site Area Emergency - AS 1.3 PNPS Basis:

Although this EAL initiating condition references the TEDE or thyroid COE as an integrated dose, field survey results are not generally reported in these dose quantities, but rather in terms of a dose rate or an air sample concentration, respectively. For this reason, the field survey EALs are based on a closed window dose rate greater than 100 mR/hr expected to continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer; or analyses of field survey samples indicate thyroid COE greater than 500 mrem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation at or beyond the site boundary and whichever is more limiting.

PNPS Basis Reference(s):

1. EP-IP-310, "Radiation Monitoring Team Activation And Response"

.. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 48 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 32 of 51 General Emergency - AG1 .1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE for the actual or projected duration of the release using actual meteorology EAL:

AG1.1 General Emergency Valid Main Stack High Range Effluent Monitor (Rl-1001-608) reading> Table A-1 column "GE for~ 15 min." (Note 1)

Note 1: The Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition will likely exceed the applicable time. If dose assessment results are available, declaration should be based on dose assessment instead of radiation monitor values. (See EAL AG1 .2.) Do not delay declaration awaiting dose assessment results.

  • ~Entergy. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 49 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 33 of 51 General Emergency - AG1 .1 TableA-1 Effluent Monitor Classification Thresholds GE SAE ALERT UE Release Point Monitor for;,: 15 min. for;,: 15 min. for;,: 15 min. for;,: 60 min.

RM-1705-1 BA/B (Panel C910 - 4E+5 cps 2E+4 cps Stack Gas units of cps)

Low Range ti)

, Rl-1001-608 0 (Panel C170 -

w Stack Gas 20 R/hr 2 R/hr ----- -----

ti)

<( units of R/hr)

(!) High Range RM-1705-32A/B Rx Bldg Vent Exhaust (Panel C910 - ----- ----- 1E+5 cps 1E+3 cps units of cps)

C RM-1705-30 200 X hi-hi 5 Radwaste Discharge alarm*

0 Effluent (Panel C910 - ----- ----- 2 x hi-hi alarm*

units of cps) not to exceed

i 8E+5 cps
  • with Radwaste discharQe not isolated Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.

The site specific monitor list in Table A-1 includes effluent monitors on all potential release pathways .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 50 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 34 of 51 General Emergency - AG1 .1 PNPS Basis:

The General Emergency effluent monitor readings are one decade greater than the Site Area Emergency value (refer to EAL AS1 .1 ).

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If dose assessment results are available, declaration should be based on dose assessment instead of the radiation monitor EAL threshold values.

(See EAL AG1 .2.) However, the emergency declaration should not be delayed awaiting dose assessment results.

Complete assumptions and inputs for these EAL threshold values are documented from the

  • calculation in Radiological Gaseous Effluent EAL Values (ref. 2). The calculated Main Stack value was rounded down to the most readable division marking on the limiting monitor for the EAL threshold.

PNPS Basis Reference(s):

1. PNPS FSAR Section 5.3, Secondary Containment System
2. Radiological Gaseous Effluent EAL Values (EALs AG1 .1, AS1 .1, AA 1.1 and AU1 .1)

Revision 0, Prepared By: Scott McCain (05/5/09), Reviewed By: Ed Salomon (05/15/09) and Approved By: Kevin Wolf (05/15/09)

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 51 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 35 of 51 General Emergency -AG1 .2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE for the actual or projected duration of the release using actual meteorology EAL:

AG1.2 General Emergency

  • Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or

> 5,000 mrem thyroid CDE at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage.

Since dose assessment is based on actual meteorology, whereas the monitor reading EAL is not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures 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 EAL.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 52 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 36 of 51 General Emergency- AG1 .2 PNPS Basis:

The General Emergency values are based on doses at or beyond the site boundary that result from an actual or imminent release of gaseous radioactivity that exceeds 1000 mrem TEDE or 5000 mrem COE thyroid for the actual or projected duration of the release.

When dose assessment results (using actual meteorology and release information) become available, they will be used to determine whether the integrated dose exceeds the threshold values of> 1000 mrem TEDE or> 5000 mrem thyroid COE at or beyond the site boundary.

Actual meteorology is specifically identified since it gives the most accurate dose assessment.

If the dose assessment results are available at the time that the classification is made, the results will be used in conjunction with this EAL for classifying the event rather than effluent radiation monitor EAL threshold values.

PNPS Basis Reference(s):

1. EN-EP-313, "Offsite Dose Assessment using the Unified RASCAL Interface"
  • e-===* Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 53 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 37 of 51 General Emergency - AG 1.3 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 1 - Offsite Rad Conditions Initiating Condition: Offsite dose resulting from an actual or imminent release of gaseous radioactivity greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE for the actual or projected duration of the release using actual meteorology EAL:

AG1.3 General Emergency

  • Field survey results indicate closed window dose rates> 1,000 mR/hr expected to continue for ~ 1 hr at or beyond the site boundary OR Analyses of field survey samples indicate thyroid COE> 5,000 mrem for 1 hr of inhalation at or beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 54 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 38 of 51 General Emergency- AG1 .3 PNPS Basis:

Although this EAL initiating condition references the TEDE or thyroid COE as an integrated dose, field survey results are not generally reported in these dose quantities, but rather in terms of a dose rate or an air sample concentration, respectively. For this reason, the field survey EALs are based on a closed window dose rate greater than 1000 mR/hr expected to continue for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or longer; or analyses of field survey samples indicate thyroid COE greater than 5000 mrem for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of inhalation at or beyond the site boundary and whichever is more limiting.

PNPS Basis Reference(s):

1. EP-IP-310, "Radiation Monitoring Team Activation And Response"
  • --==e~Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 55 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 39 of 51 Unusual Event - AU2.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - Onsite Rad Conditions & Spent Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation levels EAL:

AU2.1 Unusual Event Unplanned low water level or alarm indicating uncontrolled water level decrease in the reactor cavity or spent fuel pool with all irradiated fuel assemblies remaining covered by water

  • AND Valid area radiation monitor reading rise on ANY of the following:
  • New Fuel Vault (RIS-1815-30)
  • Refuel Floor Shield Plug Area (RIS-1815-3E)
  • Spent Fuel Pool Area (RIS-1815-3F)

Mode Applicability:

All

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 56 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 40 of 51 Unusual Event - AU2.1 NEI 99-01 Basis:

This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel or events that have resulted in, or may result in, unplanned increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.

The refueling pathway is the reactor cavity connected to the spent fuel pool through the spent fuel pool gate. While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

For refueling events where the water level drops below the RPV flange, classification would be via EAL CU2.1, CU2.2, or CU2.3. This event escalates to an Alert in accordance with EAL AA2.1 if irradiated fuel outside the reactor vessel is uncovered. For events involving

  • irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier Table for events in operating modes 1-3.

PNPS Basis:

Loss of inventory from the reactor cavity or spent fuel pool may reduce water shielding above spent fuel and cause unexpected increases in plant radiation levels. Classification as an Unusual Event is warranted as a precursor to a more serious event.

Spent fuel pool level instruments Ll-4816A and Ll-4816B, located on the back wall of the Control Room, and Spent fuel pool temperature instruments TE-4831 and TS-4807, provide a wide range indication of level and temperature in the spent fuel pool. In addition, one or more of the following annunciators may be indicative of an uncontrolled water level decrease in the reactor cavity, spent fuel pool, or fuel transfer canal (cattle chute):

  • "FUEL POOL COOLING PANEL ALARM" (C2R-D7)
  • "FUEL POOL LOW LEVEL" (C39-F1)
  • "REACTOR BASIN LOW LEVEL" (C39-E3)
  • "DRYWELL BELLOWS SEAL LEAKAGE" (C39-A1)
  • --=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 57 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 41 of 51 Unusual Event - AU2.1

  • "REFUEL BELLOWS SEAL RUPTURE" (C39-A2)
  • "FUEL POOL GATE LEAKAGE HIGH" (C39-A3)
  • "SKIMMER SURGE TANK LOW LEVEL" (C39-B2)
  • "FUEL POOL TEMP HIGH" (C39-C1)

Spent Fuel Pool Skimmer Surge Tank level is indicated on Ll-4815A (ref.1 ).

Increases in area radiation levels due to loss of water shielding above irradiated fuel outside the RPV may be detected by the following (ref. 2):

  • Refuel Floor Area Radiation Monitor(s) in alarm .
  • Refuel Floor Ventilation Process Radiation Monitor alarm .
  • Portable radiation monitors A minimum depth of water of about 10 feet over spent fuel assemblies and structures is ensured by providing no drain in the pool. In accordance with Technical Specifications Section 4.3.2, Drainage, the Spent Fuel Pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 115 ft (Tech Spec Amendment 177). (ref. 3)

This event escalates to an Alert if irradiated fuel outside the RPV is uncovered.

PNPS Basis Reference(s):

1. PNPS 2.4.31, "Reactor Basin and/or Spent Fuel Pool Draindown"
2. PNPS 5.4.3, "Refueling Floor High Radiation"
3. PNPS 2.2.85, "Fuel Pool Cooling and Filtering System"
4. EC 45088, Fukushima - Spent Fuel Pool Level Instrumentation
  • ~Entergy. NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 58 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 42 of 51 Unusual Event - AU2.2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - Onsite Rad Conditions & Spent Fuel Pool Events Initiating Condition: Unplanned rise in plant radiation levels EAL:

AU2.2 Unusual Event Unplanned valid area radiation monitor reading or survey results rise by a factor of 1000 over normal levels*

  • Normal levels can be considered as the highest reading in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> excluding the current peak value Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increased radiation levels as a result of water level decreases above irradiated fuel or events that have resulted in, or may result in, unplanned increases in radiation dose rates within plant buildings. These radiation increases represent a loss of control over radioactive material and represent a potential degradation in the level of safety of the plant.

EAL#2 This EAL addresses increases in plant radiation levels that represent a loss of control of radioactive material resulting in a potential degradation in the level of safety of the plant.

This EAL excludes radiation level increases that result from planned activities such as use of radiographic sources and movement of radioactive waste materials. A specific list of ARMs is not required as it would restrict the applicability of the threshold. The intent is to identify loss of control of radioactive material in any monitored area.

    • -Entergy . PNPS NON-QUALITY i: =~ EMERGENCY PLAN RELATED EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 59 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 43 of 51 Unusual Event - AU2.2 PNPS Basis:

The ARMs monitor the gamma radiation levels in units of mR/hr at selected areas throughout the station. If radiation levels exceed a preset limit in any channel, the Main Control Room annunciator and local alarms will be energized to warn of abnormal or significantly changing radiological conditions (ref. 1, 2).

Routine and work-specific surveys are conducted throughout the station at frequencies specified by Radiation Protection management. Work-specific surveys are conducted in accordance with the Radiological Work Permit (RWP). (ref. 3)

PNPS Basis Reference(s):

1. PNPS 2.2.62, "Area Radiation Monitoring System"
2. PNPS 6.5-160, "Calibration of the Area Radiation Monitoring System"
3. PNPS 6.3-064, "Routine Radiological Surveillance Program"

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-- Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 60 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 44 of 51 Alert - AA2.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - Onsite Rad Conditions & Spent Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the RPV EAL:

AA2.1 Alert Damage to irradiated fuel OR loss of water level (uncovering irradiated fuel outside the RPV) that causes a valid high alarm on any of the following radiation monitors (Panel C91 O/C911 ):

New Fuel Vault (RIS-1815-3D)

Refuel Floor Shield Plug Area (RIS-1815-3E)

Spent Fuel Pool Area (RIS-1815-3F)

Refuel Floor Vent Exhaust (RIS-1705-BA-D)

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increases in radiation dose rates within plant buildings and may be a.

precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant.

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

NON-QUALITY

.... PNPS

,:,..,:;::$ EMERGENCY PLAN

- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 61 of 351 Emergency Action l:.evel Technical Bases Document ATTACHMENT 9.1 CATEGORY A; ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 45 of 51 Alert - AA2.1 Increased ventilation monitor readings may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Increased background at the ventilation monitor due to water level decrease may mask increased ventilation exhaust airborne activity and needs to be considered.

While a radiation monitor could detect an increase in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

Escalation of this emergency classification level, if appropriate, would be based on EAL AS1 .1, AS1 .2, AS1 .3, AG1 .1, AG1 .2, or AG1 .3.

PNPS Basis:

When considering classification, information may come from:

  • Radiation monitor readings
  • Sampling and surveys
  • Dose projections/calculations
  • Reports from the scene regarding the extent of damage (e.g., refueling crew, Radiation Protection technicians)

This EAL is defined by the specific areas where irradiated fuel is located, such as the reactor cavity or spent fuel pool (SFP).

Spent fuel pool level instruments Ll-4816A and Ll-48168, located on the back wall of the Control Room, and Spent fuel pool temperature instruments TE-4831 and TS-4807, provide a wide range indication of level and temperature in the spent fuel pool. In addition, one or more of the following annunciators may be indicative of an uncontrolled water level decrease in the reactor cavity or spent fuel pool:

  • "FUEL POOL COOLING PANEL ALARM" (C2R-D7)
  • "FUEL POOL LOW LEVEL" (C39-F1)
  • "REACTOR BASIN LOW LEVEL" (C39-E3)

a-===- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE 0

PROCEDURES REFERENCE USE Page 62 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 46 of 51 Alert - AA2.1

  • "DRYWELL BELLOWS SEAL LEAKAGE" C39-A 1)
  • "REFUEL BELLOWS SEAL RUPTURE" (C39-A2)
  • "FUEL POOL GATE LEAKAGE HIGH" (C39-A3)
  • "SKIMMER SURGE TANK LOW LEVEL" (C39-B2)
  • "FUEL POOL TEMP HIGH" (C39-C1)

Spent Fuel Pool Skimmer Surge Tank level is indicated on Ll-4815A (ref. 1).

The high alarms of the listed ARMs (RIS-1815-3D, RIS-1815-3E, and RIS-1815-3F) activate Control Room annunciator C7 ("REFUEL FLOOR RAD HI") on Panel C904LC. The high alarm setpoint is at approximately 40 mR/hr. The ARM indication is displayed on recorder

  • 40-RR-1815-6 (Panel C902) (ref. 2, 3).

The high alarm of the Refuel Floor Exhaust PRMs activates Control Room annunciators A4

("REFL FLR VENT RAD CHAN A HI") and 84 ("REFL FLR VENT RAD CHAN B HI") on Panel C904LC. The high alarm is set at 16 mR/hr (increasing) during normal operation or

67 mR/hr (increasing) during refueling. The PRM indication is displayed on recorder RR-1705-21 on Panel C902. If both channels tripped, the following occurs (ref. 4, 5)
  • Standby Gas Treatment System starts Evacuation of refuel floor personnel to RB 91' changeout area is required if the high alarm is received on any of these monitors (ref. 6).

PNPS Basis Reference(s):

1. PNPS 2.4.31, "Reactor Basin and/or Spent Fuel Pool Draindown"
2. PNPS 6.5-160, "Calibration of the Area Radiation Monitoring System"
3. PNPS ARP-C904LC-C7
4. PNPS ARP-C904LC-A4
5. PNPS ARP-C904LC-84 6.

7.

PNPS 5.4.3, "Refueling Floor High Radiation" EC 45088, Fukushima - Spent Fuel Pool Level lnstrumention I

l_ -

~Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 63 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 47 of 51 Alert -AA2.2 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 2 - Onsite Rad Conditions & Spent Fuel Pool Events Initiating Condition: Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the RPV EAL:

AA2.2 Alert A water level drop in the reactor cavity or spent fuel pool that will result in irradiated fuel becoming uncovered

  • Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increases in radiation dose rates within plant buildings and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent an actual or substantial potential degradation in the level of safety of the plant.

Escalation of this emergency classification level, if appropriate, would be based on EAL AS1 .1, AS1 .2, AS1 .3, AG1 .1, AG1 .2, or AG1 .3 .

NON-QUALITY PNPS EP-AD-601 RELATED Revision 9

-=-- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 64 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 48 of 51 Alert - AA2.2 PNPS Basis:

When considering classification, information may come from:

  • Radiation monitor readings
  • Sampling and surveys
  • Dose projections/calculations
  • Reports from the scene regarding the extent of damage (e.g., refueling crew, radiation protection technicians)

This EAL is defined by the specific areas where irradiated fuel is located, such as the reactor

  • cavity or spent fuel pool (SFP).

Spent fuel pool level instruments Ll-4816A and Ll-4816B, located on the back wall of the Control Room, and Spent fuel pool temperature instruments TE-4831 and TS-4807, provide a wide range indication of level and temperature in the spent fuel pool. In addition, one or more of the following annunciators may be indicative of an uncontrolled water level decrease in the reactor cavity or spent fuel pool:

  • "FUEL POOL COOLING PANEL ALARM" (C2R-D7)
  • "FUEL POOL LOW LEVEL" (C39-F1)
  • "REACTOR BASIN LOW LEVEL" (C39-E3)
  • "DRYWELL BELLOWS SEAL LEAKAGE" (C39-A 1)
  • "REFUEL BELLOWS SEAL RUPTURE" (C39-A2)
  • "FUEL POOL GATE LEAKAGE HIGH" (C39-A3)
  • "SKIMMER SURGE TANK LOW LEVEL" (C39-B2)
  • "FUEL POOL TEMP HIGH" (C39-C1)

NON-QUALITY A

PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 65 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 49 of 51 Alert - AA2.2 Spent Fuel Pool Skimmer Surge Tank level is indicated on Ll-4815A (ref. 1).

PNPS Basis Reference(s):

1. PNPS 2.4.31, "Reactor Basin and/or Spent Fuel Pool Draindown"
2. EC 45088, Fukushima - Spent Fuel Pool Level Instrumentation

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 66 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 50 of 51 Alert - AA3.1 Category: A - Abnormal Rad Release/Rad Effluent Subcategory: 3 - MCR/CAS Radiation Initiating Condition: Rise in radiation levels within the facility that impedes operation of systems required to maintain plant safety functions EAL:

AA3.1 Alert Dose rates > 15 mR/hr in EITHER of the following areas requiring continuous occupancy to maintain plant safety functions:

Main Control Room (R1S-1815-2A, Panel C911)

OR CAS (by survey)

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses increased radiation levels that impact continued operation in areas requiring continuous occupancy to maintain safe operation or to perform a safe shutdown.

The cause and/or magnitude of the increase in radiation levels is not a concern of this EAL.

The Emergency Director must consider the source or cause of the increased radiation levels and determine if any other IC may be involved.

Areas requiring continuous occupancy include the Control Room and, as appropriate to the site, any other control stations that are staffed continuously, or a security alarm station.

~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 67 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.1 CATEGORY A, ABNORMAL RAD RELEASE/RAD EFFLUENT Sheet 51 of 51 Alert - AA3.1 PNPS Basis:

Areas that meet this threshold include the Main Control Room and the Central Alarm Station (CAS).

The Main C.ontrol Room Area Radiation Monitor (ARM) R1S-1815-2A on Panel C911 provides indication of area radiation levels in the Main Control Room. The high alarm of the Main Control Room ARM activates Control Room annunciator B7 ("CONTROL ROOM RAD HI") on Panel C904LC. The high alarm is set at approximately 1 mR/hr. (ref. 1, 2, 3)

The CAS area has no permanently installed area radiation monitors that may be used to assess this EAL threshold. Therefore this threshold must be assessed via local radiation survey (ref. 4).

  • PNPS Basis Reference(s):
1. PNPS 2.2.62, "Area Radiation Monitoring System"
2. PNPS 6.5-160, "Calibration of the Area Radiation Monitoring System"
3. PNPS ARP-C904LC-B7
4. PNPS 6.3-064, "Routine Radiological Surveillance Program"

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 68 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 1 of 70 Category H - Hazards EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety, or personnel safety.

The events of this category pertain to the following subcategories:

1. Natural or Destructive Phenomena Natural events include hurricanes, earthquakes, or tornados that have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety. Non-naturally occurring events that can cause damage to plant facilities and include aircraft crashes, missile impacts, etc.
2. Fire or Explosion Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of vital equipment.
3. Hazardous Gas Non-naturally occurring events that can cause damage to plant facilities and include toxic, corrosive, asphyxiant, or flammable gas leaks.
4. Security
  • Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.
5. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.
  • -: : -;: - Entergy . PNPS NON-QUALITY
  • -= EMERGENCY PLAN RELATED EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 69 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 2 of 70
6. Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE I

PROCEDURES REFERENCE USE Page 70 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 3 of 70 Unusual Event- HU1.1 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.1 Unusual Event Seismic event identified by any two of the following:

  • Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

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

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

a-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 71 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 4 of 70 Unusual Event - HU1 .1 PNPS Basis:

The method of detection with respect to emergency classification relies on the agreement of the shift operators on duty in the Control Room that the suspected ground motion is a felt earthquake as well as the actuation of the PNPS seismic instrumentation. Consensus of the Control Room operators with respect to ground motion helps avoid unnecessary classification if the seismic switches inadvertently trip or detect vibrations not related to an earthquake.

Ground motion acceleration of 0.01 g activates Control Room annunciator "SEISMIC RECORDER OPERATING" (C903R-B1) (ref. 1).

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

  • The SYSCOM seismic monitoring system is actuated by the trigger function of the recorder/sensors, which are located in the Reactor Building El. -17' CRD Quadrant, Reactor Building El. 23', and Reactor Building El. 91 '. The trigger setpoint is 0.01 g and the logic to actuate the Control Room annunciator is any one of the three recorder/sensors. The trigger actuation in any recorder/sensor will initiate the recording function in the other recorder/sensors and actuate Control Room indication for Operators in Panel C911. There are three indications to alert Operators that the trigger was actuated and that the seismic recording was/is activated (ref. 1).
  • The indicator panel that is part of ECU-370 has three red LEDs designated for the trigger function. There is one for each recorder/sensor; at least one of these is illuminated.
  • The red trigger LED for ECU-370 is illuminated indicating the logic for the system has been met. *
  • Control Room annunciator "SEISMIC RECORDER OPERATING" (C903R-B1) will be in alarm whenever the trigger senses a magnitude equal to or greater than 0.01g or 10mg.

This event escalates to an Alert under EAL HA1.1 if the earthquake exceeds Operating Basis Earthquake (OBE) levels.

PNPS Basis Reference(s):

1. PNPS 5.2.1, "Earthquake"

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 72 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 5 of 70 Unusual Event - HU1 .2 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.2 Unusual Event Tornado striking within Protected Area boundary OR Sustained high winds > 105 mph Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL is based on a tornado striking (touching down) or high winds within the Protected Area.

Escalation of this emergency classification level, if appropriate, would be based on visible damage or by other in plant conditions via EAL HA 1.2.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 73 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 6 of 70 Unusual Event - HU1 .2 PNPS Basis:

A tornado striking (touching down) within the Protected Area warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Design wind speed of 105 mph is calculated from the design basis wind loadings (ref. 1). The range of the Climatronics Model #100075 F460 Wind Speed Sensor is 0.5 to 145 mph (ref. 2).

PNPS Basis Reference(s):

1. SUDDSRF96-35 Wind and Tornado Evaluation for PNPS
2. Modification ER03114340

~--------------

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 74 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 7 of 70 Unusual Event - HU1 .3 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area (turbine)

EAL:

HU1.3 Unusual Event Turbine failure resulting in casing penetration or damage to turbine or generator seals Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant.

Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual fires and flammable gas build-up are appropriately classified via EAL HU2.1 and EAL HU3.1.

This EAL is consistent with the definition of a UE 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 EAL HA 1.4 based on damage done by projectiles generated by the failure or by the radiological releases. These latter events would be classified by the EALs in Category A or Category F.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PRO CEDURE PROCEDURES REFER ENCE USE Page 75 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 8 of 70 Unusual Event - HU1 .3 PNPS Basis:

The turbine generator stores large amounts of rotational kinetic energy in its rotor. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety related equipment.

In the event of missile ejection, the probability of a strike on a plant region is a function of the energy and direction of an ejected missile and of the orientation of the turbine with respect to the plant region.

  • PNPS Basis Reference(s):

None

~-----------------------------------

NON-QUALITY A

PNPS RELATED EP-AD-601 Revision 9

~Entergy - ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 76 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 9 of 70 Unusual Event - HU1 .4 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.4 Unusual Event Flooding in any Table H-2 area that has the potential to affect safety-related equipment needed for the current operating mode

- Reactor Building Table H-2 Internal Flooding Areas

- Rx Closed Cooling Water System Auxiliary Bays

- Turbine Building

- Diesel Generator Rooms

- Salt Service Water Bays

- Radwaste Area Mode Applicability:

All

  • ~Entergy. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EME RGENCY PLAN ADM INISTRATIVE PROCEDURE PRO CEDURES REFERENCE USE Page 77 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 10 of 70 Unusual Event - HU1 .4 NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses the effect of internal flooding caused by events such as component failures, equipment misalignment, or outage activity mishaps.

Escalation of this emergency classification level, if appropriate, would be based on visible damage via EAL HA1 .5 or by other plant conditions.

PNPS Basis:

The internal flooding areas of concern are listed in Table H-2 Internal Flooding Areas .

Flooding in these areas could have the potential to cause a reactor trip and could result in consequential failures to important systems. The potential for flooding in these areas was determined by an examination of piping systems in the area and also considered propagation of water from one area to another (ref. 1).

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.

PNPS Basis Reference(s):

1. PNPS-NE-07-00006 Rev. 0

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 78 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 11 of 70 Unusual Event - HU1 .5 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the Protected Area EAL:

HU1.5 Unusual Event Seawater bay water level> +13'6" MSL (Ll-3831A/B)

Seawater bay water level < -13'9" MSL (Ll-3831 A/B)

Mode Applicability:

All NEI 99-01 Basis:

This EAL is categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators.

This EAL addresses other site-specific phenomena (such as flood) that can also be precursors of more serious events.

PNPS Basis:

As illustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21 '6 (+21.5') MSL and well above the flood level of +13'6" MSL. Since the entrances to all structures containing equipment necessary for reactor shutdown and cooling are at elevations well above +13'6" MSL, they are protected against flooding from external sources. Seawater bay water level

< -13'9" MSL is the design minimum level for the SSW pumps.

  • -::::::-Entergy . PNPS NON-QUALITY
  • -= EMERGENCY PLAN RELATED PROCEDURE EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURES REFERENCE USE Page 79 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 12 of 70 Unusual Event - HU1 .5 PNPS Basis Reference(s):
1. PNPS 2.2.94, "Seawater System"
2. PNPS 2.4.154, "Intake Structure Fouling"
3. PNPS 2.2.32, "Salt Service Water System (SSW)"
4. PNPS 8.E.29, "Salt Service Water System Instrumentation Calibration" L___ _ _ _ _ _ - - -

.. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 e -

~=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 80 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 13 of 70 Unusual Event - HU1 .5 Figure H-1 Screen house Bay Water Levels SCRErNHOUSE GROUND LEVEL (EL. +21.S")

M!'.AH HIGl1 WAT"ER (E:L. +4.3")

M!'.AH SEA LEVEL (EL. 0.0')

MtAN LOW WATER (EL. -4.6')

BOTTOM OF SK!MMER WALL (EL. -12.D')

([I.. -20.1')

INTA.~E APRON (El.. - 24.5')

ELEV. VIEW

+16'0" Maximum rv1onitored Water Level

+*t3'6" Flood Level

+4'13" lvlean High Ttde O'O" Mean Sea Level

-4'8" Mean Lmv Tide

-7'0" Desicin Low Water Level

-'10'0" PNPS 2.4.154 action level to reduce Reactor pmver and secure affected Seawater Pump

-13'9" Desiqn Minimum Level for SSW Pumps

-14'6" Bottom of sluice gates between seawater bays and SSW bays

-15'0" Calculated worst case level. PNPS 2-4.154 action level to secure affected Seawater Pump (8' below design low water level).

-16'0" Minimum rv1onitored Water Level

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 81 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 14 of 70 Alert - HA1 .1 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area EAL:

HA1.1 Alert Seismic event identified by any two of the following:

Ground motion > Operating Basis Earthquake (0.08g) per analysis OR Control Room indication of degraded performance of systems required for the safe shutdown of the plant Mode Applicability:

All

e~Entergy- PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 82 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 15 of 70 Alert - HA 1 .1 NEI 99-01 Basis:

This EAL escalates from HU 1.1 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged but rather that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System

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 Information Center can confirm whether an earthquake has occurred in the area of the plant.

PNPS Basis:

Ground motion acceleration of 0.08g horizontal or vertical is the Operating Basis Earthquake for PNPS (ref. 1).

Ground motion acceleration of 0.01 g activates Control Room annunciator "SEISMIC RECORDER OPERATING" (C903R-B1) (ref. 1).

a-=- Entergy . NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 83 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 16 of 70 . i Alert- HA1 .1 The SYSCOM seismic monitoring system is actuated by the trigger function of the recorder/sensors, which are located in the Reactor Building El. -17' CRD Quadrant, Reactor Building El. 23', and Reactor Building El. 91 '. The trigger setpoint is 0.01 g and the logic to actuate the Control Room annunciator is any one of the three recorder/sensors. The trigger actuation in any recorder/sensor will initiate the recording function in the other recorder/sensors and actuate Control Room indication for Operators in Panel C911. There are three indications to alert Operators that the trigger was actuated and that the seismic recording was/is activated (ref. 1).

  • The indicator panel that is part of ECU-370 has three red LEDs designated for the trigger function. There is one for each recorder/sensor; at least one of these is illuminated.

The red trigger LED for ECU-370 is illuminated indicating the logic for the system has been met.

Control Room annunciator "SEISMIC RECORDER OPERATING" (C903R-B1) will be in alarm whenever the trigger senses a magnitude equal to or greater than 0.01 g or 10mg.

PNPS 5.2.1, "Earthquake," provides the guidance for determining whether the OBE earthquake threshold is exceeded by analys.is.

PNPS Basis Reference(s):

1. PNPS 5.2.1, "Earthquake"

e Entergy . NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=::=- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 84 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 17 of 70 Alert - HA1 .2 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area EAL:

HA1.2 Alert Tornado or sustained high winds> 105 mph within Protected Area boundary resulting in visible damage to any Table H-1 plant structures/equipment or Control Room indication of degraded performance of safety systems Table H-1 Safe Shutdown Areas

- Reactor Building

- Control Room

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank Mode Applicability:

All

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

~Entergy . ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 85 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 18 of 70 Alert - HA 1.2 NEI 99-01 Basis:

This EAL escalates from HU1 .2 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged but, rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System

This EAL is based on a tornado striking (touching down) or high winds that have caused visible damage to structures containing functions or systems required for safe shutdown of the plant.

PNPS Basis:

This threshold addresses events that may have resulted in Safe Shutdown Areas being subjected to forces (tornado or high winds> 105 mph, ref. 1) beyond design limits and thus damage may be assumed to have occurred to plant safety systems. Table H-1 Safe Shutdown Areas house equipment the operation of which may be needed to ensure the reactor safely reaches and is maintained shutdown (ref. 2, 3, 4).

A tornado striking (touching down) within the Protected Area resulting in visible damage warrants declaration of an Alert regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 86 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 19 of 70 Alert - HA 1.2 PNPS Basis Reference(s):

1. SUDDSRF96-35, Wind and Tornado Evaluation for PNPS
2. PNPS 5.5.1, "General Fire Procedure"
3. PNPS 5.5.2, "Special Fire Procedure"
4. PNPS 1.3.40, "Specifications For Vital Area Barrier Openings, Degradation, And Repair"
  • Entergy . PNPS NON-QUALITY

=~ EMERGENCY PLAN RELATED EP-AD-601 Revision 9

-===- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 87 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 20 of 70 Alert - HA 1.3 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area EAL:

HA1.3 Alert Vehicle crash resulting in visible damage to any Table H-1 plant structures or equipment or Control Room indication of degraded performance of safety systems

- Reactor Building

- Control Room

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank Mode Applicability:

All

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

- Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 88 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 21 of 70 Alert - HA 1.3 NEI 99-01 Basis:

This EAL escalates from an Unusual Event in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged but, rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL addresses vehicle crashes within the Protected Area that result in visible damage to vital areas or indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant.

PNPS Basis:

Table H-1 Safe Shutdown Areas house equipment the operation of which may be needed to ensure the reactor reaches and is maintained in shutdown (ref. 1, 2, 3).

If the vehicle crash is determined to be hostile in nature, the event is classified under EAL HS4.1.

PNPS Basis Reference(s):

1. PNPS 5.5.1, "General Fire Procedure"
2. PNPS 5.5.2, "Special Fire Procedure"
3. PNPS 1.3.40, "Specifications For Vital Area Barrier Openings, Degradation, And Repair

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 89 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 22 of 70 Alert - HA 1.4 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area.

EAL:

HA1.4 Alert Turbine failure-generated missiles result in any visible damage to or penetration of any Table H-1 area

  • - Reactor Building

- Control Room

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank Mode Applicability:

All

e-===:- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 90 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS

======================-

Sheet 23 of 70 Alert - HA 1.4 NEI 99-01 Basis:

This EAL escalates from HU1 .3 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged but, rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL addresses the threat to safety related equipment imposed by projectiles generated by main turbine rotating component failures. Therefore, this EAL is consistent with the definition of an Alert in that the potential exists for actual or substantial potential degradation of the level of safety of the plant.

PNPS Basis:

The turbine generator stores large amounts of rotational kinetic energy in its rotor. In the unlikely event of a major mechanical failure, this energy may be transformed into both rotational and translational energy of rotor fragments. These fragments may impact the surrounding stationary parts. If the energy-absorbing capability of these stationary turbine generator parts is insufficient, external missiles will be released. These ejected missiles may impact various plant structures, including those housing safety related equipment.

In the event of missile ejection; the probability of a strike on a plant region is.a function of the energy and direction of an ejected missile and of the orientation of the turbine with respect to the plant region.

Table H-1 Safe Shutdown Areas house equipment the operation of which may be needed to ensure the reactor safely reaches and is maintained shutdown (ref. 1, 2, 3).

  • ~Entergy. PNPS NON-QUALITY EMERGENCY PLAN RELATED EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 91 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 24 of 70 Alert - HA 1.4 PNPS Basis Reference(s):
1. PNPS 5.5.1, "General Fire Procedure"
2. PNPS 5.5.2, "Special Fire Procedure"
3. PNPS 1.3.40, "Specifications for Vital Area Barrier Openings, Degradation, and Repair'

e-~ Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 92 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 25 of 70 Alert - HA 1.5 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area EAL:

HA1.5 Alert Flooding in any Table H-2 area that results in degraded safety system performance as indicated in the Control Room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment Ta.ble H-2 Internal flooding Areas

- Reactor Building

- Rx Closed Cooling Water System Auxiliary Bays

- Turbine Building

- Diesel Generator Rooms

- Salt Service Water Bays

- Radwaste Area Mode Applicability:

All

~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 93 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 26 of 70 Alert - HA 1.5 NEI 99-01 Basis:

This EAL escalates from EAL HU1 .4 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged but, rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs .

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

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.

PNPS Basis:

The internal flooding areas of concern are listed in Table H-2 Internal Flooding Areas.

Flooding in these areas could have the potential to cause a reactor trip and could result in consequential failures to important systems. The potential for flooding in these areas was determined by an examination of piping systems in the areas and also considered propagation of water from one area to another (ref. 1).

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 94 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 27 of 70 Alert - HA 1.5 PNPS Basis Reference(s):

1. PNPS-NE-07-00006 Rev. 0

-=:- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE

. PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 95 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 28 of 70 Alert - HA 1.6 Category: H - Hazards Subcategory: 1 - Natural or Destructive Phenomena Initiating Condition: Natural or destructive phenomena affecting the plant Vital Area EAL:

HA1.6 Alert Seawater bay water level> +16'0" MSL (Ll-3831A/B)

OR Seawater bay water level< -16'0" MSL(Ll-3831A/B)

Mode Applicability:

All NEI 99-01 Basis:

This EAL escalates from HU1 .5 in that the occurrence of the event has resulted in visible damage to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by Control Room indications of degraded system response or performance. The occurrence of visible damage and/or degraded system response is intended to discriminate against lesser events. The initial report should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage.

The significance here is not that a particular system or structure was damaged but, rather, that the event was of sufficient magnitude to cause this degradation.

Escalation of this emergency classification level, if appropriate, would be based on System Malfunction EALs.

This EAL addresses other PNPS phenomena that result in visible damage to vital areas or results in indication of damage to safety structures, systems, or components containing functions and systems required for safe shutdown of the plant (such as flood) that can also be

  • precursors of more serious events .

NON-QUALITY A

PNPS EP-AD-601 RELATED Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 96 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 29 of 70 Alert - HA 1.6 PNPS Basis:

As illustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21 '6" (+21.5') MSL and well above the flood level of +13'6" MSL. Since the entrances to all structures containing equipment necessary for reactor shutdown and cooling are at elevations well above +16'0" MSL, they are protected against flooding from external sources. The specified water levels are the maximum and minimum monitored seawater bay water levels.

PNPS Basis Reference(s):

1. PNPS 2.2.94, "Seawater System"
2. PNPS 2.4.154, "Intake Structure Fouling"
3. PNPS 2.2.32, "Salt Service Water System (SSW)"
4. PNPS 8.E.29, "Salt Service Water System Instrumentation Calibration"
  • -::::::-Entergy . PNPS NON-QUALITY

= - .. RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 97 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 30 of 70 Alert - HA 1 .6 Figure H-1 Screen house Bay Water Levels SCR£E..'1HOUS£ GROUND U:VEL (EL. t*21.5')

STOP LOO

.- GUIDE MtAll HIGH WATER (CL. +4.J')

  • (EL. -20.1')

MEAN SEA LEVEL (EL. 0.0')

1'-EAN LOW WATCR (EL. -4.6')

sono1r,1 or SKIMMER WAlL (EL.

  • 12.0')

INTAKE APRON (EL. - 24.5')

ELEV. VIEW

+*t6'0" Maximum Monitored Water Level

+13'6" Flood Level

+413" Mean High Tide O'O" Mean Sea Level

-4'8" Mean Lmv Tlde

-7'0" DesiQn Low \iVater Level

-10'0" PNPS 2-4.154 action level to reduce Reactor power and secure affected Seawater Pump

-*13'9" DesiQn Minimum Level for SSW Pumps

-14'6" Bottom of sluice gates between seawater bays and SSW bays

-15'0" Calculated worst case level. PNPS 2.4.154 action level to secure affected Seawater Pump (8' belo*w design low 'Nater level).

  • -*16'0" Minimum Monitored Water Level

.. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 98 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 31 of 70 Unusual Event - HU2.1 Category: H - Hazards Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within 15 min. of detection or explosion within Protected Area boundary EAL:

HU2.1 Unusual Event Fire not extinguished within 15 min. of Control Room notification or verification of a Control Room fire alarm in any Table H-1 area (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Tabl~ H-1 ~afe Shutdown Areas

- Reactor Building

- Control Room

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank

  • ===- Entergy

- -= -

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 99 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 32 of 70 Unusual Event - HU2.1 Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses the magnitude and extent of fires that may be potentially significant precursors of damage to safety systems. It addresses the fire and not the degradation in performance of affected systems that may result.

As used here, detection is visual observation and report by plant personnel or sensor alarm indication.

The 15-minute time period begins with a credible notification that a fire is occurring or indication of a fire detection system alarm/actuation. Verification of a fire detection system alarm/actuation includes actions that can be taken within the Control Room or other nearby site-specific location to ensure that it is not spurious. An alarm is assumed to be an indication of a fire unless it is disproved within the 15-minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.

The intent of this 15-minute duration is to size the fire and to discriminate against small fires that are readily extinguished (e.g., smoldering waste paper basket).

Escalation of this emergency classification level, if appropriate, would be based on EAL HA2.1.

PNPS Basis:

Fire, as used in this EAL, means combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires.

Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

Fires that are not within an H-1 area do not meet the criteria for an UE classification. If a fire alarm in an H-1 area cannot be disproved within the 15 minutes allowed, the UE needs to be performed .

-~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 100 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 33 of 70 Unusual Event - HU2.1 PNPS Basis Reference(s):

1. PNPS 5.5.1, "General Fire Procedure"
2. PNPS 5.5.2, "Special Fire Procedure"
3. PNPS 1.3.40, "Specifications for Vital Area Barrier Openings, Degradation, and Repair

NON-QUALITY

~ PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 101 of 351 Emergency Action Level Technical Bases Document ATTACHMENT9.2 CATEGORY H, HAZARDS Sheet 34 of 70 Unusual Event - HU2.2 Category: H - Hazards Subcategory: 2 - Fire or Explosion Initiating Condition: Fire within the Protected Area not extinguished within 15 min. of detection or explosion within Protected Area boundary EAL:

HU2.2 Unusual Event Explosion within the Protected Area

  • Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses the magnitude and extent of explosions that may be potentially significant precursors of damage to safety systems. It addresses the explosion and not the degradation in performance of affected systems that may result.

As used here, detection is visual observation and report by plant personnel or sensor alarm indication.

This EAL addresses only those explosions of sufficient force to damage permanent structures or equipment within the Protected Area.

No attempt is made to assess the actual magnitude of the damage. 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 EAL HA2.1 .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 102 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 35 of 70 Unusual Event - HU2.2 PNPS Basis:

As used here, an explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials.

  • A steam line break or steam explosion that damages surrounding permanent structures or equipment would be classified under this EAL. This does not mean the emergency is classified simply because the steam line break occurred. The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EALs in Category F).

If the explosion is determined to be hostile in nature, the event is classified under security

PNPS Basis Reference(s):

None

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 103 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 36 of 70 Alert - HA2.1 Category: H - Hazards Subcategory: 2 - Fire or Explosion Initiating Condition: Fire or explosion affecting the operability of plant safety systems required to establish or maintain safe shutdown EAL:

HA2.1 Alert Fire or explosion resulting in visible damage to any Table H-1 area containing safety systems or components or Control Room indication of degraded performance of safety systems

  • - Reactor Building

- Control Room

- -; -~

Table H.:.1 Safe ShutdowrfAreas

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank Mode Applicability:

All

- ='" Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 104 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 37 of 70 Alert - HA2.1 NEI 99-01 Basis:

Visible damage is used to identify the magnitude of the fire or explosion and to discriminate against minor fires and explosions.

The reference to structures containing safety systems or components is included to discriminate against fires or explosions in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the fire or explosion was large enough to cause damage to these systems.

The use of visible damage should not be interpreted as mandating a lengthy damage assessment prior to classification. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform detailed damage assessments.

The Emergency Director also needs to consider any security aspects of the explosion.

Escalation of this emergency classification level, if appropriate, will be based on system malfunctions, fission product barrier degradation, or abnormal rad levels/radiological effluent EALs.

PNPS Basis:

Fire, as used in this EAL, means combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires.

Observation of flame is preferred but is not required if large quantities of smoke and heat are observed.

An explosion is a rapid, violent, unconfined combustion or a catastrophic failure of pressurized equipment that potentially imparts significant energy to nearby structures and materials.

A steam line break or steam explosion that damages permanent structures or equipment would be classified under this EAL. The method of damage is not as important as the degradation of plant structures or equipment. The need to classify the steam line break itself is considered in fission product barrier degradation monitoring (EAL Category F).

  • -e: ===Entergy

-===- .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 105 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 38 of 70 Alert - HA2.1 PNPS Basis Reference(s):

1. PNPS 5.5.1, "General Fire Procedure"
2. PNPS 5.5.2, "Special Fire Procedure"
3. PNPS 1.3.40, "Specifications For Vital Area Barrier Openings, Degradation, And Repair"

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 106 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 39 of 70 Unusual Event - HU3.1 Category: H - Hazards Subcategory: 3 - Hazardous Gas Initiating Condition: Release of toxic, corrosive, asphyxiant or flammable gases deemed detrimental to normal plant operations EAL:

HU3.1 Unusual Event Toxic, corrosive, asphyxiant or flammable gases in amounts that have or could adversely affect normal plant operations Mode Applicability:

All NEI 99-01 Basis:

This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect normal plant operations.

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

This 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 EAL HA3.1.

a~-Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 107 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 40 of 70 Unusual Event - HU3.1 PNPS Basis:

As used in this EAL, affecting normal plant operations means that activities at the plant site associated with routine testing, maintenance, or *equipment operations, in accordance with normal operating or administrative procedures, have been impacted. For example, the discharge of hydrogen gas from the main generator during plant shutdown activities would not constitute an emergency classification unless the release adversely affected normal plant operations. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from normal plant operations and thus would be considered to have been affected.

The release may have originated within the plant site areas or it may have originated offsite and subsequently drifted onto plant site areas. Offsite events (e.g., tanker truck accident

  • releasing toxic gases, etc.) resulting in the plant being within the evacuation area should also be considered in this EAL because of the adverse affect on normal plant operations.

Some gases are toxic by their very nature. Others, like carbon dioxide, can be lethal if it reduces oxygen to low concentrations (asphyxiant) that are immediately dangerous to life and health (IDLH). Oxygen deficient atmospheres (less than 19.5% oxygen) are considered IDLH (ref. 1). NRC position is that any time carbon dioxide is discharged in plant areas such that the area becomes uninhabitable, regardless of whether anyone is in the areas, conditions for classification exist. The EAL assumes an uncontrolled process that has the potential to affect plant operations or personnel safety. Releases occurring during planned surveillance activities or planned maintenance/tagout activities, therefore, are excluded.

The following documents provide additional information on hazardous substances and spills.

  • PNPS 5.5.4, "Response to Hazardous Material Incidents" (ref. 2)
  • Regulatory Guide 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits (IDLH Limits) for Some Hazardous Chemicals Should the release affect access to plant safe shutdown areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1 .
    • -Entergy .
  • e PNPS EMERGENCY PLAN NON-QUALITY RELATED EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURES PROCEDURE REFERENCE USE Page 108 of 351 *1 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 41 of 70 Unusual Event - HU3.1 PNPS Basis Reference(s):
1. PNPS 1.4.12, "Primary Containment Entry"
2. PNPS 5.5.4, "Response to Hazardous Material Incidents"

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 109 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 42 of 70 Unusual Event - HU3.2 Category: H - Hazards Subcategory: 3 - Hazardous Gas Initiating Condition: Release of toxic, corrosive, asphyxiant or flammable gases deemed detrimental to normal operation of the plant EAL:

HU3.2 Unusual Event Recommendation by local, county, or state officials to evacuate or shelter site personnel based on offsite event

  • Mode Applicability:

All NEI 99-01 Basis:

This EAL is based on the release of toxic, corrosive, asphyxiant or flammable gases of sufficient quantity to affect normal plant operations.

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

This 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 enclos*ed 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 EAL HA3.1 .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 110 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 43 of 70 Unusual Event - HU3.2 PNPS Basis:

This EAL is based on the existence of an uncontrolled release originating offsite and local, county, or state officials have reported the need for evacuation or sheltering of site personnel.

Offsite events (e.g., tanker truck accident releasing toxic gases, etc.) are considered in this EAL because they may adversely affect normal plant operations.

State officials may determine the evacuation area for offsite spills by using the Department of Transportation (DOT) Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials.

Should the release affect plant safe shutdown areas, escalation to an Alert would be based on EAL HA3.1. Should an explosion or fire occur due to flammable gas within an affected plant area, an Alert may be appropriate based on EAL HA2.1.

PNPS Basis Reference(s):

None

a*~Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE . Page 111 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 44 of 70 Alert - HA3.1 Category: H - Hazards Subcategory: 3 - Hazardous Gas Initiating Condition: Access to a vital area is prohibited due to release of toxic, corrosive, asphyxiant or flammable gases which jeopardizes operation of systems required to maintain safe operations or safely shut down the reactor EAL:

HA3.1 Alert Access to any Table H-1 area is prohibited due to toxic, corrosive, asphyxiant or flammable gases which jeopardize operation of systems required to maintain safe operations or safely shut down the reactor (Note 8)

Note 8: 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

. . .. . . . .. ' ' . *. . *, . \ - . ' '. ' . <'...

. *Table H-'1 Safe Shutdown *Areas*

- Reactor Building

- Control Room

- Cable Spreading Room

- 4160 Switchgear Rooms

- Diesel Generator Rooms

- Salt Service Water Bays

- Rx Closed Cooling Water System Auxiliary Bays

- Standby Gas Treatment Area, 51' Level, Turbine Bldg

- Condensate Storage Tank

RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 112 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 45 of 70 Alert - HA3.1 Mode Applicability:

All NEI 99-01 Basis:

Gases in a vital area can affect the ability to safely operate or safely shut down the reactor.

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

Declaration should not be delayed for confirmation from atmospheric testing if the atmosphere poses an immediate threat to life and health or an immediate threat of severe exposure to gases. This could be based upon documented analysis, indication of personal ill effects from exposure, or operating experience with the hazards.

If the equipment in the stated area was already inoperable or out of service before the event occurred, then this EAL should not be declared as it will have no adverse impact on the ability of the plant to safely operate or safely shut down beyond that already allowed by Technical Specifications at the time of the event.

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness, or even death.

An uncontrolled release of flammable gases 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 gases, 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 gases which can ignite/support combustion.

Escalation of this emergency classification level, if appropriate, will be based on system malfunctions, fission product barrier degradation, or abnormal rad levels/radioactive effluent EALs.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 113 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 46 of 70 Alert - HA3.1 PNPS Basis:

This EAL is based on gases that have entered a plant structure in concentrations that could be unsafe for plant personnel and, therefore, preclude access to equipment necessary for the safe operation of the plant. Table H-1 Safe Shutdown Areas contains systems that are operated to establish or maintain safe shutdown (ref. 1, 2, 3).

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

  • Flammable gases, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipmenUcomponents (acetylene - used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gases 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.

This EAL does not apply to routine inerting of the Primary Containment.

The following documents provide additional information on hazardous substances and spills.

  • PNPS 5.5.4, "Response to Hazardous Material Incidents" (ref. 4)
  • Regulatory Guide 1. 78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, Table 1, Toxicity Limits

.(IDLH Limits) for Some Hazardous Chemicals PNPS Basis Reference(s):

1. PNPS 5.5.1, "General Fire Procedure"
2. PNPS 5.5.2, "Special Fire Procedure"
3. PNPS 1.3.40, "Specifications for Vital Area Barrier Openings, Degradation, and Repair"
4. PNPS 5.5.4, "Response to Hazardous Material Incidents"

a Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=:- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 114 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 47 of 70 Unusual Event - HU4.1 Category: H - Hazards Subcategory: 4 - Security Initiating Condition: Confirmed security condition or threat which indicates a potential degradation in the level of safety of the plant EAL:

HU4.1 Unusual Event A security condition that does not involve a hostile action as reported by the Station Security Force OR A credible site-specific security threat notification OR A validated notification from NRC providing information of an aircraft threat Mode Applicability:

All NEI 99-01 Basis:

Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.

Security events which do not represent a potential degradation in the level of safety of the plant are reported under 10CFR73.71 or in some cases under 10CFR50.72. Security events assessed as hostile actions are classifiable under EALs HA4.1, HS4.1, and HG4.1.

A higher initial classification could be made based upon the nature and timing of the security threat and potential consequences. The licensee shall consider upgrading the emergency response status and emergency classification level in accordance with the site's Safeguards Contingency Plan and Emergency Plan.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 115 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 48 of 70 Unusual Event - HU4.1 First Threshold Reference is made to PNPS Security Force because these individuals are the designated personnel onsite 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 PNPS Safeguards Contingency Plan.

This threshold is based on PNPS security plans. PNPS Safeguards Contingency Plans are based on guidance provided by NEI 03-12.

Second Threshold This threshold is included to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a credible threat. Only the plant to which the specific threat is made need declare the Unusual Event.

The determination of "credible" is made through use of information found in the PNPS Safeguards Contingency Plan.

Third Threshold The intent of this threshold is to ensure that notifications for the aircraft threat are made in a timely manner and that offsite response organizations (OROs) and plant personnel are at a state of heightened awareness regarding the credible threat. It is not the intent of this threshold to replace existing nonhostile-related EALs involving aircraft.

This threshold is met when PNPS receives information regarding an aircraft threat from the 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 PNPS 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 the Alert emergency classification level via EAL HA4.1 would be appropriate if the threat involves an airliner within 30 minutes of the plant.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 116 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 49 of 70 Unusual Event - HU4.1 PNPS Basis:

Security EALs could involve events due to nonhostile and hostile actions. It is important to understand the differences between these actions and therefore the definitions of hostile action and hostile force are provided with this basis.

Hostile Action: An act toward PNPS or its plant personnel that includes the use of violent force to destroy equipment, takes hostages, and/or intimidates plant personnel 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 PNPS. Nonterrorism-based EALs should be used to address such activities (e.g., violent acts between individuals in the Owner Controlled Area.)

Hostile Force: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

FIRST THRESHOLD: The declaration of an Unusual Event for a significant security condition (e.g., an attempted unauthorized entry into the Protected Area) is appropriate to ensure positive action is taken to protect personnel safety and to notify offsite law enforcement agencies.

The term" ... as reported by the Station Security Force" infers that Station Security has evaluated the incident and has determined it is a significant security condition. These individuals are the designated onsite personnel 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 controls placed on the PNPS Safeguards Contingency Plan.

SECOND THRESHOLD: The declaration of an Unusual Event for any credible security threat is appropriate to ensure positive action is taken to maintain site security, protect personnel safety, and notify offsite law enforcement agencies. The term "credible security threat" refers to a threat which in the judgment of the Station Security Force is credible as specified in the PNPS Safeguards Contingency Plan.

THIRD THRESHOLD: PNPS 5.3.14.1, Airborne Threat", describes the actions to be taken when notified of an airborne threat.

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 117 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 50 of 70 Unusual Event - HU4.1 PNPS Basis Reference(s):

1. NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security Based Events", July 18, 2005
2. NEI White Paper, "Enhancements to Emergency Preparedness Programs for Hostile Action", November 18, 2005
3. PNPS 5.3.14, "Security Incidents"
4. PNPS 5.3.14.1, Airborne Threat"

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 118 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 51 of 70 Alert - HA4.1 Category: H - Hazards Subcategory: 4 - Security Initiating Condition: Hostile action within the Owner Controlled Area or airborne attack threat EAL:

HA4.1 Alert A hostile action is occurring or has occurred within the Owner Controlled Area as reported by the Station Security Force OR A validated notification from NRG of an airliner attack threat within 30 min. of the site Mode Applicability:

All NEI 99-01 Basis:

Note: Timely and accurate communication between Security Shift Supervision and the Control Room is crucial for the implementation of effective Security EALs.

This EAL addresses 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 onsite evacuation, dispersal, or sheltering).

~Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 119 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 52 of 70 Alert - HA4.1 First Threshold This threshold addresses the potential for a very rapid progression of events due to a hostile action. It is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the OCA. Those events are adequately addressed by other EALs.

Note that this threshold is applicable for any hostile action occurring, or that has occurred, in the Owner Controlled Area.

Second Threshold This threshold addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time .

The intent of this threshold is to ensure that notifications for the airliner attack threat are made in a timely manner and that offsite response organizations (OROs) 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 threshold is met when a plant receives information regarding an airliner attack threat from the 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 .

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

~~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 120 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 53 of 70 Alert - HA4.1 PNPS Basis:

Security EALs could involve events due to nonhostile and hostile actions. It is important to understand the differences between these actions and therefore the definitions of hostile action and hostile force are provided with this basis.

Hostile Action: An act toward PNPS or its plant personnel that includes the use of violent force to destroy equipment, takes hostages, and/or intimidates plant personnel 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 PNPS. Nonterrorism-based EALs should be used to address such activities (e.g., violent acts between individuals in the Owner Controlled Area.)

Hostile Force: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception*, equipped with suitable weapons capable of killing, maiming, or causing destruction.

FIRST THRESHOLD: This EAL is intended to address the potential for a very rapid progression of events due to an attack including:

  • Air attack (aircraft impacting the Owner Controlled Area)
  • Land-based attack (hostile force progressing across PNPS property or directing projectiles at the site)
  • Waterborne attack (hostile force on water attempting forced entry or directing projectiles at the site)
  • Bombs This EAL is intended to address the contingency for a very rapid progression of events due to a hostile attack and the possibility for additional attack. This EAL is not premised solely on the potential for a radiological release. Rather, the issue includes the need for assistance due to the possibility for significant and indeterminate damage from additional attack elements. It is appropriate for offsite response organizations to be notified and to activate in order to be better prepared to respond should protective actions become necessary.

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 121 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 54 of 70 Alert - HA4.1 For the purpose of this EAL, "Owner Controlled Area (OCA)" is the Entergy owned property on the north side of Rocky Hill Road.

SECOND THRESHOLD: PNPS 5.3.14.1, Airborne Threat", describes the action to be taken when notified of an airborne threat.

PNPS Basis Reference(s):

1. NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security Based Events", July 18, 2005
2. NEI White Paper, "Enhancements to Emergency Preparedness Programs for Hostile Action", November 18, 2005
  • 3. PNPS 5.3.14, "Security Incidents"
4. PNPS 5.3.14.1, "Airborne Threat"

e~Entergy. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 122 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 55 of 70 Site Area Emergency - HS4.1 Category: H - Hazards Subcategory: 4 - Security Initiating Condition: Hostile action within the Protected Area EAL:

HS4.1 Site Area Emergency A hostile action is occurring or has occurred within the Protected Area as reported by the Station Security Force Mode Applicability:

All NEI 99-01 Basis:

This condition represents an escalated threat to plant safety above that contained in the Alert in that a hostile force has progressed from the Owner Controlled Area to the Protected Area.

This EAL addresses the contingency for a very rapid progression of events, such as that experienced on September 11, 2001. It is not premised solely on the potential for a radiological release. Rather, the issue includes the need for rapid assistance due to the possibility for significant and indeterminate damage from additional air, land, or water attack elements.

The fact that the site is under serious attack with minimal time available for further preparation or additional assistance to arrive requires offsite response organizations' (ORO) 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 is not intended to address incidents that are accidental events or acts of civil disobedience, such as small aircraft impact, hunters, or physical disputes between employees within the Protected Area. Those events are adequately addressed by other EALs.

NON-QUALITY A PNPS.

RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 123 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 56 of 70 Site Area Emergency - HS4.1 Escalation of this emergency classification level, if appropriate, would be based on actual plant status after impact or progression of attack.

PNPS Basis:

Security EALs could involve events due to nonhostile and hostile actions. It is important to understand the differences between these actions and therefore the definitions of hostile action and hostile force are provided with this basis.

Hostile Action: An act toward PNPS or its plant personnel that includes the use of violent force to destroy equipment, takes hostages, and/or intimidates plant personnel 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 PNPS. Nonterrorism-based EALs should be used to address such activities (e.g., violent acts between individuals in the Owner Controlled Area.)

Hostile Force: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This EAL is intended to address the potential for a very rapid progression of events due to an attack including:

  • Air attack (aircraft impacting the Protected Area)
  • Land-based attack (hostile force penetrating the Protected Area)
  • Waterborne attack (hostile force on water penetrating the Protected Area)
  • Bombs breaching the Protected Area This EAL is intended to address the contingency for a very rapid progression of events due to a hostile attack and the possibility for additional attack. This EAL is not premised solely on the potential for a radiological release. Rather, the issue includes the need for assistance due to the possibility for significant and indeterminate damage from additional attack elements .
  • ~Entergy- PNPS NON-QUALITY

= =-- EMERGENCY PLAN RELATED PROCEDURE EP-AD-601 Revision 9 ADMINISTRATIVE PROCEDURES REFERENCE USE Page 124 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 57 of 70 Site Area Emergency - HS4.1 This EAL addresses the immediacy of a threat to impact site vital areas within a relatively short time. The fact that the site is under serious attack with minimal time available for additional assistance to arrive requires offsite response organizations' readiness and preparation for the implementation of protective measures.

PNPS Basis Reference(s):

1. NRC Bulletin 2005-02, "Emergency Preparedness and Response Actions for Security Based Events", July 18, 2005
2. NEI White Paper, "Enhancements to Emergency Preparedness Programs for Hostile Action", November 18, 2005
3. PNPS 5.3.14, "Security Incidents"
4. PNPS 5.3.14.1, Airborne Threat"

e1"$P*e.

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 125 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 58 of 70 General Emergency - HG4.1 Category: H - Hazards Subcategory: 4 .:. Security Initiating Condition: Hostile action resulting in loss of physical control of the facility EAL:

HG4.1 General Emergency A hostile action has occurred such that plant personnel are unable to operate equipment required to maintain safety functions

  • OR A hostile action has caused failure of spent fuel cooling systems and imminent fuel damage is likely for recently irradiated fuel (fuel has decayed < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> since reactor shutdown)

Mode Applicability:

All NEI 99-01 Basis:

First Threshold 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.

If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the threshold is not met.

NON-QUALITY 1.916 PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 126 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 59 of 70 General Emergency - HG4.1 Second Threshold 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.

PNPS Basis:

The first threshold encompasses conditions under which a hostile force has taken control of plant equipment (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. Typically 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). Temporary lineups to allow heat to be absorbed in

  • the Torus are not considered a long term heat sink.

Loss of physical control of the Control Room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account. If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

For the second threshold the term "recently irradiated fuel" is defined as fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (i.e., reactor fuel that has decayed less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following reactor shutdown) (ref. 1).

PNPS Basis Reference(s):

1. Technical Specifications Section 3.7.C
2. PNPS 5.3.14, "Security Incidents"
3. PNPS 5.3.14.1, Airborne Threat"

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 127 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 60 of 70 Alert - HA5.1 Category: H - Hazards Subcategory: 5 - Control Room Evacuation

  • Initiating Condition: Control Room evacuation has been initiated EAL:

HAS.1 Alert PNPS 2.4.143, "Shutdown from Outside Control Room," requires Control Room evacuation

  • Mode Applicability:

All NEI 99-01 Basis:

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

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

PNPS Basis:

PNPS 2.4.143, "Shutdown from Outside Control Room," provides the instructions for scramming the unit and maintaining RCS inventory from outside the Control Room. The Shift Manager (SM) determines whether the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.

PNPS Basis Reference(s):

1. PNPS 2.4.143, "Shutdown from Outside Control Room"

NON-QUALITY A PNPS

~* RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 128 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 61 of 70 Site Area Emergency - HS5.1 Category: H - Hazards Subcategory: 5 - Control Room Evacuation Initiating Condition: Control Room evacuation has been initiated and plant control cannot be established EAL:

HSS.1 Site Area Emergency Control Room evacuation has been initiated AND Control of the plant cannot be established within 15 min.

Mode Applicability:

All NEI 99-01 Basis:

The intent of this EAL is to capture those events where control of the plant cannot be re-established in a timely manner. In this case, expeditious transfer of control of safety systems has not occurred (although fission product barrier damage may not yet be indicated).

The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions.

Typically, these safety functions are reactivity control (ability to shut down the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink) I.

The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director (ED) judgment. The Emergency Director is expected to make a reasonable, informed judgment within the site-specific time for transfer that the licensee has control of the plant from the remote shutdown panel.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 129 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 62 of 70 Site Area Emergency - HS5.1 Escalation of this emergency classification level, if appropriate, would be by fission product barrier degradation or abnormal rad levels/radiological effluent EALs.

PNPS Basis:

PNPS 2.4.143, "Shutdown from Outside Control Room," provides the instructions for scramming the unit and maintaining RPV inventory from outside the Control Room. The Shift Manager determines whether the Control Room is inoperable and requires evacuation.

Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.

PNPS Basis Reference(s):

1. PNPS 2.4.143, "Shutdown from Outside Control Room"
  • =- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 130 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 63 of 70 Unusual Event - HU6.1 Category: H - Hazards Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU6.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate EITHER:

A potential degradation of the level of safety of the plant

  • A security threat to facility protection has been initiated No releases of r_adioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the UE emergency classification level.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-===- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 131 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 64 of 70 Unusual Event - HU6.1 PNPS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the PNPS Emergency Plan. Initially, the Shift Manager or the available on-shift Senior Reactor Operator (SRO) assumes the duties and responsibilities as the Emergency Director. When augmentation of the on-shift complement occurs, the Senior Nuclear Executive or a designated alternate reports to the EOF and, once briefed, relieves the Shift Manager of all Emergency Director responsibilities. Once the on-call Emergency Director assumes the Emergency Director responsibilities, overall command and control of the emergency transfers from the Control Room to the EOF. The Emergency Plant Operations Supervisor (EPOS) may relieve the on-shift Emergency Director until such time as the on-call Emergency Director arrives; however, the EPOS must report and remain in the Control Room until relieved (ref. 1).

  • The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

1. PNPS Emergency Plan Section B-3
2. Refer to PNPS EP-IP-100, "Emergency Classification and Notification", for definition of "Unusual Event" .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 132 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 65 of 70 Alert - HA6.1 Category: H - Hazards Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.

EAL:

HA6.1 Alert Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

An actual or potential substantial degradation of the level of safety of the plant 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 (1,000 mrem TEDE and 5,000 mrem thyroid CDE)

Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency classification level.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 133 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 66 of 70 Alert - HA6.1 PNPS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the PNPS Emergency Plan. Initially, the Shift Manager or the available on-shift Senior Reactor Operator (SRO) assumes the duties and responsibilities as the Emergency Director. When augmentation of the on-shift complement occurs, the Senior Nuclear Executive or a designated alternate reports to the EOF and, once briefed, relieves the Shift Manager of all Emergency Director responsibilities. Once the on-call Emergency Director assumes the Emergency Director responsibilities, overall command and control of the emergency transfers from the Control Room to the EOF. The Emergency Plant Operations Supervisor (EPOS) may relieve the on-shift Emergency Director until such time as the on-call Emergency Director arrives; however, the EPOS must report and remain in the Control Room until relieved (ref: 1) .

  • The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

1. PNPS Emergency Plan Section B-3
2. Refer to PNPS EP-IP-100, "Emergency Classification and Notification", for definition of "Alert" .

e Entergy . NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=::::- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 134 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 67 of 70 Site Area Emergency - HS6.1 Category: H - Hazards Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

HS6.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

An actual or likely major failure of plant functions needed for protection of the public OR Hostile action that results in intentional damage or malicious acts 1) toward site personnel or equipment that could lead to the likely failure of or 2) that prevent effective access to equipment needed for the protection of the public Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels (1,000 mrem TEDE and 5,000 mrem thyroid COE) beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for Site Area Emergency.

a-===- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 135 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 68 of 70 Site Area Emergency - HS6.1 PNPS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the PNPS Emergency Plan. Initially, the Shift Manager or the available on-shift Senior Reactor Operator (SRO) assumes the duties and responsibilities as the Emergency Director. When augmentation of the on-shift complement occurs, the Senior Nuclear Executive or a designated alternate reports to the EOF and, once briefed, relieves the Shift Manager of all Emergency Director responsibilities. Once the on-call Emergency Director assumes the Emergency Director responsibilities, overall command and control of the emergency transfers from the Control Room to the EOF. The Emergency Plant Operations Supervisor (EPOS) may relieve the on-shift Emergency Director until such time as the on-call Emergency Director arrives; however, the EPOS must report and remain in the Control Room until relieved (ref. 1).

  • The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

1. PNPS Emergency Plan Section B-3
2. Refer to PNPS EP-IP-100, "Emergency Classification and Notification", for definition of "Site Area Emergency" .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-===- Entergy .

EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 136 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 69 of 70 General Emergency - HG6.1 Category: H - Hazards Subcategory: 6 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

HG6.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve EITHER:

Actual or imminent substantial core degradation or melting with potential for loss of containment integrity Hostile action that results in an actual loss of physical control of the facility Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels (1,000 mrem TEDE and 5,000 mrem thyroid COE) beyond the site boundary Mode Applicability:

All NEI 99-01 Basis:

This EAL addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for General Emergency.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 137 of 351 Emergency Action Level Technical Bases Document

~;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;:;;;;;;;;;;;;

ATTACHMENT 9.2 CATEGORY H, HAZARDS Sheet 70 of 70 I General Emergency - HG6.1 PNPS Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the PNPS Emergency Plan. Initially, the Shift Manager or the available on-shift Senior Reactor Operator (SRO) assumes the duties and responsibilities as the Emergency Director. When augmentation of the on-shift complement occurs, the Senior Nuclear Executive or a designated alternate reports to the EOF and, once briefed, relieves the Shift Manager of all Emergency Director responsibilities. Once the on-call Emergency Director assumes the Emergency Director responsibilities, overall command and control of the emergency transfers from the Control Room to the EOF. The Emergency Plant Operations Supervisor (EPOS) may relieve the on-shift Emergency Director until such time as the on-call Emergency Director arrives; however, the EPOS must report and remain in the Control Room until relieved (ref. 1).

    • The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

1. PNPS Emergency Plan Section B-3
2. Refer to PNPS EP-IP-100, "Emergency Classification and Notification", for definition of "General Emergency" .

e--=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 138 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 1 of 52 Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature> 212°F); EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety.

The events of this category pertain to the following subcategories:

1. Loss of AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity.
2. ATWS/Criticality Events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification however, ATWS is intended to mean any scram failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS, and Primary Containment integrity. Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy- EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 139 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 2 of 52

3. Inability to Reach Shutdown Conditions One EAL falls into this subcategory. It is related to the failure of the plant to be brought to the required plant operating condition required by the Technical Specifications if a limiting condition for operation (LCO) is not met.
4. Instrumentation/Communications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Loss of annunciators or indicators is in this subcategory.

Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.

5. Fuel Clad Degradation During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these baseline levels (2% - 5% clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of clad damage may result in coolant activity exceeding Technical Specifications limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.
6. RCS Leakage The reactor vessel provides a volume for the coolant that covers the reactor core. The reactor vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail.

Excessive RCS leakage greater than Technical Specifications limits is utilized to indicate potential pipe cracks that may propagate to an extent threatening fuel clad, RCS, and Primary Containment integrity.

~Entergy- EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 140 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 3 of 52

7. Loss of DC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category involves total losses of vital plant 125V DC power sources.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 141 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 4 of 52 Unusual Event - SU1 .1 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite AC power to emergency buses for 15 minutes or longer EAL:

SU1.1 Unusual Event Loss of all offsite AC power (Table S-3) to emergency buses AS and A6 for 2':. 15 min.

(Note 3)

  • Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table 5-3 AC Power Sources Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

- SBO DG

e Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-:::::=- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 142 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 5 of 52 Unusual Event - SU 1.1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

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 buses.

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

PNPS Basis:

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure S-1.

The inability to provide emergency AC power from any offsite power source poses a potential degradation of reactor plant safety. If onsite power capability becomes degraded, significant losses of vital equipment operability may occur. 4160V AC buses A5 and A6 are the emergency buses. These electrical buses provide power to vital equipment such as RHR, Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses 81 and 82.

Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators. Due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers.

The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If neither emergency bus is energized by an offsite source within 15 minutes, an Unusual Event is declared under this EAL.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 143 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 6 of 52 Unusual Event - SU1 .1 PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31, "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator"
8. PNPS 2.4.144, "Degraded Voltage"
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line"
10. PNPS 2.4.A.5, "Loss of Electrical Bus A5"
11. PNPS 2.4.A.6, "Loss of Electrical Bus A6"

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 144 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 7 of 52 Unusual Event - SU1 .1 Figure 5-1 Emergency Bus Composite Diagram 24 KV 24KV 345 KV

~ UAT "A"

sigt ~SOT

~

A-8 .

"B" SUT~

801 EOG EOG

)600

)sos l'")sog l'")so1 IJS04 I

PNPS

)6os 1)609 1/601 )604 I I I REV. 001 I I I I I I 04/15/02 JEH 4160BUSA-5 4160 BUS A-6 I)so2 I'")so3 *1'")so6 I'")so7 I)sos I)602 I)603 I)606 I)607 I)608

~ ~

"Tl Gi I TO BUS B-1 PNPS I TO BUS B-2 C

o EMERGENCY BUS m

--' COMPOSITE

a-===- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 145 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 8 of 52 Alert - SA 1 .1 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to emergency buses EAL:

SA1.1 Alert AC power capability to emergency buses A5 and A6 reduced to a single power source

  • (Table S-3) for~ 15 min. such that any additional single failure would result in loss of all AC power to emergency buses (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table S-3 AC Power Sources Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 146 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 9 of 52 Alert- SA1 .1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

The condition indicated by this EAL 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 buses. Another related condition could be the loss of all offsite power and loss of onsite emergency generators with only one train of emergency buses being backfed from the unit main generator, or the loss of onsite emergency generators with only one train of emergency buses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site

  • Area Emergency in accordance with EAL SS1 .1.

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

PNPS Basis:

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure S-1.

The inability to provide emergency AC power from any offsite power source poses a potential degradation of reactor plant safety. If onsite power capability becomes degraded, significant losses of vital equipment operability may occur. 4160V AC buses A5 and A6 are the emergency buses. These electrical buses provide power to vital equipment such as RHR, Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses B1 and 82.

Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators. Due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers .

~Entergy EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 147 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 10 of 52 Alert - SA 1.1 PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31, "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator
8. PNPS 2.4.144, "Degraded Voltage"
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line"
10. PNPS 2.4.A.5, "Loss of Electrical Bus AS"
11. PNPS 2.4.A.6, "Loss of Electrical Bus A6"

~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 148 Revision 9 of 351

  • Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 11 of 52 Alert- SA1.1 Figure 5-1 Emergency Bus Composite Diagram 24 KV 24KV 345 KV

~ UAT IIAII siii 801

~SDT

~

"B" SUT~

EOG EOG

)600 I

PNPS REV. 001 04/15/02 I JEH 1...)507 1)508 )602 606 J607 )608

~ ~

"Tl G)

C I TO BUS B-1 PNPS I TO BUS B-2

o EMERGENCY BUS m

..... COMPOSITE

a-===- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 149 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 12 of 52 Site Area Emergency - SS 1.1 Category: S - System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

551.1 Site Area Emergency Loss of all offsite and all onsite AC power (Table S-3) to emergency buses A5 and A6 for

~ 15 min. (Note 3)

  • Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table 5-3* AC Power Sources Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

- SBO DG

--=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 150 Revision 9 of 351

  • Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 13 of 52 Site Area Emergency - SS1 .1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Loss of all AC power to emergency buses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, and the Ultimate Heat Sink.

Prolonged loss of all AC power to emergency buses will lead to loss of fuel clad, RCS, and Primary 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.

Escalation to General Emergency is via fission product barrier degradation EALs or EAL SG1 .1.

PNPS Basis:

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure S-1.

The inability to provide emergency AC power from any offsite power source poses a potential degradation of reactor plant safety. If onsite power capability becomes degraded, significant losses of vital equipment operability may occur. 4160V AC buses A5 and A6 are the emergency buses. These electrical buses provide power to vital equipment such as RHR, Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses B1 and B2.

Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators. Due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers.

This EAL is the hot condition equivalent of the cold condition loss of all AC power EAL CA 1.1 .

When in Cold Shutdown, Refuel, or Defueled mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the emergency buses relative to that existing when in hot conditions .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 151 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 14 of 52 Site Area Emergency - SS 1.1 PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31, "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator"
8. PNPS 2.4.144, "Degraded Voltage"
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line"
10. PNPS 2.4.A.5, "Loss of Electrical Bus A5"
11. PNPS 2.4.A.6, "Loss of Electrical Bus A6"

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 152 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 15 of 52 Site Area Emergency - SS1 .1 Figure S-1 Emergency Bus Composite Diagram 24KV 24KV 345 KV

~ UAT ~SDT SUT~

~

"8" EDG

)600 I

PNPS REV. 001 I

04/15/02 416'0 SUS A-5-:' .... *, JEH I)502 I)503 I)506 1,507 I 1)508 )602 606 )607 )608

~ ~

I TO BUS B-1 PNPS I TO BUS B-2 EMERGENCY BUS COMPOSITE

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 153 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 16 of 52 General Emergency - SG1 .1 Category: S -System Malfunction Subcategory: 1 - Loss of AC Power Initiating Condition: Prolonged loss of all offsite power and prolonged loss of all onsite AC power to emergency buses EAL:

SG1.1 General Emergency Loss of all offsite and all onsite AC power (Table S-3) to emergency buses A5 and A6 AND EITHER:

Restoration of at least one emergency bus in < 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is not likely OR RPV level cannot be restored and maintained> -125 in. or cannot be determined Table 5-3 AC Power Sources Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

- SBO DG

= -- EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 154 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 17 of 52 General Emergency- SG1 .1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Loss of all AC power to emergency buses compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, and the Ultimate Heat Sink.

Prolonged loss of all AC power to emergency buses will lead to loss of fuel clad, RCS, and containment, thus warranting declaration of a General Emergency.

This EAL is specified to assure that, in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate based on a reasonable assessment of the event trajectory.

  • The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded.

PNPS Basis:

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure S-1.

The inability to provide emergency AC power from any offsite power source poses a potential degradation of reactor plant safety. Should onsite power capability become degraded, significant losses of vital equipment operability may occur. 4160V AC buses AS and A6 are the emergency buses. These electrical buses provide power to vital equipment such as RHR, Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses 81 and 82.

Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators. Due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers .

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 155 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 18 of 52 General Emergency - SG1 .1 Indication of continuing core cooling degradation is manifested by an RPV level instrument reading of< -125 in. (RPV level is below TAF) (ref. 12). When RPV level is at or above TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the

  • EOPs specify alternate, more extreme RPV level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the fuel clad barrier.

When RPV level cannot be determined, EOPs require entry to EOP-16, "RPV Flooding". RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 13). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling

  • must be attempted. The instructions in EOP-16 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram-failure events). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.

PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31, "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator
8. PNPS 2.4.144, "Degraded Voltage"
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line"
10. PNPS 2.4.A.5, "Loss of Electrical Bus A5"
11. PNPS 2.4.A.6, "Loss of Electrical Bus A6"
12. EOP-1, "RPV Control"
13. EOP-16, "RPV Flooding"

-::::=- ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 156 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 19 of 52 General Emergency- SG1 .1 Figure 5-1 Emergency Bus Composite Diagram 24 KV 24 KV 345 KV

~ UAT ~SDT SUT~

~

"A" A-8 11911 EOG EOG

)600 I

PNPS

)605 L, IJ J 609 601 604 REV. 001 I

04/15/02 JEH 4160 BUS A-5

  • 4160 BUS A-6

)602 603 )606 )607 )608

~

'TI Gi C

PNPS I TO BUS B-2

o EMERGENCY BUS m

COMPOSITE

  • ~Entergy- PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 157 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 20 of 52 Unusual Event - SU2.1 Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Inadvertent criticality EAL:

SU2.1 Unusual Event Unplanned sustained positive period observed on nuclear instrumentation Mode Applicability:

2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

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

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

PNPS Basis:

Period meters Nl-750-4A, C, B, and Don Panel C905 identify this condition as well as annunciator "SRM PERIOD" (C905L-G9). Amber lights on Panel C905 illuminate when its SRM channel period is less than 20 seconds (seal in) (ref. 1, 2).

PNPS Basis Reference(s):

1. PNPS 2.2.64, "Source Range Monitoring System"
2. PNPS 3.M.2-5.1, "Source Range Monitor Calibration Instruction"

~--------------- -----

NON-QUALITY A

PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 158 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 21 of 52 Alert - SA2.1 Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic scram fails to shut down the reactor and the manual actions taken from the reactor control console are successful in shutting down the reactor.

EAL:

SA2.1 Alert An automatic scram failed to shut down the reactor (reactor power< 3%)

Manual actions taken at the reactor control console successfully shut down the reactor as indicated by reactor power< 3% (APRM downscale) (Note 6)

Note 6: Manual scram actions taken at the reactor control console are the following:

  • Reactor Scram push buttons
  • Reactor Mode switch in SHUTDOWN.
  • ATWS-ARI push buttons.

Mode Applicability:

1 - Run, 2 - Startup NEI 99-01 Basis:

Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.

a-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 159 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 22 of 52 Alert - SA2.1 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 shut down the plant.

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

PNPS Basis:

A reactor scram is automatically initiated by the Reactor Protection System (RPS) when certain continuously monitored parameters exceed predetermined setpoints. In Hot condition operating modes, a reactor scram may be the result of manual or automatic action in response to any of the following parameters (ref. 1):

  • IRM High-High Flux
  • APRM High Flux (15%)
  • Rx Pressure High
  • Drywell Pressure High
  • Rx Water Level Low

a.:.

-===- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 160 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 23 of 52 Alert - SA2.1

  • Turbine Stop Valve Closure Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion to bring the reactor power below the APRM downscale setpoint of 3% (ref. 2).

This EAL indicates a failure of the automatic RPS scram function to rapidly insert a sufficient

  • number of control rods to achieve reactor shutdown. The significance, therefore, is that a potential degradation of a safety system exists because a front line automatic protection system did not function in response to a plant transient. Thus, plant safety has been compromised.

Following any automatic RPS scram signal, PNPS 2.1.6 (ref. 3) prescribes insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved. Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Alert.

For the purpose of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console, Panel C905 (ref. 3, 4):

  • Reactor Scram push buttons
  • Reactor Mode switch in SHUTDOWN.
  • ATWS-ARI push buttons.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . ADMINISTRATIVE EMERGENCY PLAN PROCEDURE PROCEDURES REFERENCE USE Page 161 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 24 of 52 Alert - SA2.1 Reactor shutdown achieved by use of the EOP satellite procedures referenced in EOP-2 do not constitute a successful manual scram (ref. 5).

This EAL is not applicable if a manual scram is initiated and no RPS setpoints are exceeded.

Taking the mode switch to SHUTDOWN position is a manual scram action. When the mode switch is taken out of the RUN position, however, the nuclear instrumentation scram setpoint is lowered. If reactor power remains above the lowered setpoint, an automatic scram is initiated.

In the event that the operator identifies a reactor scram is imminent and initiates a successful manual reactor scram before the automatic scram setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions fail to reduce reactor power below 3% (ref. 2), the event escalates to the Site Area Emergency under EAL SS2.1 .

If, by procedure, operator actions include the initiation of an immediate manual scram following receipt of an automatic scram signal and there are no clear indications that the automatic scram failed ( such as a time delay following indications that a scram setpoint was exceeded), it may be difficult to determine whether the reactor was shut down because of automatic scram or manual actions. If a subsequent review of the scram actuation indications reveals that the automatic scram did not cause the reactor to be shut down, consideration should be given to evaluating the fuel for potential damage and the reporting requirements of 50.72 should be considered for the transient event.

PNPS Basis Reference(s):

1. PNPS 2.2.79, "Reactor Protection System"
2. EOP-1, "RPV Control"
3. PNPS 2.1.6, "Reactor Scram"
4. PNPS 2.2.126, "Anticipated Transient Without Scram (A TWS)"
5. EOP-2, "RPV Control, Failure-to-Scram"

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 162 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 25 of 52 Site Area Emergency - SS2.1 Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Automatic scram fails to shut down the reactor and manual actions taken from the reactor control console are not successful in shutting down the reactor EAL:

552.1 Site Area Emergency An automatic scram failed to shut down the reactor (reactor power< 3%)

Manual actions taken at the reactor control console do not shut down the reactor as indicated by reactor power ~ 3% (Note 6)

Note 6: Manual scram actions taken at the reactor control console are the following:

  • Reactor Scram push buttons
  • Reactor Mode switch in SHUTDOWN.
  • ATWS-ARI push buttons.

Mode Applicability:

1 - Run, 2 - Startup

  • -===-Entergy

-. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

. ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 163 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 26 of 52 Site Area Emergency - SS2.1 NEI 99-01 Basis:

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful. A Site Area Emergency is warranted because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS.

Manual scram actions taken at the reactor control console are any set of actions by the reactor operator(s) which causes or should cause control rods to be rapidly inserted into the core and shuts down the reactor.

Manual scram actions are not considered successful if action away from the reactor control console is required to scram (trip) the reactor. This EAL is still applicable even if actions taken away from the reactor control console are successful in shutting down the reactor because the design limits of the fuel may have been exceeded or because of the gross failure of the Reactor Protection System to shut down the plant.

Escalation of this event to a General Emergency would be due to a prolonged condition leading to an extreme challenge to either core cooling or heat removal.

PNPS Basis:

This EAL addresses any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed. For the purpose of emergency classification at the Site Area Emergency level, successful manual scram actions are those which can be quickly performed from the reactor control console, Panel C905 (ref. 1, 2):

  • Reactor Scram push buttons
  • Reactor Mode switch in SHUTDOWN.
  • ATWS-ARI push buttons.

Reactor shutdown achieved by use of the EOP satellite procedures referenced in EOP-2 do not constitute a successful manual scram (ref. 3) .

-:: :.-:- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 164 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 27 of 52 Site Area Emergency - SS2.1 The APRM downscale trip setpoint is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine whether reactor power is greater than 3% power (ref. 4).

Escalation of this event to a General Emergency would be under EAL SG2.1 or Emergency Director judgment.

PNPS Basis Reference(s):

1. PNPS 2.1.6, "Reactor Scram"
2. PNPS 2.2.126, "Anticipated Transient Without Scram (A TWS)"
3. EOP-2, "RPV Control, Failure-to-Scram"
4. EOP-1, "RPV Control"

,. PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page. 165 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 28 of 52 General Emergency - SG2.1 Category: S - System Malfunction Subcategory: 2 - ATWS/Criticality Initiating Condition: Failure of the RPS to complete an automatic scram and manual scram was not successful and there is indication of an extreme challenge to the ability to cool the core EAL:

SG2.1 General Emergency An automatic scram failed to shut down the reactor (reactor power< 3%)

  • AND All manual actions do not shut down the reactor as indicated by reactor power AND EITHER: of the following exists or has occurred due to continued power

~ 3%

generation:

RPV level cannot be restored and maintained> -150 in. or cannot be determined OR Torus water temperature and RPV pressure cannot be maintained below Heat Capacity Temperature Limit (EOP-11 Figure 2)

Mode Applicability:

1 - Run, 2 - Startup

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 166 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 29 of 52 General Emergency - SG2.1 NEI 99-01 Basis:

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed and efforts to bring the reactor subcritical are unsuccessful.

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.

PNPS Basis:

This EAL addresses the following:

  • Any automatic reactor scram signal followed by failure of the automatic scram and all subsequent manual scrams to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL SS2.1 ); and
  • Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

In addition to manual scram methods at the reactor control console (Panel C905), a reactor shutdown achieved by use of the EOP support procedures referenced in EOP-2 is also credited as a successful manual scram provided that reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist (ref. 1).

The APRM downscale trip setpoint is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, torus temperature trend) can be used to determine whether reactor power is greater than 2% power (ref. 2).

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 167 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 30 of 52 General Emergency - SG2.1 The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the fuel clad and RCS barriers.

Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above -150 in. (or cannot be determined). -150 in. is the Minimum Steam Cooling RPV Water Level (MSCRWL). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel.

RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence. When RPV level cannot be determined under ATWS conditions, EOPs require entry to EOP-26, "RPV Flooding, Failure-to-Scram"

  • (ref. 3). RPV water level indication provides the primary means of knowing whether adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-26 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures.

The Heat Capacity Temperature Limit (HCTL) is the highest torus temperature from which emergency RPV depressurization will not raise torus pressure above the Primary Containment Pressure Limit (PCPL) 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 torus level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. This threshold is met when EOP-3, "Primary Containment Control", Step TT-10 is reached (ref. 4). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature .

a-===- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 168 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 31 of 52.

General Emergency - SG2.1 It is recognized that actions will be taken in the plant Emergency Operating Procedures (EOPs) in response to the failure to scram. The conditions addressed in the EAL thresholds

("RPV level cannot be restored and maintained > -150 in ... " and "torus water temperature and RPV pressure cannot be maintained below the Heat Capacity Temperature Limit ... ")

correspond to decisions made in the EOPs. If at any time it is determined that RPV level cannot be restored and maintained > -150 in. or torus water temperature and RPV pressure cannot be maintained below the Heat Capacity Temperature Limit, a General Emergency must be declared.

PNPS Basis Reference(s):

1. EOP-2, "RPV Control, Failure-to-Scram"
2. EOP-1, "RPV Control" 3.

4.

5.

6.

EOP-26, "RPV Flooding, Failure-to-Scram" EOP-3, "Primary Containment Control" PNPS 2.1.6, "Reactor Scram" PNPS 2.2.126, "Anticipated Transient Without Scram (A TWS)" *

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 169 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 32 of 52 Unusual Event - SU3.1 Category: S - System Malfunction Subcategory: 3 - Inability to Reach Shutdown Conditions Initiating Condition: Inability to reach required shutdown within Technical Specifications limits EAL:

SU3.1 Unusual Event Plant is not brought to required operating mode within Technical Specifications LCO action statement time

  • Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Limiting Conditions for Operation (LCOs) require the plant to be brought to a required operating mode when the Technical Specifications 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 4-hour report under 10CFR50.72(b) Nonemergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate UE 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 UE is based on the time at which the LCD-specified action statement time period elapses under the PNPS Technical Specifications and is not related to how long a condition may have existed.

PNPS Basis:

None

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 170 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 33 of 52 Unusual Event - SU3.1 PNPS Basis Reference(s):

1. PNPS Technical Specifications

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 171 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 34 of 52 Unusual Event - SU4.1 Category: S - System Malfunction Subcategory: 4 - Instrumentation/Communications Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room for 15 minutes or longer EAL:

SU4.1 Unusual Event Unplanned loss of> approximately 75% of annunciators or indicators associated with safety systems on MCR Panels C903, C904, C905, C1, C3, C170, and C171 for~ 15 min.

  • (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

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

Recognition of the availability of computer-based indication equipment is considered (e.g.,

Process Computer, EPIC, SPDS, etc.).

Loss of annunciators or indicators excludes scheduled maintenance and testing activities .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

$' Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 172 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 35 of 52 Unusual Event - SU4.1 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 Specifications. The initiation of a Technical Specifications imposed plant shutdown related to the instrument loss will be reported via 10CFR50. 72. If the shutdown is not in compliance with

  • the Technical Specifications action, the UE is based on EAL SU3.1.

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

This UE 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.

PNPS Basis:

The availability of computer-based monitoring capability (i.e., Process Computer, EPIC, SPDS) is a factor at the Alert classification level but not a factor at the Unusual Event emergency classification level. Safety system annunciation and indication considered in this EAL are found on MGR Panels C903, C904, C905, C1, C3, C170, and C171. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 1, 2).

PNPS Basis Reference(s):

1. PNPS 2.3.1, "General Action for Alarm Response and Annunciator Control"
2. PNPS 2.4.155, "Loss of Annunciator System"

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 173 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 36 of 52 Unusual Event - SU4.2 Category: S - System Malfunction Subcategory: 4 - Instrumentation/Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

SU4.2 Unusual Event Loss of all Table S-2 onsite (internal) communication methods affecting the ability to perform routine operations OR Loss of all Table S-2 offsite (external) communication methods affecting the ability to perform offsite notifications Table S-2 Communicaticms Systems System Onsite Offsite (internal) (external)

Plant Telephone System (CENTREX) X X Wireless Telephone System X X Pilgrim Station Radio System X Plant Gaitronics System X Alternate Shutdown Communication X NRC-ENS Telephone, Direct Line X Satellite phones X

a PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

  • --=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 174 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 37 of 52 Unusual Event - SU4.2 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

The purpose of this EAL is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate issues with 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 nonroutine radio transmissions, individuals being sent to offsite locations, etc.) are being used to make communications possible.

Onsite/offsite communications include one or more of the systems listed in Table S-2 (ref. 1, 2, 3, 4, 5). A description of the capabilities of each system is given in Section 4.0 of PNPS 2.2.17, "Communications Systems" (ref. 2).

This EAL is the hot condition equivalent of the cold condition EAL CU4.1.

PNPS Basis Reference(s):

1. FSAR Section 10.15
2. PNPS 2.2.17, "Communications Systems"
3. PNPS 2.4.57, "Loss of Public-Address System"
4. PNPS 8.A.13, "Plant Emergency Alarms and Radio Test"
5. EP-AD-413, "Emergency Communications Test"

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 175 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 38 of 52 Alert - SA4.1 Category: S - System Malfunction Subcategory: 4 - Instrumentation/Communications Initiating Condition: Unplanned loss of safety system annunciation or indication in the Control Room with EITHER 1) a significant transient in progress or

2) compensatory indicators unavailable EAL:

SA4.1 Alert Unplanned loss of> approximately 75% of annunciators or indicators associated with safety systems on MCR Panels C903, C904, C905, C1, C3, C170, and C171 for 2: 15 min .

(Note 3)

AND EITHER:

Any significant transient is in progress, Table S-1 OR Compensatory indications are unavailable Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table 5;.1 Significanf Transients Reactor scram Runback > 25% thermal power Electrical load rejection> 25% full electrical load ECCS injection Thermal power oscillations> 10%

NON-QUALITY

~ PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 176 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 39 of 52 Alert - SA4.1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a significant transient.

"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 also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specifications. The initiation of a Technical Specifications imposed plant shutdown related to the instrument loss will be reported via 10CFRS0.72. If the shutdown is not in compliance with the Technical Specifications action, the UE is based on EAL SU3.1, "Inability to Reach Required Shutdown Within Technical Specifications Limits."

"Compensatory indications" in this context include computer-based information such as SPDS, Process Computer, EPIC, etc. 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 .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 177 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 40 of 52 Alert - SA4.1 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.

PNPS Basis:

The Process Computer, EPIC, or SPDS serves as a redundant compensatory indicator which may be utilized in lieu of normal Control Room indicators. Safety system annunciation and indication considered in this EAL are found on MCR Panels C903, C904, C905, C1, C3, C170, and C171. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 1, 2).

Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater.

PNPS Basis Reference(s):

1. PNPS 2.3.1, "General Action for Alarm Response and Annunciator Control"
2. PNPS 2.4.155, "Loss of Annunciator System"

e-:: : : - Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAf'J ADMINISTRATIVE, PROCEDURE PROCEDURES REFERENCE USE Page 178 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 41 of 52 Site Area Emergency - SS4.1 Category: S - System Malfunction Subcategory: 4 - Instrumentation/Communications Initiating Condition: Inability to monitor a significant transient in progress EAL:

554.1 Site Area Emergency Loss of> approximately 75% of the annunciators or indicators associated with safety systems on MCR Panels C903, C904, C905, C1, C3, C170, and C171 for :2: 15 min.

(Note 3)

AND Any significant transient is in progress, Table S-1 AND Compensatory nonalarming indications are unavailable Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Tallie 5;_1 'Si_~nifica.nt Transients Reactor scram Runback > 25% thermal power Electrical load rejection> 25% full electrical load ECCS injection Thermal power oscillations> 10%

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 179 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 42 of 52 Site Area Emergency - SS4.1 Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is intended to recognize the threat to plant safety associated with the complete loss of capability of the Control Room staff to monitor plant response to a significant transient.

"Planned" and "unplanned" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

Quantification is arbitrary; however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide increased monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specifications. The initiation of a Technical Specifications imposed plant shutdown related to the instrument loss will be reported via 10CFR50.72. If the shutdown is not in compliance with the Technical Specifications action, the UE is based on EAL SU3.1, "Inability to Reach Required Shutdown Within Technical Specifications 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 .

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 180 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 43 of 52 Site Area Emergency - SS4.1 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.

"Compensatory indications" in this context include computer-based information such as SPDS, Process Computer, EPIC, etc. 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.

PNPS Basis:

The availability of computer-based monitoring capability (i.e., Process Computer, EPIC, SPDS) is a factor at the Site Area Emergency classification level because they are compensatory nonalarming indications. These serve as a redundant compensatory indicator

  • which may be utilized in lieu of normal Control Room indicators. Safety system annunciation and indication considered in this EAL are found on MGR Panels C903, C904, C905, C1, C3, C170, and C171. The other annunciators and indicators are important to plant operation but are not important to safety (ref. 1, 2).

The ability to monitor EOP parameters is ultimately necessary for protection of the health and safety of the public. These parameters include those used to determine such functions as the ability to shut down the reactor, to maintain the core cooled and in a coolable geometry, to remove heat from the core, to maintain the reactor coolant system intact, and to maintain containment integrity.

Significant transients are listed in Table S-1 and include response to automatic or manually initiated functions such as trips, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10% or greater.

Due to the limited number of safety systems in operation during cold shutdown, refuel, and defueled modes, this EAL is not applicable during these modes of operation.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 181 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 44 of 52 Site Area Emergency - SS4.1 PNPS Basis Reference(s):

1. PNPS 2.3.1, "General Action For Alarm Response And Annunciator Control"
2. PNPS 2.4.155, "Loss of Annunciator System"
3. PNPS Plant-Specific and Severe Accident Management Guidelines

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 182 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 45 of 52 Unusual Event - SU5.1 Category: S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL:

SU5.1 Unusual Event Air Ejector Offgas Radiation Monitors RM-1705-3A and B (Panel C910) reading> Hi-Hi alarm for> 13 min.

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.

Escalation of this EAL to the Alert level is via the fission product barriers.

This threshold addresses PNPS radiation monitor readings that provide indication of a degradation of fuel clad integrity.

PNPS Basis:

The Air Ejector Offgas Radiation Monitors 1705-3A and B continuously monitor air ejector offgas radiation levels, initiate an alarm if gamma radiation levels exceed short-term maximum release rates, and initiate auto-closure of offgas isolation valves if these limits are exceeded for longer than 13 minutes. The monitors serve to detect fuel damage since noble gases leaking from fuel elements pass through the monitoring instruments prior to release.

The high-high trip and alarm are set to preclude exceeding the Technical Specifications offgas release limit of 500,000 µCi/sec noble gas (referenced to a 30-minute hold-up) (ref. 1).

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 183 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 46 of 52 Unusual Event - SU5.1 The timer associated with the protective function of the offgas radiation monitors allows the operator time to clear the initiating signal through power reduction. The inability to clear the initiating signal within the 13 minutes indicates a significant fuel cladding integrity problem which, in turn, is indicative of abnormal core conditions and a potential for increased radiological hazards in plant. An air ejector offgas radiation monitor hi-hi alarm which does not clear within 13 minutes therefore warrants declaration of an Unusual Event.

In the Hot modes, a steam source is available from which noncondensable gases can be separated for processing by the offgas system. The cold shutdown, refuel, and defueled modes do not afford a transfer mechanism from which the Air Ejector Offgas radiation monitors can draw a valid sample. The radiation monitors lose a valid sample source when the air ejectors are not in service (ref. 1).

PNPS Basis Reference(s):

1. PNPS ODCM Section 8.3 Steam Jet Air Ejector Monitor

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 184 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 47 of 52 Unusual Event - SU5.2 Category: S - System Malfunction Subcategory: 5 - Fuel Clad Degradation Initiating Condition: Fuel clad degradation EAL:

SU5.2 Unusual Event Reactor coolant system sample activity > 20 µCi/ml total iodine Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included because it is a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant.

Escalation of this EAL to the Alert level is via the fission product barriers.

This threshold addresses coolant samples exceeding coolant Technical Specifications for transient iodine spiking limits.

PNPS Basis:

Elevated reactor coolant activity represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This EAL addresses reactor coolant samples exceeding Technical Specifications LCOs 3.6.B which are applicable in Hot operating modes (ref. 1). A radioactivity concentration of 20 µCi/ml total iodine can be reached if there is fuel cladding failure or if there is a failure or a prolonged shutdown of the cleanup demineralizer.

PNPS Basis Reference(s):

1. PNPS Technical Specifications 3.6.B

e~Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 185 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 48 of 52 Unusual Event - SU6.1 Category: S - System Malfunction Subcategory: 6 - RCS Leakage Initiating Condition: RCS leakage EAL:

SU6.1 Unusual Event Unidentified or pressure boundary leakage > 10 gpm OR

  • Identified leakage > 25 gpm (Note 7)

Note 7: See Table F-1, Fission Product Barrier Matrix, for possible escalation above the Unusual Event due to RCS Leakage Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

This EAL is included as a UE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 GPM value for the unidentified or pressure boundary leakage was selected as it is observable with normal Control Room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances).

Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close according to design should be considered applicable to this EAL if the relief valve cannot be isolated.

The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this EAL to the Alert level is via fission product barrier degradation EALs .

-=-Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 186 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 49 of 52 Unusual Event - SU6.1 PNPS Basis:

The following alarms may be indicative of a reactor coolant leak inside containment (ref. 1):

  • "DRYWELL EQPT DRAIN SUMP DISCH HIGH TOTAL FLOW" (C20C-A3)
  • "DRYWELL FLOOR DRAIN SUMP DISCH HIGH TOTAL FLOW" (C20C-B3)
  • Any Drywell cooler leaking alarm on Panel C7L Identified leakage is:

Reactor coolant leakage into Drywell collection systems such as pump seal or valve packing leaks that is captured and conducted to a sump or collecting tank; or Reactor coolant leakage into the Drywell atmosphere from sources which are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be pressure boundary leakage.

Pressure boundary leakage is leakage through a nonisolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

Unidentified leakage is all reactor coolant leakage which is not identified leakage (ref. 2).

PNPS Basis Reference(s):

1. PNPS 2.2.77, "Drywe/1 Leak Detection Systems"
2. PNPS Technical Specifications 3.6.C

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 187 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY S, SYSTEM MALFUNCTION Sheet 50 of 52 Site Area Emergency - SS7 .1 Category: S - System Malfunction Subcategory: 7 - Loss of DC Power Initiating Condition: Loss of all essential DC power for 15 minutes or longer EAL:

SS7.1 Site Area Emergency

< 105V DC bus voltage indications on all essential 125V DC buses (Panels D16 and D17) for~ 15 min .. (Note 3)

  • Note 3: The Emergency Director should.not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

Loss of all DC power compromises ability to monitor and control plant safety functions.

Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system.

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

Escalation to a General Emergency would occur by EALs in Category A or Category F .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 188 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 51 of 52 Site Area Emergency - SS7 .1 PNPS Basis:

The essential 125V DC power distribution is illustrated in Figure S-2.

The loss of required essential 125V DC power poses a significant threat to decay heat removal capability and reactor plant safety. Operability of safety-related equipment and safety system protective functions are severely degraded. The 125V DC system provides power to CSCS initiation logics, controllers, and indications. The 125V DC system also provides control power and tripping power for high voltage AC protective devices. If not restored within a short period of time, significant system and equipment failures may be imminent depending upon plant conditions at the time of the loss.

Annunciators "A 125V DC UNDERVOLTAGE" (C3RC-A7) and "B 125V DC UNDERVOLTAGE" (C3RC-B7) alarm at 124V DC (decreasing) and signal loss of Panel 016

  • and 017, respectively.

This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU6.1.

PNPS Basis Reference(s):

1. FSAR Figure 8.6-1
2. PNPS 2.2.14, "125V DC Battery Systems"
3. PNPS 5.3.11, "Loss of Essential DC Bus 016 or 04 and 036"
4. PNPS 5.3.12, "Loss of Essential DC Bus 017 or 05 and 037"
5. PNPS ARP C3RC-A7
6. PNPS ARP C3RC-B7

-=-

-===- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE NON-QUALITY RELATED PROCEDURE

  • EP-AD-601 Revision 9 PROCEDURES REFERENCE USE Page 189 of 351 Emergency Action Level Technical Bases Document I ATTACHMENT 9.3 CATEGORY 5, SYSTEM MALFUNCTION Sheet 52 of 52 Site Area Emergency - SS7 .1 Figure S-2 Essential 125V DC Power Distribution 480V MCC 8-15 125VDC 4BOV MCC B-10 125VDC 480V MCC B-14 CONTROi. CONTROL c1' BATTERY A BATIERYB s2-1s13 62-1021
_!., -:I, 0-1 D-2 C FUSE ( FUSE
  • 125 voe CHARGER 125 VDC,BACKUP BATTERY A BATIERY CHARGER l

200a C D-11 0-2'.I 200a C 0-14 D-12 j' 0-16 _r--;;

(72-162 ~ ~

,l,. ,(

~

]'"

f l2-16B ~2-16A

(?2-171 f

AUTO TRANSFER SWITCH Y-10 MA!NI PANEL PAIEL PANEL PANEL MAINT SPARE D-4 D-7 D-8 D-5 SPARE EMERGENCY LIGHTING PANEL SOLENOID OPERATED PANEL PANEL DISTRIBUTION PANEL D-8 VALVE PANEL D-37 D-'.:-5 D-H<

25L

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 190 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 1 of 8 Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 212°F); EALs in this category are applicable only in one or more hot operating modes including Startup/Hot Standby.

EALs in this category represent threats to the defense-in-depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

A. Fuel Clad (FC): The fuel clad barrier consists of the zircalloy fuel bundle tubes that contain the fuel pellets.

B. Reactor Coolant System (RCS): 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.

C. Primary Containment (PC): The Primary Containment barrier includes the drywell, the wetwell (torus), their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 9.6). "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials. "Potential Loss" means integrity of the barrier is threatened and could be lost if conditions continue to degrade. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:

Unusual Event:

Any Joss or any potential Joss of Primary Containment Any Joss or any potential Joss of either fuel clad or RCS

~ RELATED EP-AD-601 Revision 9

-::::::-Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 191 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 2 of 8 Site Area Emergency:

Loss or potential loss of any two barriers General Emergency:

Loss of any two barriers and loss or potential loss of third barrier Note 1: The logic used for these initiating conditions reflects the following considerations:

  • The fuel clad barrier and the RCS barrier are weighted more heavily than the Primary Containment barrier. UE EALs associated with RCS and fuel clad barriers are addressed under system malfunction EALs.

At the Site Area Emergency level, there must be some ability to dynamically assess how far present conditions are from the threshold for a General Emergency. For example, if fuel clad and RCS barrier "loss" EALs existed, that, in addition to offsite dose assessments, would require continual assessments of radioactive inventory and containment integrity. Alternatively, if both fuel clad and RCS barrier "potential loss" EALs existed, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency.

  • The ability to escalate to higher emergency classes as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.
  • The Primary Containment barrier should not be declared lost or potentially lost based on exceeding Technical Specifications 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 Determine which combinations of the three barriers are lost or have a potential loss and use FU1 .1, FA1 .1, FS1 .1, and FG1 .1 to classify the event. Also, multiple events could occur which result in the conclusion that exceeding the loss or potential loss thresholds is imminent. In this imminent loss situation use judgment and classify as if the thresholds are exceeded .

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 192 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 3 of 8 Unusual Event - FU1 .1 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of Primary Containment EAL:

FU1.1 Unusual Event Any loss or any potential loss of Primary Containment (Table F-1, Attachment 9.6)

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None PNPS Basis:

Fuel clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 9.6) lists the fission product barrier thresholds, bases, and references.

Fuel clad and RCS barriers are weighted more heavily than the Primary Containment barrier.

Unlike the fuel clad and RCS barriers, the loss of either of which results in an Alert (EAL FA 1.1 ), loss of the Primary Containment barrier in and of itself does not result in the relocation of radioactive materials or the potential for degradation of core cooling capability.

However, loss or potential loss of the Primary Containment barrier in combination with the loss or potential loss of either the fuel clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1.

PNPS Basis Reference(s):

None

a-===-Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 193 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 4 of 8

-~==-~~------=""-==~--~==~=---==-

Alert - FA1 .1 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either fuel clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of either fuel clad or RCS (Table F-1, Attachment 9.6)

  • Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None PNPS Basis:

Fuel clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 9.6) lists the fission product barrier thresholds, bases, and references.

At the Alert classification level, fuel clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Primary Containment barrier, loss or potential loss of either the fuel clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Primary Containment barrier in combination with loss or potential loss of either fuel clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1 .1.

PNPS Basis Reference(s):

None

-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 194 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 5 of 8 Site Area Emergency - FS1 .1 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1, Attachment 9.6)

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None PNPS Basis:

Fuel clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 9.6) lists the fission product barrier thresholds, bases, and references.

At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions:

  • One barrier loss and a second barrier loss (i.e., loss - loss)
  • One barrier loss and a second barrier potential loss (i.e., loss - potential loss)
  • One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss)

a-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 195 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 6 of 8 Site Area Emergency - FS1 .1 At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of fuel clad and RCS barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Primary Containment integrity in anticipation of reaching a General Emergency classification.

Alternatively, if both fuel clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent.

PNPS Basis Reference(s):

None

~Entergy.

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 196 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 7 of 8 General Emergency - FG1 .1 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1, Attachment 9.6)

Mode Applicability:

1 - Run, 2 - Startup, 3 - Hot Shutdown NEI 99-01 Basis:

None PNPS Basis:

Fuel clad, RCS, and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 9.6) lists the fission product barrier thresholds, bases, and references.

At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions:

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 197 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.4 CATEGORY F, FISSION PRODUCT BARRIER DEGRADATION Sheet 8 of 8 General Emergency - FG1 .1 PNPS Basis Reference(s):

None

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

~Entergy. EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 198 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 1 of 64 Category C - Cold Shutdown/Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature~ 212°F); EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out of service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refuel, D - Defueled).

The events of this category pertain to the following subcategories:

1. Loss of AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for 4160V emergency buses.
2. RPV Level RPV water level is a measure of inventory available to ensure adequate core cooling and, therefore, maintain fuel clad integrity. The RPV provides a volume for the coolant that covers the reactor core. The RPV and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail.
3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

~Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 199 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 2 of 64

4. Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
5. Inadvertent Criticality Inadvertent criticalities pose potential personnel safety hazards as well being indicative of losses of reactivity control.
6. Loss of DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of vital 125 volt DC power sources .

~Entergy- EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 200 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 3 of 64 Unusual Event - CU1 .1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: AC power capability to emergency buses reduced to a single power source for 15 minutes or longer such that any additional single failure would result in loss of all AC power to emergency buses EAL:

CU1.1 Unusual Event AC power capability to emergency buses AS and A6 reduced to a single power source (Table C-4) for;::: 15 min. such that any additional single failure would result in loss of all AC power to emergency buses (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time .

  • ' - "d

.*

  • Table*c:4 'ACP:owerSources* *.

.. , * < '* ' " ~

  • "*-- *, ,-., .. v . ..;,*,

Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

- SBO DG

NON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 201 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 4 of 64 Unusual Event - CU 1.1 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

The condition indicated by this EAL 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 buses. The subsequent loss of this single power source would escalate the event to an Alert in accordance with EAL CA 1.1.

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

PNPS Basis :

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure C-1.

4160V AC buses A5 and A6 are the emergency buses. These electrical buses provide power to vital equipment such as RHR , Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses B1 and B2 . Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators . This EAL does consider backscuttle capability. But due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers.

This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SA1.1 .

N ON-QUALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=- Entergy . ADMINISTRATIVE PROCE DURE PROC EDURES REFERENCE USE Page 202 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 5 of 64 Unusual Event - CU1 .1 PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31 , "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator"
8. PNPS 2.4.144, "Degraded Voltage "
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line "
10. PNPS 2.4.A.5 , "Loss of Electrical Bus A5" 11 . PNPS 2.4.A.6 , "Loss of Electrical Bus A6"

NON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE PROCE DURE PROCEDURES REFERENCE USE Page 203 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 6 of 64 Unusual Event - CU1 .1 Figure C-1 Emergency Bus Composite Diagram 24 KV 24 KV 345 KV

~ UAT "A"

sigt ~SOT

~

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801 EO G EOG

) 600 I

PNPS

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-=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 204 Revision 9 of 351 Emergency Action Level Technical Bases Document

~ - -- - -- - -

ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 7 of 64 Alert - CA1 .1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 1 - Loss of Power Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer EAL:

CA1.1 Alert Loss of all offsite and all onsite AC power (Table C-4) to emergency buses A5 and A6 for

~ 15 min. (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C-4 AC Power Sources Offsite

- Startup Transformer (X4)

- Shutdown Transformer

- UAT

- Backscuttle via Main Transformer (only if already established)

Onsite

- EDGA

- EOG B

- SBO DG

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 205 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 8 of 64 Alert- CA1.1 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel, D - Defueled NEI 99-01 Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS , Containment Heat Removal, Spent Fuel Heat Removal , and the Ultimate Heat Sink.

The event can be classified as an Alert when in cold shutdown, refueling , or defueled mode because of the significantly reduced decay heat and lower temperature and pressure, increasing the time to restore one of the emergency buses, relative to that specified for the

  • Site Area Emergency EAL .

Escalating to Site Area Emergency, if appropriate, is by Category A EALs.

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

PNPS Basis:

This EAL is indicated by the loss of all offsite and onsite AC power to the emergency buses A5 and A6 .

The PNPS emergency 4160V AC electrical distribution system is illustrated in Figure C-1.

The inability to provide emergency AC power from any offsite power source poses a potential degradation of reactor plant safety. If onsite power capability becomes degraded , significant losses of vital equipment operability may occur. 4160V AC buses A5 and A6 are the emergency buses . These electrical buses provide power to vital equipment such as RHR, Core Spray, and CRD pumps as well as vital 480V AC transformers feeding buses 81 and 82.

Offsite power supply transformers are those transformers which are capable of providing power to the emergency buses independent of the onsite generators. This EAL does consider backscuttle capability. But due to the significant time required to establish the necessary switching lineups and removal of generator disconnect links, backscuttle is not considered an available offsite power supply transformer unless already established at the time of the loss of other offsite power supply transformers .

This EAL is the cold condition equivalent of the hot condition loss of all AC power EAL SS1 .1.

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 206 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 9 of 64 Alert - CA 1.1 PNPS Basis Reference(s):

1. Bechtel Drawing Electrical Single Line Diagram S-E-155
2. FSAR Section 8.1
3. FSAR Section 8.3
4. FSAR Section 8.5
5. PNPS 5.3.31 , "Station Blackout"
6. PNPS 2.2.8, "Standby AC Power System (Diesel Generators)"
7. PNPS 2.2.146, "Station Blackout Diesel Generator"
8. PNPS 2.4.144, "Degraded Voltage "
9. PNPS 2.4.A.23, "Loss/Degradation Of 23kV Line"
10. PNPS 2.4.A.5, "Loss of Electrical Bus A5" 11 . PNPS 2.4.A.6, "Loss of Electrical Bus A6"
  • --=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES N ON-Q UALITY R ELATED PROCEDURE REFERENCE USE EP-AD-601 Page 207 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 10 of 64 Alert- CA1 .1 Figure C-1 Emergency Bus Composite Diagram 24 KV 24 KV 345 KV

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o EMERGENCY BUS rn COMPOSITE

NON- QUALI TY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 208 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 11 of 64 Unusual Event - CU2.1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: RCS leakage EAL:

CU2.1 Unusual Event RPV level cannot be restored and maintained> +12 in. for ~ 15 min . (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but

  • should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown NEI 99-01 Basis:

This EAL is considered to be a potential degradation of the level of safety of the plant. The inability to maintain or restore level is indicative of loss of RCS inventory.

Relief valve normal operation should be excluded from this EAL. However, a relief valve that operates and fails to close according to design should be considered applicable to this EAL if the relief valve cannot be isolated .

Prolonged loss of RCS Inventory may result in escalation to the Alert emergency classification level via either CA2.1 or CA3.1.

PNPS Basis :

The condition of this EAL may be a precursor of more serious conditions and , as a result, is considered to be a potential degradation of the level of safety of the plant. When RPV level drops to 12 in . (low level scram setpoint) , level is well below the normal control band and

  • automatic RPS and PCIS actuations are required (ref. 1).

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EM ERGENCY PLAN ADMIN ISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 209 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 12 of 64 Unusual Event - CU2 .1 All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482 .5 inches above the bottom of the reactor vessel , 126.3 inches above the top of the active fuel (TAF). The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482 .5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero. The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll-263-106A & B, Ll-263-73A & B, Ll -1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions. Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 2).

In preparation for refueling operations, a temporary shutdown/floodup transmitter is installed

  • on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054 ). This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 2) .

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 3).

This Cold Shutdown EAL represents the hot condition EAL SU6 .1 in wh ich RCS leakage is associated with Technical Specifications limits. In Cold Shutdown, these limits are not applicable; hence , the use of RPV level as the parameter of concern in this EAL (ref. 4).

PNPS Basis Reference(s):

1. EOP-1 , "RPV Control"
2. PNPS 2.2.80 , "Reactor Vessel Level, Temperature, And Internal Pressure Instrumentation"
3. PNPS 3.M .2-40 , "Refuel Outage Tempora ry Modification Reactor Shutdown/Flood-Up Level Indication "
4. NEI/NRC EAL FAQ #2006-014

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 210 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 13 of 64 Unusual Event - CU2 .1 Figure C-3 Floodup RPV Level Indication

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NON- QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCE DURE PROCEDURES REFERENCE USE Page 211 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 14 of 64 Unusual Event - CU2.2 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL:

CU2.2 Unusual Event Unplanned RPV level drop for ~ 15 min (Note 3) below EITHER:

  • RPVflange (+182 in.)

OR RPV level band when the RPV level band is established below the RPV flange Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

5 - Refuel NEI 99-01 Basis:

This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RPV water level below the RPV flange are carefully planned and procedurally controlled . An unplanned event that results in water level decreasing below the RPV flange , or below the planned RPV water level for the given evolution (if the planned RPV water level is already below the RPV flange), warrants declaration of a UE due to the reduced RPV inventory that is available to keep the core covered.

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

--=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 212 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 15 of 64 Unusual Event - CU2.2 The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame , then it may indicate a more serious condition exists.

Continued loss of RCS Inventory will result in escalation to the Alert emergency classification level via either EAL CA2 .1 or EAL CA3 .1.

This EAL involves a decrease 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 decreases in flooded reactor cavity level , which is addressed by EAL AU2 .1, until such time as the level decreases to the level of the vessel flange.

PNPS Basis:

181.87 in. (rounded to 182 in. for readability) on the Shutdown RPV water level instrument corresponds to the RPV flange (ref. 1 ).

All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482 .5 inches above the bottom of the reactor vessel , 126.3 inches above the top of the active fuel (T AF). The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482.5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero. The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll-263-106A & B, Ll -263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions . Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 2).

In preparation for refueling operations, a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054). This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 1, 2).

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 1) .

NON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 213 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 16 of 64 Unusual Event - CU2.2 PNPS Basis Reference(s):

1. PNPS 3.M.2-40 , "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication"
2. PNPS 2.2.80 , "Reactor Vessel Level, Temperature, And Internal Pressure Instrumentation"

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERG ENCY PLAN ADM INISTRATIVE PROCEDURE PROCEDU RES REFERENCE USE Page 214 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 17 of 64 Unusual Event - CU2 .2 Figure C-3 Floodup RPV Level Indication

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e-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 215 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 18 of 64 Unusual Event - CU2.3 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Unplanned loss of RPV inventory EAL:

CU2.3 Unusual Event RPV level cannot be monitored with any unexplained RPV leakage indication , Table C-1

  • Table C-1 RPV Leakage Indications

- Drywell equipment drain sump level rise

- Drywell floor drain sump level rise

- Reactor Building equipment drain sump level rise

- Reactor Building floor drain sump level rise

- Torus level rise

- RPV make-up rate rise

- Observation of unisolable RCS leakage Mode Applicability:

5 - Refuel

e-=- Entergy . PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDU RE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 216 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 19 of 64 Unusual Event - CU2.3 NEI 99-01 Basis:

This EAL is a precursor of more serious conditions and considered to be a potential degradation of the level of safety of the plant.

Refueling evolutions that decrease RPV water level below the RPV flange are carefully planned and procedurally controlled. An unplanned event that results in water level decreasing below the RPV flange, or below the planned RPV water level for the given evolution (if the planned RPV water level is already below the RPV flange), warrants declaration of a UE due to the reduced RPV inventory that is available to keep the core covered .

Continued loss of RCS inventory will result in escalation to the Alert emergency classification level via either EAL CA2 .1 or EAL CA3.1 .

  • This EAL addresses conditions in the refueling mode when normal means of core temperature indication and RPV level indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RPV inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes . Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

PNPS Basis:

All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482 .5 inches above the bottom of the reactor vessel , 126.3 inches above the top of the active fuel (TAF) . The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482 .5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero . The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll -263-106A & B, Ll-263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001 -604A & B (Pen 3) are used to monitor RPV level during accident conditions . Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 2).

  • -~ Entergy PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 217 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 20 of 64 Unusual Event - CU2.3 In preparation for refueling operations, a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054). This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 1, 2).

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 1).

In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage. If the makeup rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified . Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory (ref. 3, 4, 5, 6, 7) .

PNPS Basis Reference(s):

1. PNPS 3.M.2-40, "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication"
2. PNPS 2.2.80, "Reactor Vessel Level, Temperature, And Internal Pressure Instrumentation"
3. PNPS 2.2 .77, "Drywe/1 Leak Detection Systems"
4. PNPS 2.5.2 .71, "Radwaste Collection System"
5. FSAR Section 4.10 - Nuclear System Leakage Rate Limits
6. FSAR Section 5.2 - Primary Containment System
7. FSAR Section 9.2 - Liquid Radwaste System

NON- QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 218 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 21 of 64 Unusual Event - CU2 .3 Figure C-3 Floodup RPV Level Indication R l!1. F'U)CJA It. IIT

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N ON- QUALITY PNPS

~ RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 219 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 22 of 64 Alert - CA2 .1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA2.1 Alert RPV level< -45 in .

OR RPV level cannot be monitored for ~ 15 min. with any unexplained RPV leakage indication , Table C-1 (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C - 1 RPV Leakage Indications

- Drywell equipment drain sump level rise

- Drywell floor drain sump level rise

- Reactor Building equipment drain sum p level rise

- Reactor Building floor drain sump level rise

- Torus level rise

- RPV make-up rate rise

- Observation of unisolable RCS leakage

--=-Entergy PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 220 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 23 of 64 Alert - CA2.1 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL serves as a precursor to a loss of ability to adequately cool the fuel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery. This condition will result in a minimum emergency classification level of an Alert.

The inability to restore and maintain level after reaching this setpoint (-45 in .) would be indicative of a failure of the RCS barrier.

If RPV level continues to lower, escalation to Site Area Emergency will be via EAL CS2 .1 or CS2 .2.

PNPS Basis :

The threshold RPV level of -45 in. is the low-low ECCS actuation setpoint (ref. 1). Low-low level switches LIS-263-72A, B, C, and D initiate Reactor Core Injection Cooling (RCIC),

Automatic Depressurization System (ADS), and start the standby diesel generator (DG) at

-46.3 in . (rounded to -45 in. for readability in accordance with EOPs). Core Spray (CS) is also initiated if Reactor pressure is less than 395 to 405 psig or after a 9 to 15.4-minute time delay.

Low level switches LS-263-72A-1, 8-1, C-1, and D-1 initiate High Pressure Core Injection (HPCI) and start the standby DG at this level. Residual Heat Removal (RHR) will also be initiated (ref. 3).

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 221 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 24 of 64 Alert - CA2.1 All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482 .5 inches above the bottom of the reactor vessel, 126.3 inches above the top of the active fuel (TAF). The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482.5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero. The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll-263-106A & B, Ll-263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions. Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 3).

In preparation for refueling operations, a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054 ). This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 2, 3).

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 2).

Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage. If the makeup rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory (ref. 4, 5, 6, 7, 8) .

NON- Q UALI TY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 222 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 25 of 64 Alert - CA2.1 PNPS Basis Reference(s):

1. EOP-1 , "RPV Control"
2. PNPS 3.M .2-40, "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication"
3. PNPS 2.2.80 , "Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation"
4. PNPS 2.2.77 , "Orywe/1 Leak Detection Systems"
5. PNPS 2.5.2.71, "Radwaste Collection System"
6. FSAR Section 4.10 - Nuclear System Leakage Rate Limits
7. FSAR Section 5.2 - Primary Containment System
8. FSAR Section 9.2 - Liquid Radwaste System

N ON- Q UALITY PNPS

~ R ELATED EP-AD-601 Revision 9

""=- Entergy . EMERGENCY PLAN ADMINISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 223 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 26 of 64 Alert - CA2 .1 Figure C-3 Floodup RPV Level Indication OV!

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  • 3.M.240 Re . 5 Page26of28

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 224 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 27 of 64 I Site Area Emergency - CS2 .1 I Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS2.1 Site Area Emergency With Containment Closure not established, RPV level< -50 in. (Note 4)

Note 4: Containment closure is the action taken to secure Primary or Secondary Containment

  • and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions . Containment closure is established when Primary or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications.

Mode Applicability:

4 - Cold Shutdown , 5 - Refuel NEI 99-01 Basis:

Under the conditions specified by this EAL , continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach , pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.

Escalation to a General Emergency is via EAL CG2 .1, CG2.2, AG1 .1, AG1 .2, or AG1 .3.

NON-Q UALI TY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 225 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 28 of 64 Site Area Emergency - CS2 .1 PNPS Basis:

When RPV level decreases to -50 in . on the normal range level indicators (i.e., the -50 to +50 instruments), water level is 5 inches below the low-low ECCS actuation setpoint (ref. 1, 4).

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery .

The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and potential loss of the fuel clad barrier.

Containment closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 2). As applied to PNPS, containment closure is established when either Primary Containment integrity or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications (ref. 3).

PNPS Basis Reference(s):

1. EOP-1 , "RPV Controf'
2. NEI 99-01 Definitions and Appendix C
3. Technical Specifications Section 3.7
4. PNPS 2.2.80 , "Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation"

N ON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 226 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 29 of 64 Site Area Emergency - CS2 .2 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS2.2 Site Area Emergency With containment closure established , RPV level < -125 in . (Note 4)

Note 4: Containment closure is the action taken to secure Primary or Secondary Containment

  • and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment closure is established when Primary or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

Under the conditions specified by this EAL, continued decrease in RPV level is ind icative of a loss of inventory control. Inventory loss may be due to an RCS breach , pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted .

Escalation to a General Emergency is via CG2 .1, CG2 .2, AG1 .1, AG1 .2 , or AG1 .3.

N ON-QUALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 227 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 30 of 64 Site Area Emergency - CS2.2 PNPS Basis :

When RPV level drops below -125 in. , core uncovery is about to occur (ref. 1).

The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level decrease and potential core uncovery.

The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and potential loss of the fuel clad barrier.

Containment closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 2). As applied to PNPS ,

containment closure is established when either Primary Containment integrity or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications (ref. 3).

PNPS Basis Reference(s):

1. EOP-1, "RPV Control"
2. NEI 99-01 Definitions and Append ix C
3. Technical Specifications Section 3.7

N ON- Q UALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 228 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 31 of 64 Site Area Emergency - CS2.3 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS2.3 Site Area Emergency RPV level cannot be monitored for ~ 30 min. (Note 3) with a loss of inventory as indicated by EITHER:

Unexplained RPV leakage indication, Table C-1 Erratic Source Range Monitor indication Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Table C~1 RPV Leakage Indications

- Drywell equipment drain sump level rise

- Drywell floor drain sump level rise

- Reactor Building equipment drain sump level rise

- Reacto r Building floor drain sump level rise

- Torus level rise

- RPV make.up rate rise

- Observation of unisolable RCS leakage

N ON-QUALI TY PNPS RE LATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 229 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 32 of 64 Site Area Emergency - CS2 .3 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-0 1 Basis:

Under the conditions specified by this EAL, continued decrease in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RCS breach , pressure boundary leakage, or continued boiling in the RPV. Thus, declaration of a Site Area Emergency is warranted.

Escalation to a General Emergency is via EAL CG2 .1, CG2 .2, AG 1 .1 , AG 1 .2, or AG 1 .3 .

  • The 30-minute duration allows sufficient time for actions to be performed to recover inventory control equipment.

PNPS Basis :

All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482.5 inches above the bottom of the reactor vessel , 126.3 inches above the top of the active fuel (TAF) . The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482 .5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero. The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll-263-106A & B, Ll-263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions. Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 2).

In preparation for refueling operations , a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054 ). This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 1, 2) .

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 1) .

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

--=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 230 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 33 of 64 Site Area Emergency - CS2.3 In this EAL, all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage . Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage. If the makeup rate to the RPV unexplainably rises above the pre-established rate , a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified . Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory (ref. 3, 4, 5, 6, 7).

PNPS Basis Reference(s):

1. PNPS 3.M.2-40 , "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication "
2. PNPS 2.2.80, "Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation"
3. PNPS 2.2.77 , "Drywe/1 Leak Detection Systems"
4. PNPS 2.5.2.71, "Radwaste Collection System"
5. FSAR Section 4.10 - Nuclear System Leakage Rate Limits
6. FSAR Section 5.2 - Primary Containment System
7. FSAR Section 9.2 - Liquid Radwaste System

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

~ Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 231 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 34 of 64 Site Area Emergency - CS2.3 Figure C-3 Floodup RPV Level Indication

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  • 3.M..2 0 Re . 5 Page26of28

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 232 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 35 of 64 General Emergency - CG2.1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG2.1 General Emergency RPV level< -125 in. for ~ 30 min. (Note 3)

Any containment challenge indication , Table C-5 Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determ ined that the condition will likely exceed the applicable time .

Table c..s Containment Challenge Indications

- Containment Closure not established (Note 4 )

- Deflagration concentrations exist inside PC

- Unplanned rise in PC pressure

- Secondary Containment area radiation> any Maximum Safe Operating Value (EOP-4, Table L)

Note 4: Containment closure is the action taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. Containment closure is established when Primary or Secondary Containment integrity is established in

  • accordance with Section 3.7 of Technical Specifications.

J

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

~ Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 233 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 36 of 64 General Emergency - CG2.1 Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-0 1 Basis:

This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored as available decay heat will cause boiling, further reducing the RPV level. With the containment breached or challenged , 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 .

NON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCED URE PROCEDURES REFERENCE USE Page 234 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 37 of 64 General Emergency - CG2.1 PNPS Basis:

When RPV level drops below -125 in ., core uncovery is about to occur (ref. 1).

Four conditions are associated with a challenge to Primary Containment (PC) integrity:

  • Containment closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures , systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 2). As applied to PNPS , containment closure is established when either Primary Containment integrity or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications (ref. 3).
  • elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs (Severe Accident Management Guides) indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term . Post-LO CA hydrogen generation primarily caused by radiolysis is a slowly evolving , long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative , therefore, of a potential threat to Primary Containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 4 ).

Except for brief periods during plant startup and shutdown, oxygen co ncentration in the Primary Containment is maintained at insignificant levels by nitrogen inertion. The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 3) and readily recognizable because 6%

hydrogen is well above the EOP-3 , "Primary Containment Control", entry condition (ref. 5).

The H2/02 system monitors Primary Containment hydrogen (H2) and oxygen (02) gas concentrations via the following containment penetration sample points: X-29E ,

X-106A-b, X-228J , X-15E, X-50A-d, and X-228C. The H2/02 system returns its samples to Primary Containment via containment penetrations X-46F and X-228K .

NON-QUALITY PNPS RELATE D EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 235 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 38 of 64 General Emergency - CG2.1 The analyzer system consists of two separate units: an analyzer panel and a remote panel. Analyzer Panels C172 and C173 are located close to Primary Containment on the 74'3" elevation of the Reactor Building . Each panel contains all essential components necessary to perform the required containment atmosphere analysis, indication, and alarm functions . Recorders AR-1001-612A and AR-1001-612B and indicating meters are located in the Control Room in Panels C174 and C175 . The backup Com sip H2/02 analyzers, C172 and C173 , are set to alarm at either 4% 02 or 4% H2 and are normally left in the "STANDBY" mode (ref. 6, 7, 8).

  • Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicate containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release .

The Secondary Containment area radiation Maximum Safe Operating Values are indicative of problems in the Secondary Containment that are spreading . The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-4 , "Secondary Containment Control", Table L (ref. 9).

PNPS Basis Reference{s):

1. EOP-1 , "RPV Control"
2. NEI 99-01 Definitions and Appendix C
3. Technical Specifications Section 3.7
4. Plant Specific Technical Guidelines for EOPs and SAGs, Section PC/G
5. EOP-3, "Primary Containment Controf'
6. PNPS 2.2.120, "Postaccident Monitoring Panel"
7. PNPS 2.2.133 , "H2!02 Analyzer and C19 Systems"
8. SAG-02 , "Containment and Radioactivity Release Controf'
9. EOP-4 , "Secondary Containment Control"

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=- Entergy . ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 236 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 39 of 64 General Emergency - CG2.1 EOP-4 Table L Secondary Containment Control Area Radiation Maximum Safe Operating Values RADIATION LEVEL VALUE (mR/hr)

NW equipment space/HPCI pump room - 17 ft 6 in . El. 1000 CRD pump room - 17 ft 6 in . El. 1000 RCIC pump room - 17 ft 6 in . El. 1000 SE equipment space - 17 ft 6 in . El. 1000 CRD HCU west area - 23 ft El. 1000 CRD HCU east area - 23 ft El. 1000 RB west area - 51 ft El. 1000 RB east area - 51 ft El. 1000 North Storage and laydown area - 74 ft 3 in. El. (H202) 1000 Fuel pool cooling pump/HX area - 74 ft 3 in . El 1000 SLC pump area - 91 ft 3 in . El. 1000 Skimmer surge tank area - 91 ft 3 in . El. 1000

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 237 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 40 of 64 General Emergency - CG2.2 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 2 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG2.2 General Emergency RPV level cannot be monitored for ~ 30 min. (Note 3) with a loss of inventory as indicated

  • by EITHER:

Unexplained RPV leakage indication, Table C-1 Erratic source range monitor indication AND Any containment challenge indication, Table C-5 Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time .

--=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 238 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 41 of 64 General Emergency - CG2.2 Table C-1 RPV Leakage Indications

- Drywell equipment drain sump level rise

- Drywell floor drain sump level rise

- Reactor Building equipment drain sump level rise

- Reactor Building floor drain sump level rise

- Torus level rise

- RPV make-up rate rise

- Observation of unisolable RCS leakage Table C-5 Containment Challenge Indications

- Containment Closure not established (Note 4)

- Deflagration concentrations exist inside PC

- Unplanned rise in PC pressure

- Secondary Containment area radiation > any Maximum Safe Operating Value (EOP-4, Table L}

Note 4: Containment closure is the action taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions . Containment closure is established when Primary or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel

e-=-Entergy PNPS N ON-Q UALI TY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE Use Page 239 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 42 of 64 General Emergency - CG2.2 NEI 99-01 Basis:

This EAL represents the inability to restore and maintain RPV level to above the top of active fuel with containment challenged. Fuel damage is probable if RPV level cannot be restored as available decay heat will cause boiling , further reducing the RPV level. With the containment breached or challenged , 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.

Sump and tank level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

PNPS Basis :

All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482.5 inches above the bottom of the reactor vessel, 126.3 inches above the top of the active fuel (TAF). The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482 .5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero. The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll-263-106A & B, Ll-263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions. Level Recorder LR-640-28 is used to monitor RPV level during shutdown conditions (ref. 2) .

N ON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=- Entergy . ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 240 of 351 Emergency Action Level Technical Bases Document

~

ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 43 of 64 General Emergency - CG2 .2 In preparation for refueling operations, a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054) . This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 1, 2) .

Figure C-3 illustrates the elevations of the temporary RPV level flood up instrument (ref. 1).

In this EAL , all water level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table C-1 . Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building

  • equipment or floor drain sump level rise may also be indicative of RCS inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode , an unexplained rise in torus level could be indicative of RHR valve misalignment or leakage . If the makeup rate to the RPV unexplainably rises above the pre-established rate , a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified . Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory (ref. 3, 4, 5, 6, 7).

Four conditions are associated with a challenge to Primary Containment (PC) integrity:

  • Containment closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 8) . As applied to PNPS , containment closure is established when either Primary Containment integrity or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications (ref. 9).
  • --=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE PROCEDURES NON-QUALITY RELATED PROCEDURE REFERENCE USE EP-AD-601 Page 241 Revision 9 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 44 of 64 General Emergency - CG2.2

  • Deflagration (explosive) mixtures in the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen . BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term.

Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition . Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction . A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative , therefore, of a potential threat to Primary Containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 10).

  • Except for brief periods during plant startup and shutdown, oxygen concentration in the Primary Containment is maintained at insignificant levels by nitrogen inertion . The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 10) and readily recognizable because 6%

hydrogen is well above the EOP-3, "Primary Containment Control", entry condition (ref. 11 ).

The H2/02 system monitors Primary Containment hydrogen (H2) and oxygen (02) gas concentrations via the following containment penetration sample points: X-29E, X-106A-b, X-228J , X-15E , X-50A-d, and X-228C. The H2/02 system returns its samples to Primary Containment via containment penetrations X-46F and X-228K.

The analyzer system consists of two separate units: an analyzer panel and a remote panel. Analyzer Panels C172 and C173 are located close to Primary Containment on the 74'3" elevation of the Reactor Building. Each panel contains all essential components necessary to perform the required containment atmosphere analysis, indication, and alarm functions . Recorders AR-1001-612A and AR-1001-612B and indicating meters are located in the Control Room in Panels C174 and C175 . The backup Comsip H2/02 analyzers, C172 and C173 , are set to alarm at either 4% 02 or 4% H2 and are normally left in the "STANDBY" mode (ref. 12, 13, 14).

  • Any unplanned increase in PC pressure in the Cold Shutdown or Refueling mode indicates a potential loss of containment closure capability. Unplanned Primary Containment pressure increases indicate containment closure cannot be assured and the Primary Containment cannot be relied upon as a barrier to fission product release.

NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 242 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 45 of 64 General Emergency - CG2.2

PNPS Basis Reference(s):

1. PNPS 3.M.2-40 , "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication "
2. PNPS 2.2.80, "Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation"
3. PNPS 2.2.77 , "Drywe/1 Leak Detection Systems"
4. PNPS 2.5.2.71, "Radwaste Collection System" *
5. FSAR Section 4.10 - Nuclear System Leakage Rate Limits
6. FSAR Section 5.2 - Primary Containment System
7. FSAR Section 9.2 - Liquid Radwaste System
8. NEI 99-01 Definitions and Appendix C
9. Technical Specifications Section 3.7
10. Plant Specific Technical Guidelines for EOPs and SAGs, Section PC/G 11 . EOP-3 , "Primary Containment Control"
12. PNPS 2.2.120, "Postaccident Monitoring Panel"
13. PNPS 2.2.133 , "H2/02 Analyzer and C19 Systems"
14. SAG-02 , "Containment and Radioactivity Release Control"
15. EOP-4 , "Secondary Containment Control and Radioactivity Release Control"

NON-QUALITY

'=' PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 243 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 46 of 64 General Emergency - CG2.2 Figure C-3 Floodup RPV Level Indication C90II~

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NON-QUALITY A PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 244 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 47 of 64 General Emergency - CG2.2 EOP-4 Table L Secondary Containment Control Area Radiation Maximum Safe Operating Values RADIATION LEVEL VALUE (mR/hr)

NW equipment space/HPCI pump room - 17 ft 6 in . El. 1000 CRD pump room - 17 ft 6 in . El. 1000 RCIC pump room - 17 ft 6 in . El. 1000 SE equipment space - 17 ft 6 in . El. 1000 CRD HCU west area - 23 ft El. 1000 CRD HCU east area - 23 ft El. 1000 RB west area - 51 ft El.

RB east area - 51 ft El.

North Storage and laydown area - 74 ft 3 in . El. (H202 )

Fuel pool cooling pump/HX area - 74 ft 3 in. El SLC pump area - 91 ft 3 in . El.

Skimmer surge tank area - 91 ft 3 in . El.

1000 1000 1000 1000 1000 1000

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 245 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 48 of 64 Unusual Event - CU3.1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL:

CU3.1 Unusual Event Any unplanned event results in RCS temperature> 212°F due to loss of decay heat removal capabil ity (Note 5)

Note 5: ILRT, CRD scram time testing , and hydrostatic testing in which RCS temperature is intentionally raised above 212°F in accordance with Technical Specifications LCO 3.14 are not applicable to these EALs.

Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL is a precursor of more serious conditions and , as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow . Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the 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 decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RCS temperatures depending on the time since shutdown .

  • Escalation to Alert would be via EAL CA2.1 based on an inventory loss or EAL CA3 .1 based on exceeding its temperature criteria.

N ON-Q UALITY PNPS

~ R ELATED EP-AD-601 Revision 9 EMERGEN CY PLAN

--=-Entergy . ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 246 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 49 of 64 Unusual Event - CU3 .1 PNPS Basis:

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specifications cold shutdown temperature limit (212 °F, ref. 1). These include (ref. 2):

  • Reactor vessel shell adjacent to reactor vessel flange (TR-263-105 red pen on Panel C904 or TR-263-104 Pt. 2 Panel C921)
  • Reactor vessel bottom head temperature TR-263-104 Pt. 9 Panel C92 1

The only other viable mechanisms for maintaining reactor coolant temperature< 212°F are to maximize RWCU cooling via the NRHX or to establish a feed/bleed with feed/condensate and RWCU reject. Should SOC be unavailable, reactor decay levels could potentially be well in excess of either of these other mechanisms . The inability to establish and maintain cold shutdown conditions under those cond itions for which it is required poses a significant threat to reactor plant safety.

N ON-QUALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 247 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 50 of 64 Unusual Event - CU3.1 The note is a reminder that for certain planned and controlled events , RCS temperature may be raised above the LCO without changing the operating mode to hot conditions. For such events, the RCS temperature specified in the Technical Specifications definition of "cold shutdown" may be considered "NA" and operation considered not to be> 212°F or in hot shutdown mode while RCS temperature is above 212°F . With increased reactor vessel fluence over time, the minimum allowable vessel temperature increases at a given pressure.

Periodic updates to the RPV PIT limit curves are performed as necessary based upon the results of analyses of irradiated surveillance specimens removed from the vessel. In the future it is expected that hydrostatic and leak testing may eventually be required with minimum reactor coolant temperatures exceeding 212°F. However, even with required minimum temperature requirements below 212°F , maintaining RCS temperature within a small band during the test can be impractical. Removal of the heat addition from recirculation pump

  • operation and reactor core decay heat is coarsely controlled by control rod drive hydraulic system flow and reactor water cleanup system nonregenerative heat exchanger operation .

Test conditions are focused on maintaining a steady state pressure , and tightly limited temperature control poses an unnecessary burden on the operator and may not be achievable in certain instances (ref. 4 ).

PNPS Basis Reference(s):

1. Technical Specifications, Section 1.0, Definitions
2. PNPS 2.1.7, "Vessel Heatup and Coo/down "
3. PNPS 2.4.25, "Loss of Shutdown Cooling"
4. Technical Specifications Section 3.14

N ON-QUALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 248 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 51 of 64 Unusual Event - CU3.2 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Unplanned loss of decay heat removal capability with irradiated fuel in the RPV EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for ~ 15 min. (Note 3)

Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown , 5 - Refuel NEI 99-01 Basis:

This EAL is a precursor of more serious conditions and , as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow . Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RPV inventory. Since the 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 decrease water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid increases in RPV temperatures depending on the time since shutdown .

J

N ON -QUALITY PN PS RE LATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADM INISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 249 of 351 Emergency Action Level Techn ical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 52 of 64 Unusual Event - CU3.2 Redundant means of RPV level indication are procedurally installed to assure that the ability to monitor level will not be interrupted . However, if all level and temperature indications were to be lost in either the cold shutdown or refueling modes, this EAL would result in declaration of a UE if both temperature and level indications cannot be restored within 15 minutes from the loss of both means of indication. Escalation to Alert would be via EAL CA2.1 based on an inventory loss or EAL CA3 .1 based on exceeding its temperature criteria .

PNPS Basis :

All Control Room RPV water level monitors have their indicated range referenced to an elevation of 482 .5 inches above the bottom of the reactor vessel , 126.3 inches above the top of the active fuel (TAF). The common zero reference for RPV water level instrumentation was chosen as the bottom of the steam separators, 482.5 inches above the bottom of the reactor vessel. The low level instruments cover the level from 142" above vessel zero to 542" above vessel zero . The wide range instrument covers the upper 400" of the vessel from 425" above vessel zero to top of the vessel. Level Indicators Ll -263-106A & B, Ll-263-73A & B, Ll-1001-650A & B, and Level Recorders LR-1001-604A & B (Pen 3) are used to monitor RPV level during accident conditions . Level Recorder LR-640-28 is used to monitor RPV level during shutdown cond itions (ref. 2).

In preparation for refuel ing operations, a temporary shutdown/floodup transmitter is installed on Reactor Building 51' Rack C2206 and the narrow range feedwater level indicator Ll-640-29B is replaced with a reactor shutdown/floodup digital level display with visual alarm points (EPIC point RXX054) . This allows monitoring RPV level between "normal range" (instrument Oto 466") through flood elevation (117') (ref. 1, 2).

Figure C-3 illustrates the elevations of the temporary RPV level floodup instrument (ref. 1 ) .

N ON- QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=- Entergy . ADMINISTRATIVE P ROCEDU RE PROCED URES REFERENCE USE Page 250 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 53 of 64 Unusual Event - CU3 .2 Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specifications cold shutdown temperature limit (212 °F, ref. 3). These include (ref. 4) :

  • Reactor vessel shell adjacent to reactor vessel flange (TR-263-105 red pen on Panel C904 or TR-263-104 Pt. 2 Panel C921)
  • Reactor vessel bottom head temperature TR-263-104 Pt. 9 Panel C921
  • EPIC points RXX030 , REC068, REC072 , REC074 , REC070 , REC076, REC078 PNPS Basis Reference(s):
1. PNPS 3.M.2-40 , "Refuel Outage Temporary Modification Reactor Shutdown/Flood-Up Level Indication "
2. PNPS 2.2.80 , "Reactor Vessel Level, Temperature, and Internal Pressure Instrumentation"
3. Technical Specifications, Section 1.0, Definitions
4. PNPS 2.1.7, "Vessel Heatup and Coo /down "

N ON-Q UALITY PNPS

~ R ELATED EP-AD-601 Revision 9

"=- Entergy . EMERGENCY PLAN ADMINISTRATIVE P ROCEDURE PROCED URES REFERENCE USE Page 251 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 54 of 64 Unusual Event - CU3.2 FIGURE C-3 FLOODUP RPV LEVEL INDICATION RtFUEl. F\.OOR EL II T

* EL 116" --------- -

~-- --------* El..112' --------

. IIEACTOR CA\l!TY

' ------~-- . 1'11.S' ,w 1

  • 111.17 279* O"

-Ir~-------- -_._,._ -

  • 1 N1 II LO l!J TElPO AV

!tiUTOOWM ll!VEL TRAHSMl1'TER

N ON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 252 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 55 of 64 Alert - CA3 .1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain plant in cold shutdown EAL:

CA3.1 Alert Any unplanned event results in RCS temperature> 212°F for> Table C-3 duration (Note 5)

OR RPV pressure increase > 15 psig due to a loss of RCS cooling Note 5: ILRT, CRD scram time testing and hydrostatic testing in which RCS temperature is intentionally raised above 212°F in accordance with Technical Specification s LCO 3.14 are not applicable to these EALs Tabfe c..a RCS R&heat Duration Threshold

  • If an RCS heal removal sy tem It In operation w, hln this Ume frame and RCS temperatur ls be g reduced, the E Lis no a tk:abt
1. RCS intact (Containment Closure NIA) 60 min.*
2. Conta*nment Closure established A o 20min.

RCS o intact

3. Containment Closure no establi ed AD Omin.

RCS ot intact

N ON- Q UALITY PNPS RE LATE D EP-AD-60 1 Revision 9

~ Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 253 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 56 of 64 Alert - CA3 .1 Mode Applicability:

4 - Cold Shutdown , 5 - Refuel NEI 99-01 Basis:

The RCS Reheat Duration Thresholds table (Table C-3) 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 . 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 Thresholds 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 or RCS inventory is reduced . The allowed 20-minute time frame was included to allow operator action to restore the heat removal function if possible.

Finally, complete loss of functions required for core cooling during refueling and cold shutdown modes when neither containment closure nor RCS integrity are established. No delay time is allowed because the evaporated reactor coolant that may be released into the containment during this heatup condition could also be directly released to the environment.

The note (*) in the table 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.

The 15 psi pressure increase addresses situations where , due to high decay heat loads, the time provided to restore temperature control should be less than 60 minutes . The RCS pressure setpoint chosen is the lowest pressure that the site can read on installed control board instrumentation .

Escalation to Site Area Emergency would be via EAL CS2.1 , CS2.2 , or CS2.3 should boiling result in significant RPV level loss leading to core uncovery.

A loss of Technical Specifications components alone is not intended to constitute an Alert .

The same is true of a momentary unplanned excursion above the Technical Specifications cold shutdown temperature limit when the heat removal function is available .

N ON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDU RE PROC EDURES REFERENCE USE Page 254 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 57 of 64 Alert - CA3 .1 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 .

PNPS Basis :

A 15 psig RPV pressure increase can be read on pressure indicators Pl-640-25A/B and pressure recorder 6-PR-640-27 at Panel C905 (ref. 1).

Several instruments are capable of providing indication of RCS temperature with respect to the Technical Specifications cold shutdown temperature limit (212°F, ref. 2) . These include (ref. 3):

  • Reactor vessel shell adjacent to reactor vessel flange (TR-263-105 red pen on Panel C904 or TR-263-104 Pt. 2 Panel C921)
  • Reactor vessel bottom head temperature TR-263-104 Pt. 9 Panel C921
  • EPIC points RXX030 , REC068, REC072 , REC074 , RECO?O , REC076 , REC078 Containment closure is the action taken to secure either Primary Containment or Secondary Containment and the associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions (ref. 4) . As applied to PNPS ,

containment closure is established when either Primary Containment integrity or Secondary Containment integrity is established in accordance with Section 3.7 of Technical Specifications (ref. 5)

NON-QUALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 255 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 58 of 64 Alert - CA3 .1 The note is a reminder that any temperature increase above 212°F is an operating mode change from cold to hot conditions . Since each EAL is assigned one or more operating modes, the set of EALs that must be monitored must now include EALs associated with hot condition operating modes. The note is a reminder that for certain planned and controlled events RCS temperature may be raised above the LCO without changing the operating mode to hot conditions. For such events, the RCS temperature specified in the Technical Specifications definition of "Cold Shutdown" may be considered "NA" and operation considered not to be> 212°F or in Hot Shutdown mode while RCS temperature is above 212°F . With increased reactor vessel fluence over time , the minimum allowable vessel temperature increases at a given pressure . Periodic updates to the RPV P/T limit curves are performed as necessary based upon the results of analyses of irradiated surveillance specimens removed from the vessel. In the future it is expected that hydrostatic and leak

  • testing may eventually be required with minimum reactor coolant temperatures exceeding 212°F. However, even with required minimum temperature requirements below 212°F, maintaining RCS temperature within a small band during the test can be impractical. Removal of the heat addition from recirculation pump operation and reactor core decay heat is coarsely controlled by control rod drive hydraulic system flow and reactor water cleanup system nonregenerative heat exchanger operation. Test conditions are focused on maintaining a steady state pressure, and tightly limited temperature control poses an unnecessary burden on the operator and may not be achievable in certain instances (ref. 6).

PNPS Basis Reference(s):

1. PNPS 8.M.2-6.1, "Reactor Pressure Readout"
2. Technical Specifications, Section 1.0, Definitions
3. PNPS 2.1.7 , "Vessel Heatup and Coo/down "
4. NEI 99-01 Definitions and Appendix C
5. Technical Specifications Section 3.7
6. Technical Specifications Section 3.14

NON-Q UALITY PNPS

~ RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 256 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 59 of 64 Unusual Event - CU4.1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 4 - Communications Initiating Condition : Loss of all onsite or offsite communications capabilities EAL:

CU4.1 Unusual Event Loss of all Table C-2 onsite (internal) communication methods affecting the ability to perform routine operations OR Loss of all Table C-2 offsite (external) communication methods affecting the ability to perform offsite notifications Table C-2 Communications Systems System Onsite Offsite (internal) (external )

Plant Telephone System (CENTREX) X X Wireless Telephone System X X Pilgrim Station Radio System X Plant Gaitronics System X Alternate Shutdown Communication X NRC-ENS Telephone, Direct Line X Satellite phones X

  • --=- Entergy .

PN PS EM ERGENCY PLAN ADM INISTRATIVE PROCEDURES N ON-QUALITY R ELATED PROCEDU RE REFERENCE USE EP-AD-601 Page 257 of Revision 9 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 60 of 64 Unusual Event - CU4.1 Mode Applicability:

4 - Cold Shutdown , 5 - Refuel , Defueled NEI 99-01 Basis:

The purpose of this IC and its associated EAL is to recognize a loss of communications capability that either defeats the plant operations staff's 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 10CFR50. 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.

PNPS Basis :

Onsite/offsite communications include one or more of the systems listed in Table C-2 (ref. 1, 2, 3, 4, 5). A description of the capabilities of each system is given in Section 4.0 of PNPS 2.2 .17, "Communications Systems" (ref. 2) .

This EAL is the cold condition equivalent of the hot condition EAL SU4.2.

PNPS Basis Reference(s) :

1. FSAR Section 10.15
2. PNPS 2.2.17 , "Communications Systems"
3. PNPS 2.4.57, "Loss of Public-Address System"
4. PNPS 8.A.13, "Plant Emergency Alarms And Radio Test"
5. EP-AD-413, "Emergency Communications Test"

\

NON- QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 258 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 61 of 64 Unusual Event - CU5.1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 5 - Inadvertent Criticality Initiating Condition: Inadvertent criticality EAL:

CUS.1 Unusual Event Unplanned sustained positive period observed on nuclear instrumentation Mode Applicability:

4 - Cold Shutdown, 5 - Refuel NEI 99-01 Basis:

This EAL addresses criticality events that occur in cold shutdown or refueling modes such as fuel misloading events. This EAL indicates a potential degradation of the level of safety of the plant, warranting a UE classification.

Escalation would be by Emergency Director judgment.

PNPS Basis :

Period meters Nl-750-4A, C, B, and Don Panel C905 identify this condition as well as annunciator "SRM PERIOD" (C905L-G9). Amber lights on Panel C905 illuminate when its SRM channel period is less than 20 seconds (seal in) (ref. 1, 2) .

PNPS Basis Reference(s):

1. PNPS 2.2.64, "Source Range Monitoring System"
2. PNPS 3.M.2-5.1, "Source Range Monitor Calibration Instruction"

N ON-Q UALITY PNPS

~ R ELATED EP-AD-601 Revision 9

-=- Entergy . EM ERG ENCY PLAN ADMINISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 259 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 62 of 64 Unusual Event - CU6 .1 Category: C - Cold Shutdown/Refueling System Malfunction Subcategory: 6 - Loss of DC Power Initiating Condition: Loss of requ ired DC power for 15 minutes or longer EAL:

CU6.1 Unusual Event

< 105V DC bus voltage indications on all Technical Specifications required 125V DC buses for ~ 15 min . (Note 3)

  • Note 3: The Emergency Director should not wait until the applicable time has elapsed but should declare the event as soon as it is determined that the condition will likely exceed the applicable time.

Mode Applicability:

4 - Cold Shutdown , 5 - Refuel NEI 99-01 Basis:

The purpose of this 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 in accordance with EAL CA3 .1.

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

NON-QUALI TY PNPS EP-AD-601 RELATED Revision 9

-=-Entergy EMERGENCY PLAN ADMINISTRATIVE PROCEDURES PROCE DURE REFERENCE USE Page 260 of 351 e l Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 63 of 64 Unusual Event - CU6.1 PNPS Basis:

The essential 125V DC power distribution is illustrated in Figure C-2.

The loss of required essential 125V DC power poses a significant threat to decay heat removal capability and reactor plant safety. Operability of safety-related equipment and safety system protective functions are severely degraded . The 125V DC system provides power to CSCS initiation logics, controllers, and indications. The 125V DC system also provides control power and tripping power for high voltage AC protective devices . If not restored within a short period of time , significant system and equipment failures may be imminent depending upon plant conditions at the time of the loss.

Annunciators "A 125V DC UNDERVOLTAGE" (C3RC-A7) and "B 125V DC UNDERVOL TAGE" (C3RC-B7) alarm at 124V DC (decreasing) and signal loss of Panel D16

  • and D17 , respectively.

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS7 .1.

PNPS Basis Reference(s):

1. FSAR Figure 8.6-1
2. PNPS 2.2 .14, "125V DC Battery Systems"
3. PNPS 5.3.11 , "Loss of Essential DC Bus 016 or 04 and 036"
4. PNPS 5.3.12, "Loss of Essential DC Bus 017 or 05 and 03 7"
5. PNPS ARP C3RC-A7
6. PNPS ARP C3RC-B7
  • -=- Entergy .

PNPS EMERGENCY PLAN ADMINISTRATIVE NON-QUALITY RELATED PROCEDURE

  • EP-AD-601 Revision 9 PROCEDURES REFERENCE USE Page 261 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.5 CATEGORY C, COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION Sheet 64 of 64 Unusual Event - CU6.1 Figure C-2 Essential 125V DC Power Distribution 480V MCC B-15 125VDC 460V r,. CC B-10 125 V OC 480V MCC B-14 CONTROi.. CONT ROL c:J..1' BATTERY B EIATTERYA 52-1s13 (52.1 021 D-1

~ D-2 C FUSE C 125 voe CHARGER 125 VDC~CKUP BATTERY A I BATIERY CHARGER l

200a C 0.11 0-29 200a C D- 14 j' D-16

(,k 72-162 _ _ _ _ _....,

f

'A'"1

-166 i2-16A AUTO TRANSFER SWITCH Y-10

~

MAJNT PANEL PANEL PANEL PANEL SPARE [).4 0-3 0 -5 D-7 RE EMERGENCY LIGHTING PANEL SOLENOID OPERATED p El PANEL VALVE PANEL D-36 DISTRIBUTION P ANEL 0-6 0 -:37 0-l ll 25L ll&tJJ&'.1

NON- QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=--Entergy . ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 262 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 1 of 65 Introduction Table F-1 lists the threshold conditions that define the loss and potential loss of the three fission product barriers (fuel clad, Reactor Coolant System , and Primary Containment). The table is structured so that each of the three barriers occupies adjacent columns . Each fission product barrier column is further divided into two columns ; one for loss thresholds and one for potential loss thresholds.

The first column of the table (to the left of the Fuel Clad Barrier Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

A. RPV Level B.

C.

D.

PC Pressure/Temperature Isolation Rad E. Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the categories.

The intersection of each row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss , the word "None" is entered in the cell.

Thresholds are assigned sequential numbers to facilitate referencing . If a cel l in Table F-1 contains more than one numbered threshold, each of the numbered thresholds , if exceeded ,

signifies a loss or potential loss of the barrier. It is not necessary to exceed all of the thresho lds in a category before declaring a barrier loss/potential loss.

Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier loss and potential loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers .

NON-QUALITY PNPS

~ RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=- Entergy ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 263 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss M ATRIX AND TECHNICAL BASES Sheet 2 of 65 When equipped with knowledge of plant conditions related to the fission product barriers, the EAL user first scans down the category column of Table F-1, locates the likely category, and then reads across the fission product barrier Loss and Potential Loss thresholds in that category to determine whether a threshold has been exceeded . If a threshold has not been exceeded, the EAL user proceeds to the next likely category and continues review of the thresholds in the new category If the EAL user determines that any threshold has been exceeded, by definition the barrier is lost or potentially lost - even if multiple thresholds in the same barrier column are exceeded ,

only that one barrier is lost or potentially lost. The EAL user must examine each of the three fissio n product barriers to determine whether other barrier thresholds in the category are lost or potentially lost. For example, if containment radiation is sufficiently high , a loss of the fuel clad and RCS barriers and a potential loss of the Primary Containment barrier can occur.

Barrier losses and potential losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, FA 1.1 , and FU1 .1 to determine the appropriate emergency classification.

In the remainder of this Attachment, the fuel clad barrier threshold bases appear first, followed by the RCS barrier, and finally the Primary Containment barrier threshold bases. In each barrier, the bases are given showing the "loss" threshold followed by the "potential loss" threshold beginning with Category A then sequentially to E .

PNPS NON-QUALITY RELATED EP-AD-601 Revision 9

~ EMERGENCY PLAN P ROCEDURE

-:::::.--Entergy ADMINISTRATIVE PROCEDURES REFERENCE USE Page 264 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 3 of 65 Table F-1 Fission Product Barrier Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Loss  : Potential Loss Loss Potential Loss Loss

Potential Loss A. RPV Level 1. Primary Containment Flooding 1

5. RPV level cannot be 7. RPV level cannot be restored 1

None None

22. SAG entry required required due to EITHER: restored and maintained and maintained > -125 in.

RPV water leve l ca nnot be > -125 in. or can not be or cannot be determined restored and maintained determined above -150 in. (MSCRWL}

OR RPV water leve l cannot be determined and core damage is occurring B. PC Pressu re 8. PC pressure > 2.2 psig 16.PC pressure rise followed by a rapid unexplained 23. Toru s bottom pressure > 60 psig and rising

/ Tem perat ure due to RCS leakage drop in PC pressure 24 . Deflagration concentrations exist inside PC None None None

17. PC pressure response not consistent with LOCA conditions 25. Torus water temperature and RPV pressure c:a nn ot be maintained below Heat Capacity Temperature
Limit (EOP- 11 Figure 2)

C. Iso lation 9. Release pathway exists 13. RCS leakage > 50 gpm inside the 18. Failure of any valve in any one line lo close outside primary containment drywell AND resulting from isolation failure Direct downstream pathway to the environment in an y of the following 14. Unisolable primary system exists after PC isolation signal (excluding normal process discharge outside primary system flowpaths from an containment 19. Intentional PC venting per EOPs unisolable system) : AND None None - Main steam line A valid entry condit ion to EOP-4 20. Unisolable primary system discharge outside PC None

- HPCI steam line exists due to Secondary resulting in Secondary Containment area radiation

- RC IC steam line Conta inment area radiation or or temperature above any Maximum Safe

- RWCU temperature above any Maximum Operating Value (EOP-4 , Table L)

- Feedwater Normal Operating Value {EOP-4 ,

Table H)

10. Emergency RPV depressurization req uired D. Rad 2. Drywell High Range Area ' 11 . Drywe ll High Range Area '  : 26. Drywell High Range Area Radiation Monitor (RIT-Radiation Monitor (RIT- Radiation Monitor (RIT- 1001 -606Aand 8 ) > 8,000 R/hr 1001-606A and 8 ) 1001 -606A and 8) OR

> 800 R/hr > 65 R/hr Torus High Range Area Radiation Monitor (RIT-OR OR 1001-607Aand B) > 500 R/hr Torus High Range Area Torus High Range Area Radiation Monitor (RIT- None Radiation Monitor (RIT- None Non, 1001-607A and 8 ) 1001-607A and 8)

> 50 R/hr > 4 R/hr

3. Primary coo lant activity

> 300 µCi/gm 1-131 dose equivalent  :  :

4. Any condition in the 6 . Any condition in the 12. Any condition in the 15. Any condition in the opinion of 21. Any condition in the opin ion of the Emergency 27. Any condition in the opinion of the Emergency E. Judgment opinion of the Emergency opinion of the opinion of the Emergency the Emergency Director that Director that indicates loss of the PC barrier Director that indicates potential loss of the PC Director that indicates loss Emergen cy Director Director that indicates loss indicates potential loss of the barrier of the Fuel Clad barrier that indicates potential of the RCS barrier RCS barrier loss of the Fuel Clad barrier i i

N ON-Q UALI TY PNPS

~ R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 265 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 4 of 65 Fuel Clad RPV Level - Loss Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

1. Primary Containment Flooding required due to EITHER of the following:
  • RPV water level cannot be restored and maintained above -150 in .

(MSCRWL)

RPV water level cannot be determined and core damage is occurring NEI 99-01 Basis:

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 .

N ON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROC EDURES REFERENCE USE Page 266 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 5 of 65 Fuel Clad RPV Level - Loss PNPS Basis :

Severe Accident Guidelines (SAGs) are entered from EOP-1 , EOP-2 , EOP-3 , EOP-16, and EOP-26 when Primary Containment flooding is required. Primary Containment flooding is required when core cooling is severely challenged. These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined . SAG entry is required when (ref. 1):

  • RPV water level cannot be restored and maintained above -150 in. (MSCRWL) (ref. 1).
  • RPV water level cannot be determined and core damage is occurring (ref. 2, 3).

The above EOP conditions represent a challenge to core cooling and are the minimum values to assure core cooling without further degradation of the clad .

This threshold is also a potential loss of the Primary Containment barrier (PC P-Loss A.22) .

Since SAG entry occurs after core uncovery has occurred , a loss of the RCS barrier exists (RCS Loss A.7). SAG entry, therefore , represents a loss of two barriers and a potential loss of a third, which requires a General Emergency classification .

PNPS Basis Reference(s):

1. EOP-1 , "RPV Control"
2. EOP-16, "RPV Flooding"
3. EOP-26 , "RPV Flooding, Failure-To-Scram "
4. EOP-2 , "RPV Control, Failure-To-Scram"
5. EOP-3 , "Primary Containment Control"
  • 1

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 267 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 6 of 65 Fuel Clad RPV Level - Potential Loss Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

5. RPV level cannot be restored and maintained> -125 in. or cannot be determined

This threshold is the same as the RCS barrier loss threshold and corresponds to the site-specific water level at the top of the active fuel. Thus, this threshold indicates a potential loss of the fuel clad barrier and a loss of RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

PNPS Basis:

An RPV level instrument reading of -125 in. indicates RPV level is at the top of active fuel (TAF) (ref. 1). When RPV level is at or above TAF, the core is completely submerged . Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened , the EOPs specify alternate, more extreme RPV level control measures in order to restore and maintain adequate core cooling . Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the fuel clad barrier .

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

--=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 268 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 7 of 65 Fuel Clad RPV Level - Potential Loss When RPV level cannot be determined , EOPs require entry to EOP-16, "RPV Flooding ", or EOP-26, "RPV Flooding, Failure-to-Scram". RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained (ref. 2, 3). When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted . The instructions in EOP-16/26 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram -failure events) (ref. 2, 3). If RPV water level cannot be determined with respect to the top of active fuel , a potential loss of the fuel clad barrier exists.

Note that EOP-2, "RPV Control, Failure -to-Scram ", may require intentionally lowering RPV

  • water level to TAF and controlling level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and TAF (ref. 4). Under these conditions , a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - ATWS/Criticality EALs.

If core uncovery is threatened, the EOPs specify alternate, more extreme RPV level control measures in order to restore and maintain adequate core cooling , including depressurization and restoration with low pressure pumps.

Determination of "restore and maintain" is based on the actions driven by Emergency Operating Procedures to restore level. The inability to reverse the RPV level lowering trend after lining up injection sources and injecting , including the use of low pressure systems following an Emergency Depressurization , would warrant classification .

PNPS Basis Reference(s):

1. EOP-1 , "RPV Control"
2. EOP-16, "RPV Flooding"
3. EOP-26 , "RPV Flooding, Failure-To-Scram"
4. EOP-2 , "RPV Control, Failure-To-Scram"

NON-Q UALI TY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 269 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 8 of 65 Fuel Clad PC Pressure/Temp - Loss Barrier: Fuel Clad Category: B. PC Pressure/Temperature Degradation Threat: Loss Threshold:

I None

a--=- Entergy PNPS NON-QUALITY RELATED EP-AD-601 Revision 9 EMERGENCY PLAN PROCEDURE ADMINISTRATIVE PROCEDURES REFERENCE USE Page 270 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 9 of 65 Fuel Clad PC Pressure/Temp - Potential Loss Barrier: Fuel Clad Category: B. PC Pressure/Temperature Degradation Threat: Potential Loss Threshold:

N ON -QUALITY PNPS R ELATED EP-AD-601 Revision 9 EM ERGENCY PLAN

-=-Entergy ADM INISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 271 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 10 of 65 Fuel Clad Isolation - Loss Barrier: Fuel Clad Category: C. Isolation Degradation Threat: Loss Threshold:

~I None _ _ _ _ _ _l

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGEN CY PLAN ADMINI STRATIVE P ROCEDURE PROCED URES REFERENCE USE Page 272 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 11 of 65 Fuel Clad Isolation - Potential Loss Barrier: Fuel Clad Category: C. Isolation Degradation Threat: Potential Loss Threshold:

None

N ON-QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 273 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 12 of 65 Fuel Clad Rad - Loss Barrier: Fuel Clad Category: D. Rad Degradation Threat: Loss Threshold:

2. Drywell High Range Area Radiation Monitor (RIT-1001-606A and B) > 800 R/hr OR Torus High Range Area Radiation Monitor (RIT-1001-607A and B) > 50 R/hr NEI 99-01 Basis:

This threshold reading is a value which indicates the release of reactor coolant with elevated activity indicative of fuel damage , into the drywell.

Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within Technical Specifications and are therefore indicative of fuel damage.

This value is higher than that specified for RCS barrier loss threshold. Thus, this threshold indicates a loss of both fuel clad barrier and RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

There is no potential loss threshold associated with this item .

NON- Q UALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 274 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 13 of 65 Fuel Clad Rad - Loss PNPS Basis:

The Drywell and Torus High Range Area Radiation Monitor readings indicate the release of reactor coolant into the Primary Containment with elevated activity indicative of fuel damage.

The readings were derived assuming (ref. 1):

  • The reactor has been shutdown for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
  • 2% fuel clad damage
  • No drywell sprays in operation A LOCA-depressurized system The instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory into the drywell atmosphere The monitor reading of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown provides a realistic time for a degraded fuel condition to develop in a fast-breaking accident, which results in conservative threshold values for releases to Primary Containment with less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown (ref. 1).

In order to reach this fuel clad barrier loss threshold , a loss of the RCS barrier has already occurred (see RCS Loss 0 .11 ). This threshold, therefore , represents at least a Site Area Emergency classification.

PNPS Basis Reference(s):

1. EP-IP-330, "Core Damage"

N ON-Q UALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADM INISTRATIVE P ROCEDU RE PROCEDURES REFERENCE USE Page 275 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 14 of 65 Fuel Clad Rad - Loss Barrier: Fuel Clad Category: D. Rad Degradation Threat: Loss Threshold:

3. Primary coolant activity> 300 µCi/gm 1-131 dose equivalent

The reactor coolant activity of 300 µCi/gm 1-131 equivalent is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the fuel clad barrier is considered lost.

There is no potential loss threshold associated with this item .

PNPS Basis:

None PNPS Basis Reference(s):

1. NEI 99-01 Revision 5

NON-QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 276 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 15 of 65 Fuel Clad Rad - Potential Loss Barrier: Fuel Clad Category: D. Rad Degradation Threat: Potential Loss Threshold:

NON-QUALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 277 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 16 of 65 Fuel Clad Judgment - Loss Barrier: Fuel Clad Category: E. Judgment Degradation Threat: Loss Threshold:

4. Any condition in the opinion of the Emergency Director that indicates loss of the fuel clad barrier

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the fuel clad barrier is lost. In addition , the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.

PNPS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining whether the fuel clad barrier is lost. Such a determination should include imminent barrier degradation , barrier monitoring capability , and dominant accident sequences .

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns ,

readings from portable instrumentation , and consideration of offsite monitoring results .

N ON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN

--=- Entergy ADMIN ISTRATIVE PROCE DURE PROCE DU RES REFERENCE USE Page 278 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 17 of 65 Fuel Clad Judgment - Loss

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs . The Emergency Director should be mindful of the loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

N ON-Q UALITY PNPS RELATED EP-AD-601 Revision 9 EM ERG ENCY PLAN

--=-Entergy . ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 279 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 18 of 65 Fuel Clad Judgment - Potential Loss Barrier: Fuel Clad Category: E. Judgment Degradation Threat: Potential Loss Threshold:

6. Any condition in the opinion of the Emergency Director that indicates potential loss of the fuel clad barrier

This threshold addresses any other factors that are to be used by the Emergency Director in determin ing whether the Fuel Clad barrier is potentially lost. In addition , the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered potentially lost.

PNPS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining whether the fuel clad barrier is potentially lost. Such a determination should include imminent barrier degradation , barrier monitoring capability , and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns ,

read ings from portable instrumentation , and consideration of offsite monitoring results .

NON- QUALITY PNPS RELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 280 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 19 of 65 Fuel Clad Judgment - Potential Loss

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

NON-QUALITY PNPS RE LATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 281 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 20 of 65 Reactor Coolant System RPV Level - Loss Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Loss Threshold:

7. RPV level cannot be restored and maintained> -125 in. or cannot be determined

The loss threshold for RPV water level corresponds to the level that is used in EOPs to indicate challenge of core cooling.

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

There is no potential loss threshold associated with this item.

PNPS Basis :

An RPV level instrument reading of -125 in. indicates RPV level is at the top of active fuel

{T AF) ( ref. 1). TAF is significantly lower than the normal operating RPV level control band . To reach this level , RPV inventory loss would have previously required isolation of the RCS and Primary Containment barriers and initiation of all ECCS . If RPV level cannot be maintained above TAF , ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition , a LOCA event is a loss of the RCS barrier .

N ON-QUALITY PNPS R ELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 282 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 21 of 65 Reactor Coolant System RPV Level - Loss When RPV level cannot be determined , EOPs require entry to EOP-16 , "RPV Flooding", or EOP-26, "RPV Flooding, Failure-to-Scram". RPV water level indication provides the primary means of knowing whether adequate core cooling is being maintained (ref. 2, 3). When all means of determining RPV water level are unavailable , the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP-16/26 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold the Minimum Steam Cooling Pressures (in scram-failure events) (ref. 2, 3).

If RPV water level cannot be determined with respect to the top of active fuel , a potential loss of the fuel clad barrier exists.

Note that EOP-2 , "RPV Control, Failure-to-Scram", may require intentionally lowering RPV

  • water level to TAF and controlling level between the Minimum Steam Cooling RPV Water Level (MSCRWL) and TAF (ref. 4). Under these conditions , a high-power ATWS event exists and requires at least a Site Area Emergency classification in accordance with the System Malfunction - ATWS/Criticality EALs.

If core uncovery is threatened , the EOPs specify alternate, more extreme RPV level control measures in order to restore and maintain adequate core cooling , including depressurization and restoration with low level pumps.

Determination of "restore and maintain" is based on the actions driven by Emergency Operating Procedures to restore level. The inability to reverse the RPV level lowering trend after lining up injection sources and injecting , including the use of low pressure systems following an Emergency Depressurization , would warrant classification .

PNPS Basis Reference(s):

1. EOP-1 , "RPV Control"
2. EOP-16 , "RPV Flooding"
3. EOP-26 , "RPV Flooding, Failure-To-Scram "
4. EOP-2 , "RPV Control, Failure-To-Scram"

NON-Q UALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-- Entergy ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 283 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 22 of 65 Reactor Coolant System RPV Level - Potential Loss Barrier: Reactor Coolant System Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

I None

N ON- Q UALITY PNPS R ELATED EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 284 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 23 of 65 Reactor Coolant System PC Pressure/Temp - Loss Barrier: Reactor Coolant System Category: B. PC Pressure/Temperature Degradation Threat: Loss Threshold:

8. PC pressure> 2.2 psig due to RCS leakage NEI 99-01 Basis:

The loss threshold for Primary Containment pressure is based on the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.

There is no potential loss threshold associated with this item .

PNPS Basis:

The drywell high pressure scram setpoint is an entry condition to EOP-1, "RPV Control", and EOP-3 , "Primary Containment Control" (ref. 1, 2, 3). Normal Primary Containment pressure control functions (e.g. , operation of drywell cool ing , SGTS , etc.) are specified in EOP -3 in advance of less desirable but more effective functions (e.g ., operation of drywell or torus sprays, etc.).

In the PNPS design basis, Primary Containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend . Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control Primary Containment vent/purge (ref. 4 ).

N ON-Q UALITY PNPS R ELATED EP-AD-601 Revision 9

""=- Entergy . EMERGENCY PLAN ADMIN ISTRATIVE P ROCEDURE PROCEDURES REFERENCE USE Page 285 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 24 of 65 Reactor Coolant System PC Pressure/Temp - Loss The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect Primary Containment pressure. PC pressure greater than 2.2 psig with corollary indications (drywell temperature , humidity, etc.)

should therefore be considered a loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 2.2 psig should not be considered an RCS barrier loss.

PNPS Basis Reference(s):

1. EOP-1 "RPV Control," Entry Condition
2. EOP-3 "Primary Containment Control," Entry Condition
3. PNPS Technical Specifications 3.1
4. PNPS 2.4.44 , "Loss of Drywe/1 Area Coolers"

NON- QUALITY PNPS RELATED EP-AD-601 Revision 9 EMERGENCY PLAN

-=-Entergy ADMINISTRATIVE PROCEDURE PROCEDURES REFERENCE USE Page 286 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TE CHNICAL BASES Sheet 25 of 65 Reactor Coolant System PC Pressure/Temp - Potential Loss Barrier: Reactor Coolant System Category: B. PC Pressure/Temperature Degradation Threat: Potential Loss Threshold:

N ON- Q UALI TY PNPS RE LATE D EP-AD-601 Revision 9

-=- Entergy . EMERGENCY PLAN ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 287 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 26 of 65 Reactor Coolant System Isolation - Loss Barrier: Reactor Coolant System Category: C. Isolation Degradation Threat: Loss Threshold:

9. Release pathway exists outside Primary Containment resulting from isolation failure in any of the following (excluding normal process system flow paths from an unisolable system):

- Main steam line

- HPCI steam line

- RCIC steam line

- RWCU

- Feedwater NEI 99-01 Basis:

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

Other large high-energy line breaks such as HPCI , 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 .

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The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside Primary Containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flow path exists, emergency declaration under this threshold would not be required. Similarly, if the emergency response requires the normal process flow of a system outside Primary Containment (e.g. , EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves) , the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss C.18) barriers and justifies declaration of a Site Area Emergency (i .e., loss or potential loss of any two barriers) .

Even though RWCU and Feedwater systems do not contain steam , they are included in the list because an unisolable break could result in the high pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS .

PNPS Basis Reference(s):

1. FSAR Sections 4.4 , 4.5, 4.6 , 4.11 (MSL)
2. FSAR Section 6.3 (HPCI)
3. FSAR Section 4.7 (RCIC)
4. FSAR Section 4.9 (RWCU)
5. FSAR Section 11.9 (FW)
6. PNPS 2.2 .21, "High Pressure Coolant Injection System (HPCI)"
7. PNPS 2.2.22, "Reactor Core Isolation Cooling System (RCIC)"
8. PNPS 2.2.83, "Reactor Cleanup System"
9. PNPS 2.2.92 , "Main Steam Line Isolation and Turbine Bypass Valves"

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10. Emergency RPV depressurization required

Plant symptoms requiring emergency RPV depressurization in accordance with PNPS 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.

PNPS Basis :

Plant symptoms requiring emergency RPV depressurization (RPV-ED) are specified in the EOPs (ref. 1, 2, 3, 4, 5, 6, 7, 8, 9, 10). If emergency RPV depressurization is required , the plant operators are directed to open SRVs and keep them open regardless of any subsequent radiological release rate (ref. 7, 8, 9, 10) .

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1. EOP-1 , "RPV Control"
2. EOP-2 , "RPV Control, Failure-to-Scram"
3. EOP-3 , "Primary Containment Control"
4. EOP-4 , "Secondary Containment Control"
5. EOP-5, "Radioactivity Release Control"
6. EOP-8 , "Steam Cooling" 7 EOP-16 , "RPV Flooding"
8. EOP-17 , "Emergency RPV Depressurization "
9. EOP-26 , "RPV Flooding, Failure-to-Scram"
10. EOP-27 , "Emergency RPV Oepressurization, Failure-to-Scram"

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113. RCS leakage > 50 GPM inside the drywell

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 primary system leak rate information is unavailable, other indicators of RCS leakage should be used.

PNPS Basis :

RCS leakage inside the drywell is normally determined by monitoring drywell equipment and floor drain sump pump-out rates . This method of monitoring leakage may be isolated as part of the drywell isolation and thus may be unavailable . If primary system leak rate information is unavailable, other indicators of RCS leakage should be used (ref. 1). Inventory loss events, such as a stuck open SRV, should not be considered when referring to "RCS leakage" because they are not indications of a break which could propagate .

PNPS Basis Reference(s):

1. PNPS 2.2.77 , "Drywe/1 Leak Detection Systems"

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14. Unisolable primary system discharge outside Primary Containment AND A valid entry condition to EOP-4 exists due to Secondary Containment area rad iation or temperature above any Maximum Normal Operating Value (EOP-4, Table H)

NEI 99-01 Basis:

Potential loss of RCS based on primary system leakage outside the Primary Containment is determined from site-specific temperature or area radiation Max Normal setpoints in the areas of the main steam line tunnel , main turbine generator, RCIC , HPCI , etc., which indicate a direct path from the RCS to areas outside Primary Containment.

The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage warrant an Alert classification. An unisolable leak which is indicated by a high alarm setpoint escalates to a Site Area Emergency when combined with containment barrier loss threshold C.20 (after a containment isolation) and a General Emergency when the fuel clad barrier criteria are also exceeded.

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The presence of elevated general area temperatures or radiation levels in the Secondary Containment may be indicative of unisolable primary system leakage outside the Primary Containment. The Maximum Normal Operating Values (Table F-2 below) define this RCS threshold because they signify the onset of abnormal system operation. When parameters reach this level , equipment failure or misoperation may be occurring . Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area . The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-4, Secondary Containment Control , Table H (ref. 1).

  • In general , multiple indications should be used to determine whether a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the Secondary Containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g ., room flooding , high area temperatures , reports of steam in the Secondary Containment, an unexpected rise in feedwater flow rate , or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the Secondary Containment. For example, a high RWCU area temperature may be indicative of increased ambient temperatures due to seasonal changes that simply indicate repositioning of ventilation dampers is needed . Although a Table H temperature has reached its Maximum Normal Operating Value , the Shift Manager determines the entry condition to EOP-4 to be invalid and need not execute EOP-4. For such conditions, this EAL threshold is not met.

PNPS Basis Reference(s):

1. EOP-4 , "Secondary Containment Control"

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NW equipment space/HPCI pump room - 17 ft 6 in. El.

CRD pump room - 17 ft 6 in. El.

RCIC pump room - 17 ft 6 in. El.

SE equipment space - 17 ft 6 in. El. Unexplainable CRD HCU west area - 23 ft El. Loose CRD HCU east area - 23 ft El. Surface Contamination RB west area - 51 ft El. or Radiation, RB east area - 51 ft El. procedure North Storage and laydown area - 74 ft 3 in. El. (H202 5.3.33 Fuel pool cooling pump/HX area - 74 ft 3 in . El SLC pump area - 91 ft 3 in. El.

Skimmer surge tank area - 91 ft 3 in. El.

TEMPERATURE VALUE (°F)

RWCU Holding Pump Area - 74 ft El. TE-1291-60C 105 RWCU Filter Area - 74 ft El. TE-1291 -60A 11 5 RWCU Backwash Tank Area - 51 ft El. TE-1291-60B 105 RWCU Pump "A" Room -51 ft El. TE-1291 -60D 105 RWCU Pump "B" Room - 51 ft El. TE-1291 -60E 105 RWCU Heat Exchanger Room - 51 ft El.

  • TE-1 291-60F 115 RCIC Piping Area - Torus Compt
  • TE-1360-23A 105 RCIC Turbine Area - Stairwell TE-1 360-23B 115 HPCI Piping Area - Torus Compt TE-2374A 105 HPCI Turbine Area - 17 ft El. TE-2374B 115 RWCU Piping Area - 36 ft El. Mezzanine TE-1291-60H 105 RCIC Tip Room - 23 ft El. TE-1360-23C 105 Main Steam Tunnel - 23 ft El. TE-261-22A 140 HPCI Piping Area - 23 ft El. ("B" RHR Valve Room ) TE-2374C 105 RHR "B" & *o* Pump Area - Stairwell TE-1001 -92A 115 RHR "A" & "C" Pump Area - 6 ft El. TE-1001 -92B 11 5 RHR "A' & *c* Pump Area - Pipewell TE-1001-92G 105 RWCU & RHR Valve Room - 23 ft El. ("A" RHR Viv Rm) TE-100 1-92F 105 RHR Fuel Pool Heat Exchanger Room - 74 ft El.
  • TE-100 1-92H 105 HP CI Compartment H&V Cooler (Panel C-61 B)
  • 24-TI-H-38/39 100 RHR A Quadrant (SE) H&V Cooler (Panel C-61A)
  • 24-TI-H-34/36 100 RHR B Quadrant (NW) H&V Cooler (Panel C-61A)
  • 24-TI-H-35/37 100 CRD Quadrant (NE) H&V Cooler (Panel C-6 1A) 24-TI-H-42/43 100 RCIC Quadrant (SW) H&V Cooler (Panel C-61A)
  • 24-TI-H-40/41 100
  • Readings available on Kaye Computer

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ATTACHMENT 9.6 FISSION PRODUCT BARRIER LOSS/POTENTIAL Loss MATRIX AND TECHNICAL BASES Sheet 34 of 65 Reactor Coolant System I Rad - Loss I Barrier: Reactor Coolant System Category: D. Rad Degradation Threat: Loss Thresho ld:

11. Drywell High Range Area Rad iation Monitor (RIT-1001-606A and B) > 65 R/hr OR Torus High Range Area Radiation Monitor (RIT-1001-607A and B) > 4 R/hr NEI 99-01 Basis:

The loss threshold reading is a value which indicates the release of reactor coolant to the Primary Containment.

This reading will be less than that specified for fuel clad barrier loss threshold 0.2. Thus, this threshold would be indicative of an RCS leak only. If the radiation monitor reading increased to that value specified by fuel clad barrier threshold, fuel damage would also be indicated.

There is no potential loss threshold associated with this item.

PNPS Basis:

The drywell threshold is based on the Technical Specifications maximum allowable coolant activity uniformly dispersed into the Primary Containment 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reactor shutdown . The corresponding torus radiation threshold is 0.5 R/hr. A value of 4 R/hr has been selected, however, to provide a readable on-scale indication. The Orywell and Torus High Range Area Radiation Monitor range is 1 to 1 E7 R/hr .

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1. EP-IP-330 , "Core Damage"

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I None

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12. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier NEI 99-01 Basis:

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

PNPS Basis :

The Emergency Director judgment threshold addresses any other factors relevant to determining whether the RCS barrier is lost. Such a determination should include imminent barrier degradation , barrier monitoring capability , and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to the recognition of the inability to reach safety acceptance criteria before completion of all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns ,

readings from portable instrumentation , and consideration of offsite monitoring results .

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  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

NON-QUALITY PNPS

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15. Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier NEI 99-01 Basis :

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

PNPS Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining whether the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation , barrier monitoring capability , and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to the inability to reach final safety acceptance criteria before completing all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns ,

readings from portable instrumentation , and consideration of offsite monitoring results .

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  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

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I None

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I 22 . SAG entry required

There is no loss threshold associated with this item .

The potential loss requirement for Primary Containment flooding indicates adequate core cooling cannot be established and maintained and that core melt is possible. Entry into Primary Containment flood ing procedures 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 increased potential for containment fai lure. In conjunction with reactor vessel water level "loss" thresholds in the fuel clad and RCS barrier columns , this threshold will result in the declaration of a General Emergency - loss of two barriers and the potential loss of a third .

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Severe Accident Guidelines (SAGs) are entered from EOP-1 , EOP-2 , EOP-3 , EOP-16, and EOP-26 when Primary Containment flooding is required . Primary Containment flooding is required when core cooling is severely challenged . These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined . SAG entry is required when (ref. 1):

  • RPV water level cannot be restored and maintained above -150 in . (MSCRWL) (ref. 1).
  • RPV water level cannot be restored and maintained at or above -175 in . (elevation of
  • the jet pump suction) and no core spray subsystem flow can be restored and maintained equal to or greater than 3,600 GPM (design core spray flow) (ref. 1).
  • RPV water level cannot be determined and core damage is occurring (ref. 2, 3).

The above EOP conditions, if not restored and maintained , represent a potential core melt sequence which could lead to RPV failure and increased potential for containment failure.

This threshold is also a loss of the fuel clad barrier (FC Loss A.1 ). Since SAG entry occurs after core uncovery has occurred, a loss of the RCS barrier exists (RCS Loss A.7) . SAG entry, therefore , represents a loss of two barriers and a potential loss of a third , wh ich requires a General Emergency classification .

PNPS Basis Reference(s):

1. EOP-1, "RPV Control"
2. EOP-16, "RPV Flooding"
3. EOP-26 , "RPV Flooding, Failure-to-Scram"
4. EOP-2 , "RPV Control, Failure-to-Scram"
5. EOP-3, "Primary Containment Control"

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16. PC pressure rise followed by a rapid unexplained drop in PC pressure

Rapid unexplained loss of pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase from a high energy line break indicates a loss of containment integrity. Primary Containment pressure should increase as a result of mass and energy release into containment from a LOCA. Thus, Primary Containment pressure not increasing under these conditions indicates a loss of containment integrity.

This indicator relies on operator recognition of an unexpected response for the condition and therefore does not have a specific value associated with it. The unexpected response is important because it is the indicator for a containment bypass condition .

PNPS Basis :

None PNPS Basis Reference(s):

None

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17. PC pressure response not consistent with LOCA cond itions NEI 99-01 Basis:

Primary Containment pressure should increase as a result of mass and energy release into containment from a LOCA.

This indicator relies on operator recognition of an unexpected response for the cond ition 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 .

PNPS Basis :

The calculated pressure and temperature responses of the Primary Containment are shown in FSAR Figures 5.2-1 through 5.2-6 (ref. 1, 2). These figures show that the maximum calculated drywell pressure is well below the design allowable pressure (ref. 1, 2, 3). Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response (ref. 2).

PNPS Basis Reference(s):

1. FSAR Figures 5.2-1 through 5.2-6
2. FSAR Table 5.2-1
3. PNPS-NE-07-00006 Rev. 0

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23 . Torus bottom pressure> 60 psig and rising

The potential loss pressure is based on the Primary Containment design pressure.

PNPS Basis:

When torus bottom pressure reaches the maximum allowable value (60 psig) (ref. 1 ), Primary Containment venting may be required even if offsite radioactivity release rate limits will be exceeded (ref. 2). The torus bottom pressure value of 60 psig is based on the Primary Containment design pressure as demonstrated in the PNPS accident analysis (ref. 1 ). If this threshold is exceeded , a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a potential loss of the Primary Containment barrier even if a containment breach has not occurred .

PNPS Basis Reference(s):

1. FSAR Table 5.2-1
2. EOP-3 , "Primary Containmen t Control"

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24. Deflagration concentrations exist inside PC NEI 99-01 Basis:

None PNPS Basis:

Deflagration (explosive) mixtures in the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAMGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term . Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving , long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction . A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore , of a potential threat to Primary Containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1).

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The H2/02 system monitors Primary Containment hydrogen (H2) and oxygen (02) gas concentrations via the following containment penetration sample points: X-29E , X-106A-b ,

  • X-228J , X-15E , X-50A-d , and X-228C. The H2/02 system returns its samples to Primary Containment via containment penetrations X-46F and X-228K.

The analyzer system consists of two separate units: an analyzer panel and a remote panel.

Analyzer Panels C172 and C173 are located close to Primary Containment on the 74'3" elevation of the Reactor Building. Each panel contains all essential components necessary to perform the required containment atmosphere analysis, indication , and alarm functions.

Recorders AR-1001-612A and AR-1001-612B and indicating meters are located in the Control Room in Panels C174 and C175. The backup Comsip H2/02 analyzers, C172 and C173 , are set to alarm at either 4% 02 or 4% H2 and are normally left in the "STANDBY" mode (ref. 3, 4, 5).

PNPS Basis Reference(s):

1. Plant Specific Technical Guidelines for EOPs and SAGs, Section PC/G
2. EOP-3 , "Primary Containment Control"
3. PNPS 2.2.120 , "Postaccident Monitoring Panel"
4. PNPS 2.2.133 , "H2/02 Analyzer and C19 Systems"
5. SAG-02 , "Containment and Radioactivity Release Control"

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25 . Torus water temperature and RPV pressure cannot be maintained below Heat Capacity Temperature Limit (EOP-11 Figure 2)

NEI 99-01 Basis:

  • 1 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 ;

OR

  • Suppression chamber pressure above Primary Containment Pressure Limit A while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment ven t.

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.

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The Heat Capacity Temperature Limit (HCTL) is the highest torus temperature from which emergency RPV depressurization will not raise torus pressure above the Primary Containment Pressure Limit (PCPL) 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 torus 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 the Primary Containment barrier. This threshold is met when EOP-3 , "Primary Containment Control",

Step TT-10 is reached (ref. 1).

  • PNPS Basis Reference(s):
1. EOP-3 , "Primary Containment Control"

NON-QUALITY PNPS

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18. Failure of any valve in any one line to close AND Direct downstream pathway to the environment exists after PC isolation signal NEI 99-01 Basis:

This threshold addresses incomplete containment isolation that allows direct release to the environment.

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

PNPS Basis:

This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of Primary Containment integrity.

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Examples include unisolable main steam line, HPCI steam line , or RCIC steam line breaks ;

unisolable RWCU system breaks; and unisolable containment atmosphere vent paths. If the main condenser is available with an unisolable main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored ,

however, and do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using the Category A, Abnormal Rad Release/Rad Effluent, EALs.

  • The existence of an in-line charcoal filter (SGTS) does not make a release path indirect since the filter is not effective at removing fission 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 . Since Secondary Containment is not one of the three EAL fission product barriers, a direct unisolable release into Secondary Containment should therefore be considered a downstream pathway to the environment.

The threshold is met if the breach is not isolable from the Control Room or an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation from the Control Room should be made prior to the emergency classification. If operator actions from the Control Room are successful , this threshold is not applicable. Credit is not given for operator actions taken in-plant (outside the Control Room) to isolate the breach .

EOP-3, "Primary Containment Control", Step P-7 may specify Primary Containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions , with a valid containment isolation signal, the Primary Containment barrier should be considered lost.

PNPS Basis Reference(s):

1. EOP-3 , "Primary Containment Control"

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119. Intentional PC venting per EOPs NEI 99-01 Basis:

The EOPs may direct containment isolation valve logic(s) to be intentionally bypassed regardless of radioactivity release rates. Under these conditions , with a valid containment isolation signal , the containment should also be considered lost if containment venting is actually performed.

Intentional venting of Primary Containment for Primary Containment pressure or combustible gas control in accordance with EOPs to the Secondary Containment and/or the environment is considered a loss of containment. Containment venting for pressure when not in an accident situation should not be considered.

PNPS Basis:

EOP-3 , "Primary Containment Control", Step P-7 may specify Primary Containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). The threshold is met when the operator begins venting the Primary Containment in accordance with EOP-3 , not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in EOP -3 Step P-1 to control drywell pressure below the drywell high pressure scram setpoint do not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below Technical Specifications LCO limits.

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1. EOP-3 , "Primary Containment Control"

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20. Unisolable primary system discharge outside PC resulting in Secondary Containment area radiation or temperature above any Maximum Safe Operating Value (EOP-4 ,

Table L)

NEI 99-01 Basis:

This loss threshold addresses the presence of area radiation or temperature Max Safe Operating Values indicating unisolable primary system leakage outside the Primary Containment after a containment isolation. The indicators should be confirmed to be caused by RCS leakage.

There is no potential loss threshold associated with this item .

PNPS Basis :

The Maximum Safe Operating Values define this Primary Containment barrier threshold because they are indicative of problems in the Secondary Containment that are spreading and pose a threat to achieving a safe plant shutdown . This threshold addresses problematic discharges outside Primary Containment that may not originate from a high-energy line break.

The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP-4, Secondary Containment Control , Table L (ref. 1) (Table F-3 below) .

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PNPS Basis Reference(s):

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NW equipment space/HP CI pump room - 17 ft 6 in. El. 1000 CRD pump room - 17 ft 6 in. El. 1000 RCIC pump room - 17 ft 6 in. El. 1000 SE equipment space - 17 ft 6 in . El. 1000 CRD HCU west area - 23 ft El. 1000 CRD HCU east area - 23 ft El. 1000 RB west area - 51 ft El. 1000 RB east area - 51 ft El. 1000 North Storage and laydown area - 74 ft 3 in . El. (H 202) 1000 Fuel pool cooling pump/HX area - 74 ft 3 in. El 1000 SLC pump area - 91 ft 3 in. El. 1000 Skimmer surge tank area - 91 ft 3 in. El. 1000 TEMPERATURE VALUE (°F)

(Temperature areas are sepa rated by dashed lines)

RWC U Holding Pump Area - 74 ft El. TE-1291-60C 120 RWCU Filter Area - 74 ft El. TE-1291-60A 130 RWCU Backwash Tank Area - 51 ft El. TE- 1291 -60B 214 RWCU Pump "A" Room - 51 ft El. TE-1291-600 213 RWCU Pump "B" Room - 51 ft El. TE-1291 -60E 213 RWCU Heat Exchanger Room - 51 ft El. TE-1291-60F 215 RC IC Piping Area - Torus Compt TE-1 360-23A 258 RCIC Turbine Area - Stairwell TE-1 360-23B 175 HP CI Piping Area - Torus Compt TE-2374A 258 HPCI Turbine Area - 17 ft El. TE-2374B 175 RWCU Piping Area - 36 ft El. Mezzanine TE-1 291 -60H 238 RCIC Tip Room - 23 ft El. TE-1360-23C 224 Main Steam Tunnel - 23 ft El. TE-261-22A 289 HP CI Piping Area - 23 ft El. (B" RHR Valve Room) TE-2374C 309 RHR "B" & "D" Pump Area - Stairwell TE-1001-92A 200 RHR "A" & "C" Pump Area - 6 ft El. TE-1001-92B 200 RHR "A" & "C" Pump Area - Pipewell TE-1001-92G 224 RWCU & RHR Valve Room - 23 ft El. ("A" RHR Viv Rm) TE-1001 -92F 251 RHR Fuel Pool Heat Exchanger Room - 74 ft El. TE-1001 -92H 120

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I None

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26 . Drywell High Range Area Radiation Monitor (RIT-1001-606A and B) > 8,000 R/hr OR Torus High Range Area Radiation Monitor (RIT-1001 -607A and B) > 500 R/hr NEI 99-01 Basis:

The potential loss threshold reading is a value that indicates significant fuel damage well in excess of that required for loss of RCS and fuel clad . A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladd ing allows rad ioactive material to be released from the core into the reactor coo lant.

Regard less of whether containment is challenged , this amount of activity in containment, if released , could have such severe consequences that it is prudent to treat th is as a potential loss of containment such that a General Emergency declaration is warranted . NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents,"

indicates that such conditions do not exist when the amount of clad damage is less than 20%.

There is no loss threshold associated with this item .

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The Drywell and Torus High Range Area Radiation Monitor readings indicate the release of reactor coolant into the drywell with elevated activity indicative of fuel damage. The reading was derived assuming (ref. 1):

  • The reactor has been shutdown for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
  • 20% fuel clad damage
  • No drywell sprays in operation A LOCA-depressurized system The instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory into the drywell atmosphere The monitor reading of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown provides a realistic time for a degraded fuel condition to develop in a fast-breaking accident, which results in conservative threshold values for releases to containment with less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown (ref. 1).

In order to reach this Primary Containment barrier potential loss threshold , a loss of the RCS barrier (RCS Loss 0.11) and a loss of the fuel clad barrier (FC Loss 0.2) have already occurred. This threshold , therefore , represents at a General Emergency classification .

PNPS Basis Reference(s):

1. EP-IP-330, "Core Damage"

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21 . Any condition in the opinion of the Emergency Director that indicates loss of the PC barrier

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the containment barrier is lost. In addition , the inability to monitor the barrier should also be incorporated in this threshold as a factor in Emergency Director judgment that the barrier may be considered lost.

The containment barrier should not be declared lost based on exceeding Technical Specifications action statement criteria unless there is an event in progress requiring mitigation by the containment barrier. When no event is in progress (loss or potential loss of either fuel clad and/or RCS) , the containment barrier status is addressed by Technical Specifications .

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The Emergency Director judgment threshold addresses any other factors relevant to determining whether the Primary Containment barrier is lost. Such a determination should include imminent barrier degradation , barrier monitoring capability , and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns ,

readings from portable instrumentation , and consideration of offsite monitoring results.

  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

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27. Any condition in the opinion of the Emergency Director that indicates potential loss of the PC barrier

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

The containment barrier should not be declared potentially lost based on exceeding Technical Specifications action statement criteria unless there is an event in progress requiring mitigation by the containment barrier. When no event is in progress (loss or potential loss of either fuel clad and/or RCS) , the containment barrier status is addressed by Technical Specifications .

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The Emergency Director judgment threshold addresses any other factors relevant to determining whether the Primary Containment barrier is potentially lost. Such a determination should include imminent barrier degradation , barrier monitoring capability, and dominant accident sequences.

  • Imminent barrier degradation exists if the degradation will likely occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on a projection of current safety system performance. The term "imminent" refers to recognition of the inability to reach safety acceptance criteria before completion of all checks.
  • Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation , and consideration of offsite monitoring results.
  • Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

The "Judgment" classifications should not be used if an applicable EAL classification has already been determined as included in the EAL chart.

PNPS Basis Reference(s):

None

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EAL:

EU1 .1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY .

  • Mode Applicability:

All NEI 99-01 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 .

The results of the ISFSI Safety Analysis Report (SAR) per NUREG 1536 or SAR referenced in the cask('s) Certificate of Compliance and the related NRC Safety Evaluation Report identify natural phenomena events and accident conditions that could potentially effect the CONFINEM ENT BOUNDARY. 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.) .

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An Unusual Event is categorized as an event of sufficient magnitude that the CONFINEMENT BOUNDARY of a cask loaded with spent fuel is damaged such that it is breached .

Confinement Boundary means the outline formed by the sealed , cylindrical enclosure of the Multi-Purpose Canister (MPC) shell welded to a solid baseplate, a lid welded around the top circumference of the shell wall , the port cover plates welded to the lid , and the closu re ring welded to the lid and MPC shell providing the redundant sealing .

HI-STO This diagram is also applicable fo r the MPC-68 PNPS Basis Reference(s):

1. Holtec International FSAR for the Hi-Storm 100 Cask System revision 9
2. EC 53194 - Dry Fuel Project- Design Basis Threat Evaluation (DST Blast Effects on Loaded HI-Storm)
3. EC28039 - Dry Fuel Storage Operations

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9.8 BACKGROUND

AND DISCUSSION Sheet 1 of 10 9.

8.1 BACKGROUND

(1] EALs are the plant-specific indications, conditions , or instrument readings that are utilized to classify emergency conditions defined in the PNPS Emergency Plan (ref. 2.1 [6].

In 1992, the NRC endorsed NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels", as an alternative to NUREG-0654 EAL guidance.

NEI 99-01 (NUMARC/NESP-007) Revision 5 represents the most recently accepted methodology. Enhancements over earlier revisions include:

(a) Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions .

  • (b)

(c)

Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls).

Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

(2] Using NEI 99-01 Revision 5, Pilgrim Station conducted an EAL implementation upgrade project that produced the EALs discussed herein .

9.8.2 FISSION PRODUCT BARRIERS

[1] Many of the EALs derived from the NEI methodology are fission product barrier based.

That is, the conditions that define the EALs are based upon loss or potential loss of one or more of the three fission product barriers. "Loss" and "potential loss" signify the relative damage and threat of damage to the barrier. "Loss" means the barrier no longer assures containment of radioactive materials ; "potential loss" infers an increased probability of barrier loss and decreased certainty of maintaining the barrier .

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9.8 BACKGROUND

AND DISCUSSION Sheet 2 of 10

[2] The primary fission product barriers are:

(a) A - Fuel Clad (FC): The fuel clad barrier consists of the zircalloy fuel bundle tubes that contain the fuel pellets.

(b) B - Reactor Coolant System (RCS): 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.

(c) C - Primary Containment (PC): The Primary Containment barrier includes the drywell, the wetwell (torus) , their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves .

9.8.3 EMERGENCY CLASSIFICATION BASED ON FISSION PRODUCT BARRIER DEGRADATION

[1] The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

(a) Unusual Event:

Any loss or any potential loss of Primary Containment (b) Alert:

Any loss or any potential loss of either fuel clad or RCS (c) Site Area Emergency:

Loss or potential loss of any two barriers (d) General Emergency:

Loss of any two barriers and loss or potential loss of third barrier

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9.8 BACKGROUND

AND DISCUSSION Sheet 3 of 10 9.8.4 EAL RELATIONSHIP TO EOPS

[1] Where possible, the EALs have been made consistent with and utilize the conditions defined in the PNPS Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification , they define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened . When these symptoms are clearly representative of one of the NEI Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

  • 9.8.5 SYMPTOM-BASED VS. EVENT-BASED APPROACH

[1] To the extent possible , the EALs are symptom -based. That is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate . Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized .

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9.8 BACKGROUND

AND DISCUSSION Sheet 4 of 10 9.8.6 EAL ORGANIZATION (1] The PNPS EAL scheme includes the following features:

(a) Division of the EAL set into three broad groups:

(1) EALs applicable under all plant operating modes - This group would be reviewed by the EAL user any time emergency classification is considered.

(2) EALs applicable only under hot operating modes - This group would only be reviewed by the EAL user when the plant is in Hot Sh utdown , Startup ,

or Run mode.

(3) EALs applicable only under cold operating modes - This group would only be reviewed by the EAL user when the plant is in Cold Shutdown , Refuel ,

or Defueled mode.

(b) The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL user for a given plant condition ,

reduces EAL user reading burden , and thereby speeds identification of the EAL that applies to the emergency.

(c) Within each of the three EAL groups described above, assignment of EALs to categories/subcategories - category and subcategory titles are selected to represent conditions that are operationally significant to the EAL user.

Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The proposed PNPS EAL categories/subcategories and their relationship to NEI Recognition Categories are listed below.

(d) The primary tool for determining the emergency classification level is the EAL wall chart. The user of the EAL wall chart may (but is not required to) consult the EAL Technical Bases in order to obtain additional information concerning the EALs under classification consideration . The user should consult Sections 9.8.7 and 9.8.8 and Attachments 9.1 through 9.7 of this document for such information.

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9.8 BACKGROUND

AND DISCUSSION Sheet 5 of 10 (e) EAL Groups, Categories, and Subcategories EAL Group/Category I EAL Subcategory Any Operating Mode:

A - Abnormal Rad Release/Rad Effluent 1 - Offsite Rad Conditions 2 - Onsite Rad Conditions & Spent Fuel Pool Events 3 - MCR/CAC Rad iation H - Hazards 1- Natural or Destructive Phenomena 2- Fire or Explosion 3- Hazardous Gas 4- Security 5- Control Room Evacuation 6- Judgment

  • E - IFSFI Hot Conditions :

S - System Malfunction None 1 - Loss of AC Power 2 - ATWS/Criticality 3 - Inability to Reach Shutdown Conditions 4 - Instrumentation/Communications 5 - Fuel Clad Degradation 6 - RCS Leakage 7 - Loss of DC Power F - Fission Product Barrier Degradation None Cold Conditions :

C - Cold Shutdown/Refuel System 1- Loss of AC Power Malfunction 2- RPV Level 3- RCS Temperature 4- Communications 5- Inadvertent Criticality 6- Loss of DC Power

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9.8 BACKGROUND

AND DISCUSSION Sheet 6 of 10 9.8.7 TECHNICAL BASES INFORMATION

[1] EAL technical bases are provided in Attachment 9.1 through 9. 7 for each EAL according to EAL group (Any, Hot, Cold) , EAL category (A, H, E, S, F, and C), and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL , the following information is provided:

(a) Category Letter & Title (b) Subcategory Number & Title (c) Initiating Condition (IC)

(d) Site-specific description of the generic IC given in NEI 99-01

[2] EAL Identifier (enclosed in rectangle)

(a) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

(1) First character (letter): Corresponds to the EAL category as described above (A, H, E, S, F, or C) .

(2) Second character (letter): The emergency classification (G , S, A , or U).

(3) Third character (number): Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1 ). If a category does not have a subcategory, th is character is assigned the number one (1 ).

(4) Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL , it is given the number one (1 ).

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9.8 BACKGROUND

AND DISCUSSION Sheet 7 of 10

[3] Classification (enclosed in rectangle) . These emergency classifications include:

(a) Unusual Event(U)

(b) Alert (A)

(c) Site Area Emergency (S)

(d) General Emergency (G)

[4] EAL (enclosed in rectangle)

(a) Exact wording of the EAL as it appears in the EAL classification matrix.

[5] Mode Applicability

  • (a) One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Run , 2 - Startup , 3 - Hot Shutdown, 4 - Cold Shutdown , 5 - Refuel , D - Defueled , All , or N/A - Not Applicable. (See Section 9.8 .8 for operating mode definitions.)

[6] Basis:

(a) A generic basis section provides a description of the rationale for the EAL as provided in NEI 99-01. This is followed by a plant-specific basis section that provides PNPS-relevant information concerning the EAL.

[7] PNPS Basis Reference(s):

(a) Site-specific source documentation from which the EAL is derived .

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9.8 BACKGROUND

AND DISCUSSION Sheet 8 of 10 9.8.8 OPERATING MODE APPLICABILITY

[1] There are six Operating Modes, as follows:

(a) Mode 1: Run (1) Reactor is critical and the mode switch is in RUN. In this mode, the reactor system pressure is at or above 785 psig and the Reactor Protection System is energized with APRM protection and RBM interlocks in service.

(b) Mode 2: Startup

( 1) The mode switch is in STARTUP . In this mode the reactor protection scram trip , initiated by main steam line isolation valve closure , is bypassed when reactor pressure is less than 600 psig , the low pressure main steam line isolation valve closure trip is bypassed , the Reactor Protection System is energized with IRM neutron monitoring system trips and control rod withdrawal interlocks in service.

(c) Mode 3: Hot Shutdown (1) The mode switch is in SHUTDOWN , no core alterations are being performed, and reactor coolant temperature is> 212°F.

(d) Mode 4: Cold Shutdown (1) The mode switch is in SHUTDOWN , no core alterations are being performed , and reactor coolant temperature is ~ 212°F .

(e) Mode 5: Refuel (1) The mode switch is in REFUEL and reactor coolant temperature is

~ 212°F. During normal reactor shutdowns, the mode switch may be placed in the REFUEL position to proceed with manual control rod insertion. Although the mode switch is in REFUEL position for this evolution , refuel mode applicability for the purpose of EAL classification is defined by reactor coolant temperature as well as mode switch position .

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9.8 BACKGROUND

AND DISCUSSION Sheet 9 of 10 (f) Mode DEF: Defueled (1) Reactor vessel contains no irradiated fuel (full core off-load during refueling or extended outage).

[2] The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action is initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made , the declaration shall be based on the mode that existed at the time the event occurred.

[3] 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.

  • 9.8.9 VALIDATION OF INDICATIONS, REPORTS, AND CONDITIONS

[1] All emergency classifications shall be based upon valid indications, reports, or conditions. An indication , report, or condition is considered to be valid when it is verified by 1) an instrument channel check or 2) indications on related or redundant indicators or 3) by direct observation by plant personnel such that doubt related to the indicator's operability, the condition's existence , or the report's accuracy is removed.

Implicit in this definition is the need for timely assessment.

9.8.10 PLANNED VS. UNPLANNED EVENTS

[1] Planned evolutions involve preplanning to address the lim itations imposed by the condition, the performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the site 's Technical Specifications. Activities which cause the site to operate beyond that allowed by the site 's Technical Specifications , planned or unplanned, may result in an EAL threshold being met or exceeded . Planned evolutions to test, manipulate , repair, perform maintenance or modifications to systems and equipment that result in an EAL value being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license.

However, these conditions may be subject to the reporting requirements of 10CFR50 .72 .

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AND DISCUSSION Sheet 10 of 10 9.8.11 CLASSIFYING TRANSIENT EVENTS

[1] 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 ind icated and terminate the emergency once assessment shows that there were no consequences from the event and other termination criteria are met.

[2] 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 .

[3] 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 longer exists. In these cases an emergency should not be declared .

[4] Reporting requirements of 10CFR50 .72 are applicable and the guidance of NUREG-1022, Event Reporting Guidelines 10CFR50.72 and 50 .73, should be applied .

9.8.12 IMMINENT EAL THRESHOLDS

[1] 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 . While this is particularly prudent at the higher emergency classes (the early classification may permit more effective implementation of protective measures), it is nonetheless applicable to all emergency classes .

9.8.13 TREATMENT OF MULTIPLE EVENTS

[1] 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 ; an Alert and a Site Area Emergency is a Site Area Emergency.

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"=- Entergy . ADMINISTRATIVE PROCEDU RE PROCEDURES REFERENCE USE Page 339 of 351 Emergency Action Level Technical Bases Document ATTACHMENT 9.9 ABBREVIATIONS/ACRONYMS Sheet 1 of 6 AC ..... ............ .... Alternating Current ADS .......... ....... . Automatic Depressurization System APRM ......... ...... Average Power Range Monitor ARI .......... .... ...... Alternate Rod Insertion associated instrumentation ATWS .... ... ........ Anticipated Transient Without Scram BIIT .................. . Boron Injection Initiation Temp BWR .......... .... ... Boiling Water Reactor CAC ... .... ..... ..... . Containment Atmospheric Control CAD ... ............... Containment Atmospheric Dilution COE ....... ..... ...... Committed Dose Equivalent

  • CFR ....... .. ........ . Code of Federal Regulations cps .... .......... ... ... Counts per Second CRD .. ................ Control Rod Drive CS ......... ... .. ... .... Core Spray CSCS ...... ... .. ... .. Core Standby Cooling System CST ... ............ .... Condensate Storage Tank DC ..... ....... ...... ... Direct Current Dem in ... ..... ....... Demineralizer DHRP ................ Decay Heat Removal Pressure DW ........ .. ... ....... Drywell DWS IL ..... ...... .. . Drywell Spray Initiation Limit EAL ........... .... ... . Emergency Action Level ECCS ...... ........ .. Emergency Core Cooling Systems ECL .. .... ...... ...... . Emergency Classification Level ED ....... ......... .... . Emergency Director El. .... .... ... .. ...... ... Elevation

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GE ...... ... ....... ... .. General Emergency GPM ..... .. ... .. .... .. Gallons Per Minute H& V ........ ... ....... Heating and Ventilation H2 ... .... .. ....... ..... Hydrogen HCTL ... .. .. .... .. ... Heat Capacity Temperature Limit HCU .......... ..... ... Hydraulic Control Unit HOO ..... .. ........... Headquarters (NRC) Operations Officer HPCI .. ... ..... ... .... High Pressure Coolant Injection hr ... ... .......... .. ..... Hour HX ..... ... ... .. .. .. .... Heat Exchanger IC .. ... ...... .. .. .. ... .. Initiating Condition IDLH .. .. ...... .... .... Immediately Dangerous to Life and Health

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IPEEE ........... .... Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI .... ... ...... .... Independent Spent Fuel Storage Installation Keff ...... ... ... ... .... Effective Neutron Multiplication Factor lb ... ..... ............ ... Pound(s)

LCO .................. Limiting Condition for Operation LER ................... Licensee Event Report LFL. .. ................. Lower Flammability Limit LI .... ...... ... .......... Level Indicator LOCA ................ Loss of Coolant Accident LPCI .. ... .... ... ...... Low Pressure Coolant Injection LPSI .................. Low Pressure Safety Injection LR ... .. ........... ..... Level Recorder LWR ...... ..... .. ..... Light Water Reactor MDRIR .... ....... ... Minimum Debris Retention Injection Rate MDSL ........... ... .. Minimum Debris Submergence Level min .... ... .... .. .... .. . Minimum MNSDHR .... .... .. Minimum Number of SRVs Required for Decay Heat Removal MNSRED ... ... .... Minimum Number of SRVs Required for Emergency Depressurization mR ..... ..... .......... Milliroentgen/Millirem (as appropriate to the context and the units of radiation dose measurement) mrem ...... .......... . Millirem MSCP ..... .......... Minimum Steam Cooling Pressure MSIV ............. ... . Main Steam Isolation Valve MSL ........ .......... Main Steam Line MW ....... .... ...... .. Megawatt

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PSP ................. .. Pressure Suppression Pressure PSTG ... ........... .. Plant Specific Technical Guidelines PWR .......... ....... Pressurized Water Reactor

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' .. ........... ... .. .... .. . Feet

" ........ .. .... .. ..... ... Inches

% .... .... ....... .. ...... Percent

& .... ................... Ampersand ("and")

~F .. ..... .... ....... ... . Degrees Fahrenheit

> ............... ...... ... Greater Than

< .. .. ...... ... ..... ... ... Less Than

~ ..... ...... ....... ...... Greater Than or Equal To

~ .. ........... .... ...... . Less Than or Equal To

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PNPS NEI 99-01 Example EAL IC EAL AU1 .1 AU1 1 AU1 .2 AU1 1 AU1.3 AU1 3

  • AU2.1 AU2 .2 AA1 .1 AA1 .2 AU2 AU2 AA1 AA1 1

2 1

1 AA1 .3 AA1 3 AA2 .1 AA2 2 AA2 .2 AA2 1 AA3 .1 AA3 1 AS1.1 AS1 1 AS1.2 AS1 2 AS1 .3 AS1 4 AG1.1 AG1 1 AG1.2 AG1 2 AG1.3 AG1 4

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1, 2 1

CA1.1 CA3 1 CA2.1 CA1 1, 2 CA3 .1 CA4 1, 2 CS2.1 CS1 1 CS2 .2 CS1 2 CS2.3 CS1 3 CG2 .1 CG1 1 CG2.2 CG1 2 EU1 .1 E-HU1 1 FU1 .1 FU1 1 FA1 .1 FA1 1

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  • HU1 .5 HU2.1 HU2.2 HU1 HU2 HU2 5

1 2

HU3.1 HU3 1 HU3.2 HU3 2 HU4.1 HU4 1, 2, 3 HU6.1 HU5 1 HA1.1 HA1 1 HA1.2 HA1 2 HA1.3 HA1 5 HA1.4 HA1 4 HA1 .5 HA1 3 HA1.6 HA1 6

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1 1, 2 HG6.1 HG2 1 SU1.1 SU1 1 SU2.1 SUS 1 SU3.1 SU2 1 SU4 .1 SU3 1 SU4.2 SU6 1, 2 SU5.1 SU4 1 SU5.2 SU4 2 SU6.1 SUS 1, 2 SA1.1 SAS 1 SA2.1 SA2 1

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  • SG2 .1 SG2 1

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AU1 .1 18 SU1 .1 141 FU1.1 192 AU1 .2 23 SU2.1 157 FA1 .1 193 AU1 .3 27 SU3.1 169 FS1.1 194 AU2.1 55 SU4.1 171 FG1.1 196 AU2.2 58 SU4.2 173 1 265 AA1.1 30 SU5.1 182 2 273 AA1.2 34 SU5.2 184 3 275 AA1.3 38 SU6.1 185 4 277 AA2.1 60 SA1.1 145 5 267 AA2.2 63 SA2.1 158 6 279 AA3.1 66 SA4.1 175 7 281 AS1.1 40 SS1 .1 149 8 284 AS1 .2 44 SS2.1 162 9 287 AS1.3 46 SS4.1 178 10 289 AG1 .1 48 SS7.1 187 11 295 AG1.2 AG1 .3 HU1.1 HU1 .2 HU1 .3 C ategorv H Hazar d s 51 53 70 72 74 SG1.1 SG2.1 Category C - Cold Shutdown/Refueling s;ystem Mal f unction CU1 .1 153 165 200 12 13 14 15 16 17 18 298 291 292 300 305 306 312 HU1.4 76 CU2 .1 208 19 314 HU1 .5 78 CU2 .2 211 20 316 HU2 .1 98 CU2 .3 215 21 323 HU2 .2 101 CU3.1 245 22 303 HU3.1 106 CU3.2 248 23 307 HU3.2 109 CU4.1 256 24 308 HU4.1 114 CU5.1 258 25 310 HU6.1 130 CU6.1 259 26 321 HA1 .1 81 CA1 .1 204 27 325 HA1.2 84 CA2.1 219 HA1 .3 87 CA3.1 252 HA1 .4 89 CS2.1 224 HA1.5 92 CS2.2 226 Category E - ISFSI HA1.6 95 CS2.3 228 (Independent Spent Fuel HA2.1 103 CG2.1 232 Storage Installation)

HA3.1 111 CG2.2 237 ! EU1 .1 327  !

HA4.1 118 HA5.1 127 I Unusual Event HA6.1 132 HS4.1 122 I Alert HS5.1 128 HS6.1 134 I Site Area Emergency HG4.1 125 HG6.1 136 I General Emergency

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Document Number Document Title EP-IP-100 Emergency Classification and Notification EP-IP-100.1 Emergency Action Levels (EALs)

EAL Wall Chart EN-EP-313 Offsite Dose Assesment using the Unified RASCAL Interface EP-IP-310 Offsite Monitoring Team Activation and

Response

EP-IP-330 Core Damage EP-IP-400 Protective Action Recommendations PNPS 5.3.14 Security Incidents PNPS 5.3 .14.1 Airborne Threat

ATTACHMENT 9.2 10 CFR 50.54(a)(3) SCREENING SHEET 1 OF 4 Procedure/Document Number: EP-A0-601 I Revision: 9 Equipment/Facility/Other: Pilgrim Nuclear Power Station

Title:

Emergency Action Level Technical Bases Document 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]):

This revision of the PNPS EAL Technical Bases Document includes the following changes :

1) Attachment 9.2, Sh eet 11 of 70, Unusual Event - HU 1.5, PNPS Basis:

From:

"As illustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21 '6" MSL and well below the flood level of +13'6" MSL."

To:

"As il lustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21'6" (+21 .5') MSL and well above the flood level of + 13'6" MSL."

2} Attachment 9.2, Sheet 29 of 70, Alert - HA 1.6, PNPS Basis:

From:

"As il lustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21 '6" MSL and well below the flood level of +13'6" MSL."

To:

"As illustrated in Figure H-1 (ref. 1, 2), ground level at the screenhouse is +21 '6" (+21 .5') MSL and well above the flood level of +13'6" MSL."

3) Attachment 9.1, Sheet 5 of 5' , Unusual Event - AU 1.1, PNPS Basis:

From :

"recorder 40-R R-1705-19" To:

"recorder 40-RR-1705-24".

Part II. Activity Previously Reviewed? DYES [81 NO Is this activity fully bounded by an NRC approved 10 CFR 50.90 submittal or 50.54(q)(3) Continue to Evaluation is next part Alert and Notification System Design Report? NOT required .

Enter If YES. identify bounding source document number/approval reference and justification ensure the basis for concluding the source document fully bounds the below and complete Part proposed change is documented below: VI.

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 activity.(Refer to EN-Ll-100)

APPLICABILITY CONCLUSION

[81 If there are no other controlling change processes, continue the 50.54(q)(3) Screening.

0 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 below.

0 One or more controlling change processes are selected and fully bounds all aspects of the activity. 50.54(q)(3)

Evaluation is NOT required . ldentifv controllinq chanqe processes below and complete Part VI.

CONTROLLING CHANGE PROCESSES 10 CFR 50.54(q)

Part IV. Editorial Change DYES [81 NO 50.54(q )(3J Evaluation Cominue to nex t Is this activity an editorial or typographical change such as formatting , paragraph is NOT required. Enter part numbering, spelling, or punctuation that does not change intent? justi fi cati on and Justification : continue 10 nex1 part or complele Part VJ as applicable.

EN-EP-305 REV 6

ATTACHMENT 9.2 10 CFR 50.54(a)(3) SCREENING SHEET20F4 Procedure/Document Number: EP-AD-601 I Revision: 9 Equipment/Facility/Other: Pilgrim Nuclear Power Station

Title:

Emergency Action Level Technical Bases Document Part V. Emergency Planning ElemenUFunction 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 NU REG-0654/FEMA REP-1 Section II?

1. Responsibility for emergency response is assigned. [1] 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]
3. The process ensures that on shift emergency response responsibi lities 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. [3] 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] D
8. Procedures for notification of State and local governmental agencies are capable of alerling 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] LJ
15. Adequate facilities are maintained to support emergency response. [8] LJ
16. Adequate equipment is maintained to support emergency response . (8] D
17. Methods, systems. and equipment for assessment of radioactive releases are in use. [9] D
18. A range of publ ic PA Rs is available for implementation during emergencies. (1 O} D
19. Evacuation time estimates for the population located in 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 OJ
20. A range of protective actions is available for plant emergency workers during emergencies, including 0 those for hostile action events.[1 OJ EN-EP-305 REV 6

ATTACHMENT9.2 10 CFR 50.54(0)(3) SCREENING SHEET3 OF4 Procedure/Document Number: EP-AD-601 I Revision: 9 Equipment/Facility/Other: PIigrim Nuclear Power Station

Title:

Emergency Action Level Technical Bases Document 21 . The resources for controlling radio logical exposures for emergency workers are established. (1 1J D

22. Arrangements are made for medical services for contam inated, injured individuals. (12) D
23. Plans for recovery and reentry are developed. (13) u
24. A drill and exercise program (including radiological , medical, health physics and other program D areas) is 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 critique process in order to identify weaknesses. (14)
26. Identified weaknesses are corrected . (14) D
27. Training is provided to emergency responders. (15) D
28. Responsibil ity for emergency plan development and review is established. (16) D
29. Planners responsible for emergency plan development and maintenance are properly trained . (16) D APPLICABILITY CONCLUSION

@ If 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.

0 If any Part V criteria are checked, complete Part VI and perfo rm a 50.54(q)(3) Evaluation.

BASIS FOR CONCLUSION

1) This change to the EAL Technical Bases Document corrects an inadvertent and inappropriate use of the term "below" versus "above" when referencing comparative water heights within Figure H-1, "Screenhouse Bay Water Levels". Figure H-1 is included as a visual aid within PNPS EALs HA1 .6 and HU1 .5 and provides a cross-cutting view of bay water levels applied to screenhouse elevations.

This figure clearly depicts a ground level of the screenhouse at +21 '6" MSL (Mean Sea Level) accompanied with a comparison reference to the Flood Level of + 13'6" MSL (Mean Sea Level) . It is apparent through this discussion that +21 '6" (above) MSL would be well above, not below, a flood level of+ 13'6" (above) MSL. The change to this procedure corrects the use of the word "below" by replacing this word w ith "above".

Further, in this same PNPS Basis discussion, application of description of elevations using the term "well above" is used correctly when referencing the bay flood level of + 16'0" of the Maximum Monitored Water Level (HA1 .6) and +13'6" Flood Level (HU1 .5) .

EALs HA1 .6 and HU1 .5 as contained within EP-IP-100.1, Emergency Action Levels, and this procedure, EP-AD-601, are not impacted or changed as a result of this correction to the PNPS Basis discussion. Discussion with PNPS Training, Nuclear Operations confirm that the EALs present plain and specific wording for recognition of entry and upgraded conditions, and that PNPS Operations staff have demonstrated clear understanding of the elevations involved with these EALs as demonstrated through successful drill completion . This is a minor legacy issue within the PNPS Basis discussion of the EAL Technical Bases Document, found to have originated in Revision o of the document and discovered upon Entergy Corporate review in support of Decommissioning activities.

2) The term "(+21 .5')" from Figure H-1 was added parenthetically to the PNPS Basis discussion to enhance recognition of equivalency with the existing term "+21 '6" which is used within the proximate tables and discussion.
3) This change to the EAL Technical Bases Document corrects a reference to an inappropriate recorder for indications of Reactor Building Ventilation effluent. Per GE Process Radiation Monitoring System electrical print drawing M1 U114-5, the applicable recorder is 40-RR-1705-24, not 40-RR-1705-19, on Panel C902.

EAL AU1 .1 as contained within EP-IP* 100.1, Emergency Action Levels, and this procedure, EP-AD-EN-EP-305 REV 6

ATIACHMENT 9.2 10 CFR 50.54(Q)(3) SCREENING SHEET40F 4 Procedure/Document Number: EP-AD-601 I Revision: 9 EquipmenUFacility/Other: Pilgrim Nuclear Power Station

Title:

Emergency Action Level Technical Bases Document 601, are not impacted or changed as a result of this correction to the referenced recorder within the PNPS Basis discussion .

This is a minor legacy issue within the PNPS Basis discussion of the EAL Technical Bases Document, found to have originated in Revision O of the document and discovered upon Entergy Corporate review in support of Decommissioning activities.

This activity does not change intent, facilities , equipment or processes for this procedure or affect any planning standard elements . This activity does not affect the PNPS Emergency Plan . No further evaluation is required for this activity.

Part VI. Signatures:

Preparer Name (Print} Da~

Karen Larson-Sullivan (Optional} Reviewer Name (Print) Date:

A((ti Reviewer Name (Print} Revie~r Signature Date:

Aaron Magee ;_., *.--:7 r:J{'-*:,

-< t.---:J -*

Nuclear EP Project Manager I /

/

Approver Name (Print) Date:

Donna Calabrese Manager, Emergency Planning or designee EN-EP-305 REV 6

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