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NEI, Draft G of NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 7 - Change Summary
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Site:	Nuclear Energy Institute
Issue date:	07/31/2022
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To:	; Office of Nuclear Reactor Regulation
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Page 1 of 52 NEI 99-01 Revision 7 Change Summary July 2022

Page 2 of 52 This document summarizes the changes made in NEI 99-01, Revision 7.

NEI 99-01 Section NEI 99-01 Rev. 7 Change Summary Executive Summary Made editorial changes to improve clarity and readability. There were no intent changes.

1. Regulatory Background Deleted the Permanently Defueled Station section since the generic defueled ICs/EALs were removed from NEI 99-01. The new location for this EAL guidance will be DG-1346, Emergency Planning for Decommissioning Nuclear Power Reactors [proposed new Regulatory Guide 1.235]. In the meantime, licensees can continue to use the NRC-endorsed guidance in NEI 99-01, Revision 6, to develop EALs for a permanently defueled station.

Added a section on Immediate Notification Requirements per 10 CFR 50.72 to promote better awareness of the relationship between non-emergency notification requirements and emergency declarations.

Updated the information in the Spent Fuel Pool Monitoring Instrumentation section (e.g., Order EA-12-051 was replaced with 10 CFR 50.155). There were no intent changes.

Added section on Decommissioning Facility. The information reflects comments from both the NRC staff and the NEI Decommissioning Working Group. The guidance is aligned with NRC-approved License Amendment Requests related to EAL changes at decommissioning facilities.

Updated references to documents (e.g., added new ones, removed old ones, etc.).

Made editorial changes to improve clarity and readability.

2. Key Terminology Used in NEI 99-01 Added guidance to section 2.4, Fission Product Barrier Threshold, to better explain the relationship between the FPB thresholds and the radiological release EALs in Recognition Category A.

Updated references to documents (e.g., added new ones, removed old ones, etc.).

Made editorial changes to improve clarity and readability.

Page 3 of 52 NEI 99-01 Section NEI 99-01 Rev. 7 Change Summary

3. Design of the NEI 99-01 Emergency Classification Scheme Removed the ECL attributes (Section 3.1.1 through 3.1.4) as this information is no longer needed by the industry.

Removed discussion of a Station Blackout based on the change to IC SG1 (i.e., the SBO coping time is no longer considered in the EAL).

Deleted reference to Permanently Defueled Station EALs since the generic defueled ICs/EALs were removed from NEI 99-01.

The new location will be DG-1346, Emergency Planning for Decommissioning Nuclear Power Reactors [proposed new Regulatory Guide 1.235]. In the meantime, licensees can continue to use the NRC-endorsed guidance in NEI 99-01, Revision 6, to develop EALs for a permanently defueled station.

Added several statements to help licensees better understand NRC staff expectations concerning the content of a scheme conversion LAR.

Updated references to documents (e.g., added new ones, removed old ones, etc.).

Made editorial changes to improve clarity and readability.

4. Site-Specific Scheme Development Guidance Added several statements to help licensees better understand NRC staff expectations concerning the content of a scheme conversion LAR.

Revised section 4.3, Instrumentation Used in EALs, to provide more detail and incorporate operating experience (e.g., from EP findings). The changes also incorporated information from EPFAQ 2015-12.

Updated references to documents (e.g., added new ones, removed old ones, etc.).

Made editorial changes to improve clarity and readability.

5. Guidance on Making Emergency Classifications Replaced a reference to NRC NSIR/DPR-ISG-01 with text from the document.

Deleted a paragraph with guidance on not waiting to declare since this information appears in the Notes of the appropriate EALs (i.e., it was duplicative information).

Deleted table in section 5.5, Emergency Classification Level Downgrading and Termination, based on feedback that the information was not useful.

Deleted section 5.7, Classification of Short-Lived Events, based on feedback that the information was potentially confusing. The salient points are addressed in section 5.6, Classification of Transient Conditions.

Page 4 of 52 NEI 99-01 Section NEI 99-01 Rev. 7 Change Summary Revised section 5.8, Retraction of the Notification of an Emergency Declaration, to provide better guidance and address operating experience (e.g., ROP FAQ 21-02).

Made editorial changes to improve clarity and readability.

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 5 of 52 The table below summarizes the changes made to the Initiating Conditions and Emergency Action Levels in all Recognition Categories. As a general statement, the changes also included updating references to supporting documents (e.g., added new ones, removed old ones, etc.)

where needed, and making editorial changes to improve clarity and readability. Also, conforming changes supporting the addition, revision or deletion of an IC or EAL were made where necessary (e.g., references in one IC to another IC that was relocated or deleted were changed as appropriate).

Due to the width of the table columns and table formatting constraints, the appearance of an EAL (e.g., indentation) in this document may differ slightly from the appearance of the corresponding EAL in Revision 7.

Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis IC AU1 EAL #1 EAL #2 EAL #3 Release of gaseous or liquid radioactivity greater than 2 times the (site-specific effluent release controlling document) limits for 60 minutes or longer.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered events present a very low safety risk to the public. Activation of the site emergency plan and ERO mobilization would not be necessary to effectively respond to the event. Sites have sufficient procedures and capabilities to respond to this condition without the declaration of an emergency (e.g., use of Radiation Protection and Chemistry resources for locating and assessing airborne or waterborne releases). Depending on event-specific conditions, some plant response actions may be required by Technical Specifications and the site will make a report to the NRC in accordance with the requirements in 10 CFR Part 20 and/or 10 CFR 50.72. Finally, IC AA1 appropriately bounds releases that begin to present some elevated risk to the public (i.e., an airborne release with offsite consequences at or above 1% of the EPA PAG).

IC AU2 EAL #1 (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

(site-specific level indications).

IC AU2 EAL #1 (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

(site-specific level indications).

No change.

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 6 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis AND

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.

(site-specific list of area radiation monitors)

AND

b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.

(site-specific list of area radiation monitors)

N/A N/A IC AU3 EAL #1 EAL #2 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown New IC. See discussion below on NEI 99-01 Rev. 6 IC AA3 for change description and basis.

IC AA1 EAL #1 (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

(site-specific monitor list and threshold values)

N/A None - deleted.

This EAL was deleted because it may lead to an inappropriate emergency classification. The EAL values were calculated using assumed source terms and assumed meteorological conditions (affecting plume transport and dispersion). The assumed source terms and meteorological conditions will likely be different than those present during an actual event, perhaps significantly so. The preferred approach is to perform a dose assessment at the time of the event using actual effluent monitor readings and meteorological conditions; this approach will yield the emergency classification most reflective of the actual plant conditions. Sites maintain the capability to perform a dose assessment at all times (i.e., both on-shift and when the TSC and EOF activated). With respect to the use of averaged meteorological data from a plant computer, the differences between any two consecutive data sets (e.g., 15-minute averages delivered on the quarter hour) would not be

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 7 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis significant; therefore, performing an initial dose projection using the immediately preceding meteorological data set, if necessary, is not expected to meaningfully impact on the accuracy of the results.

A developer note was added that directs developers to verify that the emergency response facilities responsible for performing dose projections, including the Control Room, have a reliable dose assessment capability (i.e., primary and backup). Beyond this guidance, it is important to note that for events more likely to result in releases approaching or exceeding the EPA PAGs (i.e., involving challenges to multiple fission product barriers), plant indications are available to support timely and accurate emergency classifications should a dose assessment capability be lost. In addition, survey data from onsite and/or offsite monitoring teams would also be available to support emergency classification decision-making.

IC AA1 EAL #2 (2) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point).

IC AA1 EAL #1 (1) Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-specific dose receptor point).

Renumbered EAL based on the change discussed above.

Included discussion in Developer Notes concerning guidance in the 2017 EPA PAG Manual (EPA-400/R-17/001, PAG Manual:

Protective Action Guides and Planning Guidance for Radiological Incidents) per EPFAQ 2017-01.

IC AA1 EAL #3 (3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem N/A None - deleted.

This EAL is unnecessary as it is bounded by other EALs. Given the effluent dilution and dispersion that could reasonably be expected to occur between the source of the liquid (e.g., a tank) and the site boundary, it is highly unlikely that the specified doses could be reached. To do so would require a source term that is greater than that typically available during

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 8 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis thyroid CDE at or beyond (site-specific dose receptor point) for one hour of exposure.

normal operations (e.g., need some level of fuel defects or cladding failure). If a higher source term were present, then another EAL would already be met (e.g., a potentially lost or lost fission product barrier). In addition, an event covered by the EAL would generally be reported to the NRC as required by 10 CFR 50.72(b)(2)(xi). Finally, this type of event would not impact the ability of the site to implement the Emergency Plan or Security Plan, or require ERO mobilization or offsite support to address. It is also noted that State and local public safety and environmental officials, upon being notified of a spill, would take actions to minimize the risk to the public.

IC AA1 EAL #4 (4) Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

IC AA1 EAL #2 (2), Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 10 mR/hr are expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

Renumbered EAL based on the change discussed above.

Included discussion in Developer Notes concerning guidance in the 2017 EPA PAG Manual (EPA-400/R-17/001, PAG Manual:

Protective Action Guides and Planning Guidance for Radiological Incidents) per EPFAQ 2017-01.

IC AA2 EAL #1 Significant lowering of water level above, or damage to, irradiated fuel.

No change to IC or EALs, but expanded the guidance in the Developer Notes pertaining to instrumentation that requires manual actions to place in service.

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 9 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis EAL #2 EAL #3 EAL #2 EAL #3 IC AA3 EAL #1 EAL #2 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

IC AU3 EAL #1 EAL #2 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

This IC and the EALs were relocated from an Alert level to an Unusual Event level; no changes were made to the IC or EAL wording. The change was made based on a reassessment of the potential impact of the event and associated operating experience. Sites have plans and resources for responding to off-normal radiological conditions (e.g., those needed to meet NRC requirements). A response to off-normal radiological conditions does not require a full activation of the site ERO, which would occur following an Alert declaration. The declaration of an Unusual Event will ensure that key ERO managers are made aware of the event and available to support the response if needed. Should the event have operational consequences, or lead to more significant radiological consequences, enough to warrant an Alert or higher classification, then the emergency declaration would be based on another IC.

IC AS1 EAL #1 (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

(site-specific monitor list and threshold values)

N/A None - deleted.

This EAL was deleted because it may lead to an inappropriate emergency classification. The EAL values were calculated using assumed source terms and assumed meteorological conditions (affecting plume transport and dispersion). The assumed source terms and meteorological conditions will likely be different than those present during an actual event, perhaps significantly so. The preferred approach is to perform a dose assessment at the time of the event using actual effluent monitor readings and meteorological conditions; this approach will yield the emergency classification most reflective of the actual plant conditions. Sites maintain the capability to perform a dose assessment at all times (i.e., both on-shift and

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 10 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis when the TSC and EOF activated). With respect to the use of averaged meteorological data from a plant computer, the differences between any two consecutive data sets (e.g., 15-minute averages delivered on the quarter hour) would not be significant; therefore, performing an initial dose projection using the immediately preceding meteorological data set, if necessary, is not expected to meaningfully impact on the accuracy of the results. Finally, it should be noted that NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, indicates that no sequences resulted in a large early release, even considering external events and unsuccessful mitigation. This is a result of research conducted over the last several decades that has shown that phenomena earlier believed to lead to a large early release are of extremely low probability or not physically feasible.

A developer note was added that directs developers to verify that the emergency response facilities responsible for performing dose projections, including the Control Room, have a reliable dose assessment capability (i.e., primary and backup). Beyond this guidance, it is important to note that for events more likely to result in releases approaching or exceeding the EPA PAGs (i.e., involving challenges to multiple fission product barriers), plant indications are available to support timely and accurate emergency classifications should a dose assessment capability be lost. In addition, survey data from onsite and/or offsite monitoring teams would also be available to support emergency classification decision-making.

IC AS1 EAL #2 (2) Dose assessment using actual meteorology indicates doses greater IC AS1 EAL #1 (1) Dose assessment using actual meteorology indicates doses greater Renumbered EALs based on the change discussed above.

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 11 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis EAL #3 than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site-specific dose receptor point).

(3) Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

EAL #2 than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site-specific dose receptor point).

(2) Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 100 mR/hr are expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.

Included discussion in Developer Notes concerning guidance in the 2017 EPA PAG Manual (EPA-400/R-17/001, PAG Manual:

Protective Action Guides and Planning Guidance for Radiological Incidents) per EPFAQ 2017-01.

IC AS2 EAL #1 Spent fuel pool level at (site-specific Level 3 description).

No change to IC or EAL, but expanded the guidance in the Developer Notes pertaining to instrumentation that requires manual actions to place in service.

IC AG1 EAL #1 (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:

(site-specific monitor list and threshold values)

N/A None - deleted.

This EAL was deleted because it may lead to an inappropriate emergency classification. The EAL values were calculated using assumed source terms and assumed meteorological conditions (affecting plume transport and dispersion). The assumed source terms and meteorological conditions will likely be different than those present during an actual event, perhaps significantly so. The preferred approach is to perform a dose assessment at the time of the event using actual effluent monitor readings and meteorological conditions; this approach

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 12 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis will yield the emergency classification most reflective of the actual plant conditions. Sites maintain the capability to perform a dose assessment at all times (i.e., both on-shift and when the TSC and EOF activated). With respect to the use of averaged meteorological data from a plant computer, the differences between any two consecutive data sets (e.g., 15-minute averages delivered on the quarter hour) would not be significant; therefore, performing an initial dose projection using the immediately preceding meteorological data set, if necessary, is not expected to meaningfully impact on the accuracy of the results. Finally, it should be noted that NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, indicates that no sequences resulted in a large early release, even considering external events and unsuccessful mitigation. This is a result of research conducted over the last several decades that has shown that phenomena earlier believed to lead to a large early release are of extremely low probability or not physically feasible.

A developer note was added that directs developers to verify that the emergency response facilities responsible for performing dose projections, including the Control Room, have a reliable dose assessment capability (i.e., primary and backup). Beyond this guidance, it is important to note that for events more likely to result in releases approaching or exceeding the EPA PAGs (i.e., involving challenges to multiple fission product barriers), plant indications are available to support timely and accurate emergency classifications should a dose assessment capability be lost. In addition, survey data from onsite and/or offsite monitoring teams would also be available to support emergency classification decision-making.

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 13 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis IC AG1 EAL #2 EAL #3 (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (site-specific dose receptor point).

(3) Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

IC AG1 EAL #1 EAL #2 (1) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (site-specific dose receptor point).

(2) Field survey results indicate EITHER of the following at or beyond (site-specific dose receptor point):

Closed window dose rates greater than 1,000 mR/hr are expected to continue for 60 minutes or longer.

Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.

Renumbered EAL based on the change discussed above.

Added text to the Basis to better explain the relationship between IC AG1 and IC FG1.

Included discussion in Developer Notes concerning guidance in the 2017 EPA PAG Manual (EPA-400/R-17/001, PAG Manual:

Protective Action Guides and Planning Guidance for Radiological Incidents) per EPFAQ 2017-01.

IC AG2 EAL #1 Spent fuel pool level cannot be restored to at least (site-specific Level 3 description) for 60 minutes or longer.

No change to IC or EAL, but expanded the guidance in the Developer Notes pertaining to instrumentation that requires manual actions to place in service.

IC CU1 EAL #1 EAL #2 UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV

[BWR]) inventory for 15 minutes or longer.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered events present a very low safety risk to the public - the plant is in a cold condition (RCS 200°F) with significant water volumes in the RCS/RPV or available for addition. Further, activation of the site emergency plan and ERO mobilization would not be

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 14 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis necessary to effectively respond to the event. During Cold Shutdown and Refueling modes, stations typically have a large contingent of operations and technical staff onsite 24/7 to work the outage; the ready availability of this staff ensures a prompt response. If the event resulted in a significant level drop or protracted loss of level indication, then it would be classified as an Alert under IC CA1, Loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory. Depending on event circumstances, it may also be reported to the NRC in accordance with 10 CFR 50.72.

IC CU2 EAL #1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered event presents a very low safety risk to the public since the plant is in a cold condition (RCS 200°F). The event would be addressed by the requirements in plant Technical Specifications (e.g., immediately restore another required power source to OPERABLE status). Further, activation of the site emergency plan and ERO mobilization would not be necessary to effectively respond to the event. During Cold Shutdown and Refueling modes, stations typically have a large contingent of operations and technical staff onsite 24/7 to work the outage; the ready availability of this staff ensures a prompt response. If the event resulted in a total loss of AC power, then it would be classified as an Alert under IC CA2, Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Depending on event circumstances, it may also be reported to the NRC in accordance with 10 CFR 50.72.

IC CU3 EAL #1 (1) UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification N/A None - deleted.

This EAL is unnecessary as the covered event presents a very low safety risk to the public - although the cold shutdown temperature limit would be exceeded, bulk boiling of the RCS is not imminent. Activation of the site emergency plan and ERO

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 15 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis cold shutdown temperature limit).

mobilization would not be necessary to effectively respond to the event. During Cold Shutdown and Refueling modes, stations typically have a large contingent of operations and technical staff onsite 24/7 to work the outage; the ready availability of this staff ensures a prompt response. If the event persisted for greater than a time period specified in Table CA3-1, then it would be classified as an Alert under IC CA3, Inability to maintain the plant in cold shutdown. Depending on event circumstances, it may also be reported to the NRC in accordance with 10 CFR 50.72.

IC CU3 EAL #2 (2) Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV

[BWR]) level indications for 15 minutes or longer.

IC CU3 EAL #1 (1) Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV

[BWR]) level indications for 15 minutes or longer.

Renumbered EAL based on the change discussed above.

IC CU4 EAL #1 Loss of Vital DC power for 15 minutes or longer.

No change to IC or EAL. Deleted Developer Note on battery voltage - information was judged to be unnecessary since site-specific values should be considered.

IC CU5 EAL #1 EAL #2 EAL #3 Loss of all onsite or offsite communications capabilities.

No change to IC or EAL. Added Developer Note guidance to address operating experience with electronic/internet-based notification methods (e.g., ROP FAQ 20-04).

N/A N/A IC CU6 Internal flooding affecting a SAFETY SYSTEM component required for the current operating mode.

This IC is the relocated EAL #2 from IC HU3, which was replaced with a new IC and EAL.

IC CA1 EAL #1 (1) Loss of (reactor vessel/RCS

[PWR] or RPV [BWR])

inventory as indicated by IC CA1 EAL #1 (1) Loss of (reactor vessel/RCS

[PWR] or RPV [BWR])

inventory as indicated by Added a note and basis information to clarify classification expectations if the point of the leakage is above the vessel

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 16 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis EAL #2 level less than (site-specific level).

(2) a. (Reactor vessel/RCS

[PWR] or RPV [BWR])

level cannot be monitored for 15 minutes or longer AND

b. UNPLANNED increase in (site-specific sump and/or tank) levels due to a loss of (reactor vessel/RCS [PWR] or RPV

[BWR]) inventory.

EAL #2 level less than (site-specific level).

(2) a. (Reactor vessel/RCS

[PWR] or RPV [BWR]) level cannot be (monitored

[PWR] or determined

[BWR]) for 30 minutes or longer.

AND

b. EITHER of the following:

1. UNPLANNED increase in (site-specific sump and/or tank) levels due to a loss of (reactor vessel/RCS [PWR] or RPV

[BWR]) inventory.

OR

2. Visual observation of UNISOLABLE RCS leakage.

flange.

In EAL #2.a, added a provision for BWRs to use the term determined per EPFAQ 2019-04.

Changed the 15 minutes criterion in EAL #2.a to 30 minutes to align the EAL more closely with the definition of an Alert (i.e., it was determined that 15 minutes was not long enough to say there was a potential substantial reduction in the level of plant safety). This is appropriate given the RCS conditions during shutdown, available large water volumes, large on-site staff during outages, and bounding for escalation provided by IC CS1.

Added EAL statement (2).b.2 since visual observation could also identify unisolable leakage.

IC CA2 EAL #1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

CA2 EAL #1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

No change to IC or EAL. Added a note and basis information on credit for non-safety-related power sources; this addition addressed EPFAQ 2015-15. Added information to Developer Note section on the basis for the 15 minutes used in the EAL.

IC CA3 EAL #1 (1) UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown IC CA3 EAL #1 (1) UNPLANNED increase in RCS temperature to greater than (site-specific Technical Specification cold shutdown To address operating experience, added two notes and basis information on:

1) How to assess a temperature excursion if the decay heat removal function is available, and

2) Sources to use for RCS temperature information if

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 17 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis temperature limit) for greater than the duration specified in the following table.

temperature limit) for greater than the duration specified in the Table CA3-1, RCS Heatup Duration Thresholds.

reliable RCS indications are not available.

IC CA3 EAL #2 (2) UNPLANNED RCS pressure increase greater than (site-specific pressure reading). (This EAL does not apply during water-solid plant conditions. [PWR])

N/A None - deleted.

This EAL is unnecessary as the covered event presents a very low safety risk to the public. The assessment of the EAL is problematic during the specified modes because there may be periods where 1) the instrumentation needed to measure RCS pressure is not available and 2) the RCS is not intact. In addition, many plants are challenged to read small changes in RCS pressure during shutdown conditions with available instrumentation. RCS temperature indications are highly reliable and sufficient to identify and assess an RCS temperature increase. Should an issue occur with temperature indications during the Cold Shutdown and Refueling mode, it would be resolved quickly since stations typically have a large contingent of operations and technical staff onsite 24/7 to work the outage.

IC CA6 EAL #1 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

(1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Internal or external flooding event High winds or IC CA6 EAL #1 Hazardous event affecting two or more SAFETY SYSTEM trains.

(1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Internal or external flooding event High winds or tornado The IC and EAL were revised to incorporate concepts first captured in EPFAQs 2016-02 and 2018-04. Although these EPFAQs were a starting point, the information in both were significantly evolved during the development of Revision 7 to address lessons learned from operating experience and comments from the NRC staff. The key point is that an event would need to impact two or more safety system trains to be considered an actual or potential substantial degradation of the level of safety of the plant (i.e., an Alert).

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 18 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis tornado strike FIRE EXPLOSION (site-specific hazards)

Other events with similar hazard characteristics as determined by the Shift Manager AND

b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

OR

2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

strike FIRE EXPLOSION (site-specific hazards)

Other events with similar hazard characteristics as determined by the Shift Manager AND

b. The event has resulted in BOTH of the following:

1. Indications of degraded performance on a SAFETY SYSTEM train.

AND

2.

EITHER of the following:

a)

VISIBLE DAMAGE to a second SAFETY SYSTEM train.

OR b)

Indications of degraded performance to a second SAFETY SYSTEM train.

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Rev. 7 Wording Change Summary/Basis N/A N/A IC CA7 EAL #1 Control Room evacuation resulting in transfer of plant control to alternate locations.

This IC and EAL were relocated from the H Recognition Category to the C and S Recognition Categories.

IC CS1 EAL #1 EAL #2 EAL #3 (1) a. CONTAINMENT CLOSURE not established.

AND

b. (Reactor vessel/RCS

[PWR] or RPV [BWR])

level less than (site-specific level).

(2) a. CONTAINMENT CLOSURE established.

AND

b. (Reactor vessel/RCS

[PWR] or RPV [BWR])

level less than (site-specific level).

(3) a. (Reactor vessel/RCS

[PWR] or RPV [BWR])

level cannot be monitored for 30 minutes or longer.

AND

b. Core uncovery is indicated by ANY of the following:

(Site-specific radiation monitor) reading IC CS1 EAL #1 EAL #2 EAL #3 (1) a. CONTAINMENT CLOSURE not established.

AND

b. (RHR flow is lost and not restored within 30 minutes [PWR] or RPV level less than (site-specific level) [BWR]).

(2) a. CONTAINMENT CLOSURE established.

AND

b. (Reactor vessel/RCS level less than (site-specific level) [PWR] or Adequate core cooling cannot be assured

[BWR)]).

(3) a. (Reactor vessel/RCS

[PWR] or RPV [BWR])

level cannot be (monitored [PWR] or determined [BWR]) for 30 minutes or longer.

AND

b. Core uncovery is For the EAL 1.b, replaced the reactor vessel/RCS level criterion with a loss of RHR flow for 30 minutes. This change directly focuses the EAL on a loss of RHR flow, which is what the prior wording on reactor vessel/RCS level was concerned with (i.e.,

replaced the cause [low level leading to RHR suction loss] with the effect [lost RHR flow]). This change will also address EAL development issues experienced by some sites due to range limitations of available instruments.

For EAL 2.b, replaced RPV level criterion with Adequate core cooling cannot be assured. This change incorporates concepts first captured in EPFAQ 2019-04. Although the EPFAQ was a starting point, the information was evolved during the development of Revision 7. The key point is that operators would use whatever core cooling methods are specified in EOPs (which are developed appropriate to the plant design) and would make the declaration if it was determined that adequate core cooling cannot be assured. This approach is consistent with BWROG guidance for the development of EOPs.

In EAL #3.a, added a provision for BWRs to use the term determined per EPFAQ 2019-04.

Added a bullet to EAL (3).b since visual observation could also identify unisolable leakage.

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Rev. 7 Wording Change Summary/Basis greater than (site-specific value)

Erratic source range monitor indication

[PWR]

UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery (Other site-specific indications) indicated by ANY of the following:

(Site-specific radiation monitor) reading greater than (site-specific value)

Erratic source range monitor indication

[PWR]

UNPLANNED increase in (site-specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery Visual observation of UNISOLABLE RCS leakage of sufficient magnitude to make core uncovery likely (Other site-specific indications)

N/A N/A IC CS7 EAL #1 Challenge to core cooling safety function with Control Room evacuated.

This is the relocated IC and EAL from IC HS6. See discussion below for IC HS6.

IC CG1 EAL #1 EAL #2 Loss of (reactor vessel/RCS

[PWR] or RPV [BWR])

inventory affecting fuel clad integrity with containment IC CG1 EAL #1 Extended loss of core decay heat removal capability.

This IC and the associated EALs were revised to address issues with the current wording. The goal was to reduce challenges posed by the existing wording associated with assessing core and containment conditions while shutdown.

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Rev. 7 Wording Change Summary/Basis challenged.

For example:

Some Containment Closure measures may be temporary and may not have remote indications Instrumentation may be out-of-service for maintenance or repair Reliance on judgment calls concerning the magnitude of changes to tank or sump levels Radiation monitor readings were calculated based on assumed conditions and these may be different than actual conditions The revised wording should promote more timely and accurate emergency classifications. Additional supporting information is contained in the Basis and Developer Notes of the revised IC.

IC E-HU1 EAL #1 Damage to a loaded cask CONFINEMENT BOUNDARY.

(1) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than (2 times the site-specific cask specific technical specification allowable radiation level) on the surface of the spent fuel cask.

IC E-HU1 EAL #1 Damage to a loaded spent fuel cask.

(1) A closed window survey indicates EITHER of the following:

a. For a loaded spent fuel cask on the ISFSI pad - A general area dose rate greater than 10x normal radiation levels at any point along the pad boundary.

OR

b. For a loaded spent fuel cask in transit to the ISFSI This IC and EAL were revised to address operating experience.

For many sites, the EAL described in Rev. 6 was challenging to assess and to maintain as different cask technologies were placed into service. The revised wording eliminates the technical specification criterion (the source of the issues with the Rev. 6 EAL) and focuses instead on a measured dose rate.

This approach is used in other EALs (e.g., IC AU3), and should promote more timely and accurate emergency classifications.

Additional supporting information in contained in the Basis and Developer Notes.

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Rev. 7 Wording Change Summary/Basis pad - A cask dose rate greater than 10x the dose rate measured at the time the cask was sealed, at approximately the same distance.

FPB Table 9-F-2 Fuel Clad Barrier Loss 1.A (Site-specific indications that reactor coolant activity is greater than 300 Ci/gm dose equivalent I-131).

IC SA9 Reactor coolant activity > 2%

fuel clad failure.

This threshold, which addresses RCS activity levels indicative of significant fuel clad damage, was relocated to new IC SA9. The change was made recognizing that a site would be challenged to assess the threshold in a timely manner due to safety considerations associated with obtaining and analyzing high-dose rate RCS samples. The inability to make a timely threshold assessment could lead to an untimely (late) emergency classification; this potential outcome is unacceptable given the importance of the FPB matrix in the emergency classification scheme. It was determined that the threshold does have some value as a backup to other ICs/EALs; therefore, it was retained and relocated to a new Alert IC in the S Recognition Category.

FPB Table 9-F-2 Fuel Clad Barrier Loss 2.A Primary containment flooding required.

FPB Table 9-F-2 Fuel Clad Barrier Loss 2.A SAG entry required.

This threshold was changed to align with the decision-making guidance in the Emergency Procedure and Severe Accident Guidelines (EPG/SAGs), issued by the BWROG. The EPG/SAGs are used by BWR licensees to create their site-specific EOPs and SAGs. Changes made in EPG/SAGs Revision 3 necessitated this threshold change - refer to EPFAQ 2015-04. The threshold remains appropriate for the guidance in EPG/SAGs Revision 4.

FPB Table 9-F-2 Fuel Clad Barrier Potential RPV water level cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or FPB Table 9-F-2 Fuel Clad Barrier Potential RPV water level cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or cannot be No change to the threshold.

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Rev. 7 Wording Change Summary/Basis Loss 2.A cannot be determined.

Loss 2.A determined.

FPB Table 9-F-2 Fuel Clad Barrier Loss 4.A Primary containment radiation monitor reading greater than (site-specific value).

N/A None - deleted.

The primary containment radiation monitor reading is calculated using an assumed source term and instantaneous dispersal of the RCS inventory into primary containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the containment monitors could see radioactive shine from piping sources, which also affects assessment accuracy. For these reasons, the primary containment radiation monitor reading was removed. The identification of a significant challenge to the fuel clad barrier will be made using the safety-related indications for RPV water level. These indications are highly reliable (e.g., subject to the requirements in 10 CFR 50.65), and used to support diagnostic and mitigation actions in EOPs. This approach will result in more timely and accurate assessments of the status of the Fuel Clad Barrier than would be available from a primary containment radiation monitor.

FPB Table 9-F-2 Row 5, Other Indications, and Row 6, Emergency Director Judgment See wording in Rev. 6.

FPB Table 9-F-2 Row 5, Emergency Director Judgment See wording in Rev. 7.

The Other Indications row was deleted because experience has indicated that this row is seldom used. If a site has an indicator that is readily available to assess the status of a fission product barrier, then it is included in one of the thresholds in rows 1 through 4. The deletion of the Other Indications row moved up the Emergency Director Judgment row (from 6 to 5), so the associated thresholds were renumbered as 5.A and 5.B.

- This change affects all 6 columns in Table 9-F-2. **

FPB Table 9-F-2 Primary containment pressure greater than (site-specific value) due to RCS FPB Table 9-F-2 Primary containment pressure greater than (site-specific No change to the threshold.

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Rev. 7 Wording Change Summary/Basis RCS Barrier Loss 1.A leakage.

RCS Barrier Loss 1.A value) due to RCS leakage.

FPB Table 9-F-2 RCS Barrier Loss 2.A RPV water level cannot be restored and maintained above (site-specific RPV water level corresponding to the top of active fuel) or cannot be determined.

No change to the threshold.

FPB Table 9-F-2 RCS Barrier Loss 3.A UNISOLABLE break in ANY of the following: (site-specific systems with potential for high-energy line breaks).

No change to the threshold. The Basis section was revised to incorporate information from EPFAQ 2018-02.

FPB Table 9-F-2 RCS Barrier Loss 3.B Emergency RPV Depressurization.

No change to the threshold. The Basis section was revised to incorporate information from EPFAQ 2015-03.

FPB Table 9-F-2 RCS Barrier Potential Loss 3.A UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature OR

2. Max Normal Operating Area Radiation Level.

FPB Table 9-F-2 RCS Barrier Potential Loss 3.A UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Normal Operating Temperature OR

2. Max Normal Operating Area Radiation Level.

No change to the threshold.

FPB Table 9-F-2 Primary containment radiation monitor reading N/A None - deleted.

The primary containment radiation monitor reading is calculated using an assumed source term and instantaneous

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Rev. 7 Wording Change Summary/Basis RCS Barrier Loss 4.A greater than (site-specific value).

dispersal of the RCS inventory into primary containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the containment monitors could see radioactive shine from piping sources, which also affects assessment accuracy. For these reasons, the primary containment radiation monitor reading was removed. The identification of a significant challenge to the RCS barrier will be made using the safety-related indications for primary containment pressure, RPV water level, and detection of high energy line breaks. These indications are highly reliable (e.g., subject to the requirements in 10 CFR 50.65), and used to support diagnostic and mitigation actions in EOPs. This approach will result in more timely and accurate assessments of the status of the RCS Barrier than would be available from a primary containment radiation monitor.

FPB Table 9-F-2 CNMT Barrier Loss 1.A UNPLANNED rapid drop in primary containment pressure following primary containment pressure rise.

No change to the threshold.

FPB Table 9-F-2 CNMT Barrier Loss 1.B Primary containment pressure response not consistent with LOCA conditions.

No change to the threshold.

FPB Table 9-F-2 CNMT Barrier Primary containment pressure greater than (site-specific value).

FPB Table 9-F-2 CNMT Primary containment pressure greater than (site-specific value).

No change to the threshold.

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Rev. 7 Wording Change Summary/Basis Potential Loss 1.A Barrier Potential Loss 1.A FPB Table 9-F-2 CNMT Barrier Potential Loss 1.B (site-specific explosive mixture) exists inside primary containment.

FPB Table 9-F-2 CNMT Barrier Potential Loss 1.B (site-specific deflagration mixture) exists inside primary containment.

Changed explosive to deflagration to incorporate information from EPFAQ 2019-04. Deflagration is the concentration of concern in BWR EOPs/SAGs. Revised the Basis accordingly.

FPB Table 9-F-2 CNMT Barrier Potential Loss 1.C HCTL exceeded.

No change to the threshold but revised the Basis to remove a reference to Primary Containment Pressure Limit A to reflect information in EPFAQ 2019-04. Limit A is no longer used in BWR EPG/SAGs. Also revised the Developer Note to incorporate information from EPFAQ 2019-04; again the goal was to maintain alignment with BWR EPG/SAGs.

FPB Table 9-F-2 CNMT Barrier Potential Loss 2.A Primary containment flooding required.

FPB Table 9-F-2 CNMT Barrier Potential Loss 2.A It cannot be determined that core debris will be retained in the RPV.

Changed the threshold to incorporate the wording discussed in EPFAQ 2019-04. The change aligns the threshold with the appropriate diagnostic decision point described in the BWROG EPG/SAGs Revision 4. The Basis was revised accordingly.

FPB Table 9-F-2 CNMT Barrier Loss 3.A UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal.

No change to the threshold but revised the Basis to incorporate information from EPFAQ 2015-06. The new information provides clarity on the term direct path.

FPB Table 9-F-2 Intentional primary containment venting per FPB Table 9-F-2 Intentional primary containment venting per Added the term SAGs per EPFAQ 2019-04 since venting could be directed in SAG steps as well. Also revised the Basis to add

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Rev. 7 Wording Change Summary/Basis CNMT Barrier Loss 3.B EOPs.

CNMT Barrier Loss 3.B EOPs/SAGs.

information from EPFAQ 2019-04 dealing with releases due to intentional containment venting.

FPB Table 9-F-2 CNMT Barrier Loss 3.C UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

FPB Table 9-F-2 CNMT Barrier Loss 3.C UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

1. Max Safe Operating Temperature.

OR

2. Max Safe Operating Area Radiation Level.

No change to the threshold.

N/A N/A FPB Table 9-F-2 CNMT Barrier Potential Loss 3.A and 3.B A. Dose assessment using actual meteorology indicates doses greater than 750 mrem TEDE at or beyond (site-specific dose receptor point).

OR B. Field survey results indicate closed window dose rates greater than 750 mR/hr at or beyond (site-specific dose receptor point) that are expected to continue for 60 minutes or longer.

These thresholds are a replacement for the existing radiation monitor reading threshold in Containment Barrier Potential Loss 4.A. They are set to 75% of the lower limit of the EPA PAG for sheltering-in-place or evacuation of the public. Releases of this magnitude are far greater than normal containment leakage and, when combined with the loss of the fuel clad and RCS barriers, warrant the declaration of a General Emergency.

The resulting PARs, which may include an evacuation, will be more appropriate given that a significant release is in progress (i.e., one that is well in excess what would be expected from normal containment leakage).

FPB Table 9-F-2 Primary containment radiation monitor reading N/A None - deleted.

This threshold was replaced with Containment Barrier Potential Loss 3.A and 3.B; see discussion above for these new

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Rev. 7 Wording Change Summary/Basis CNMT Barrier Potential Loss 4.A greater than (site-specific value).

thresholds. The containment radiation monitor reading is calculated using an assumed source term and instantaneous dispersal of the RCS inventory into primary containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the actual containment monitor readings are influenced by in-containment conditions such as the use of sprays, natural deposition/plateout, atmospheric leakage, natural and forced convection, filters, and suppression pools. The existing threshold also makes possible the declaration of a General Emergency and issuance of an evacuation PAR even though there is no (and may never be) a radiological release that exceeds EPA PAGs offsite. The replacement thresholds, Containment Barrier Potential Loss 3.A and 3.B, will result in more appropriate emergency classifications and PARs.

FPB Table 9-F-3 Fuel Clad Barrier Potential Loss 1.A A. RCS/reactor vessel level less than (site-specific level).

No change to the threshold.

FPB Table 9-F-3 Fuel Clad Barrier Loss 2.A A. Core exit thermocouple readings greater than (site-specific temperature value).

No change to the threshold.

FPB Table 9-F-3 Fuel Clad A. Core exit thermocouple readings greater than (site-specific temperature FPB Table 9-F-3 Fuel Clad A. Core exit thermocouple readings greater than (site-specific temperature value).

No change to the threshold.

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Rev. 7 Wording Change Summary/Basis Barrier Potential Loss 2.A value).

Barrier Potential Loss 2.A FPB Table 9-F-3 Fuel Clad Barrier Potential Loss 2.B B. Inadequate RCS heat removal capability via steam generators as indicated by (site-specific indications).

N/A None - deleted.

A reassessment of this threshold concluded that it should be removed because the condition does not present an immediate threat to the Fuel Clad Barrier. During this condition, operators (following EOPs) will initiate a feed and bleed cooldown of the RCS. Absent an additional failure, this method of cooldown is sufficient to prevent a challenge to the Fuel Clad Barrier. Should an additional failure occur and lead to an actual Fuel Clad Barrier challenge, then another Potential Loss or Loss threshold would be met, ensuring an appropriate escalation of the emergency classification level.

FPB Table 9-F-3 Fuel Clad Barrier Loss 3.A A. Containment radiation monitor reading greater than (site-specific value).

N/A None - deleted.

The containment radiation monitor reading is calculated using an assumed source term and instantaneous dispersal of the RCS inventory into the containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the containment monitors could see radioactive shine from piping sources, which also affects assessment accuracy. For these reasons, the containment radiation monitor reading was removed. The identification of a significant challenge to the fuel clad barrier will be made using the safety-related indications for core exit thermocouples and reactor vessel level. These indications are highly reliable (e.g., subject to the requirements in 10 CFR 50.65), and used to support diagnostic and mitigation actions in EOPs. This approach will result in more timely and accurate assessments of the status of the Fuel Clad Barrier than would be available from a containment radiation monitor.

FPB Table 9-B. (Site-specific indications IC SA9 Reactor coolant activity > 2%

This threshold, which addresses RCS activity levels indicative of

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Rev. 7 Wording Change Summary/Basis F-3 Fuel Clad Barrier Loss 3.B that reactor coolant activity is greater than 300 Ci/gm dose equivalent I-131).

fuel clad failure.

significant fuel clad damage, was relocated to new IC SA9. The change was made recognizing that a site would be challenged to assess the threshold in a timely manner due to safety considerations associated with obtaining and analyzing high-dose rate RCS samples. The inability to make a timely threshold assessment could lead to an untimely (late) emergency classification; this potential outcome is unacceptable given the importance of the FPB matrix in the emergency classification scheme. It was determined that the threshold does have some value as a backup to other ICs/EALs; therefore, it was retained and relocated to a new Alert IC in the S Recognition Category.

FPB Table 9-F-3 Row 5, Other Indications, and Row 6, Emergency Director Judgment See wording in Rev. 6.

FPB Table 9-F-3 Row 5, Emergency Director Judgment See wording in Rev. 7.

The Other Indications row was deleted because experience has indicated that this row is seldom used. If a site has an indicator that is readily available to assess the status of a fission product barrier, then it is included in one of the thresholds in rows 1 through 4. The deletion of the Other Indications row moved up the Emergency Director Judgment row (from 6 to 5), so the associated thresholds were renumbered as 5.A and 5.B.

- This change affects all 6 columns in Table 9-F-3. **

FPB Table 9-F-3 RCS Barrier Loss 1.A A. An automatic or manual ECCS (SI) actuation is required by EITHER of the following:

1. UNISOLABLE RCS leakage OR

2. SG tube RUPTURE.

FPB Table 9-F-3 RCS Barrier Loss 1.A A. RCS subcooling has been lost.

This threshold was revised based on operating experience. A loss of subcooling is the fundamental indication that the available inventory control/makeup systems cannot adequately maintain RCS pressure and inventory against the mass loss through the leak. This condition represents a loss of the RCS Barrier.

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Rev. 7 Wording Change Summary/Basis FPB Table 9-F-3 RCS Barrier Potential Loss 1.A A. Operation of a standby charging (makeup) pump is required by EITHER of the following:

1. UNISOLABLE RCS leakage OR

2. SG tube leakage.

FPB Table 9-F-3 RCS Barrier Potential Loss 1.A A. An automatic or manual ECCS (SI) actuation is required by EITHER of the following:

1. UNISOLABLE RCS leakage OR

2. SG tube RUPTURE.

This threshold was revised based on operating experience.

Given the change above, it was determined that the ECCS (SI) actuation threshold would more appropriately define a potential loss of the RCS Barrier. The change also provides a threshold with better alignment to the definition and risk level of an Alert (because a potential loss of the RCS will lead to an Alert declaration).

FPB Table 9-F-3 RCS Barrier Potential Loss 1.B B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications).

No change to the threshold.

FPB Table 9-F-3 RCS Barrier Potential Loss 2.A A. Inadequate RCS heat removal capability via steam generators as indicated by (site-specific indications).

No change to the threshold.

FPB Table 9-F-3 RCS Barrier Loss 3.A A. Containment radiation monitor reading greater than (site-specific value).

N/A None - deleted.

The containment radiation monitor reading is calculated using an assumed source term and instantaneous dispersal of the RCS inventory into the containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the containment monitors could see radioactive shine from piping sources, which also affects assessment accuracy. For these reasons, the containment radiation monitor reading was removed. The

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Rev. 7 Wording Change Summary/Basis identification of a significant challenge to the RCS barrier will be made using the safety-related indications relied upon to assess challenges to the RCS Integrity and RCS Heat Removal safety functions, or initiate an ECCS (SI) actuation. These indications are highly reliable (e.g., subject to the requirements in 10 CFR 50.65), and used to support diagnostic and mitigation actions in EOPs. This approach will result in more timely and accurate assessments of the status of the RCS Barrier than would be available from a containment radiation monitor.

FPB Table 9-F-3 CNMT Barrier Loss 1.A A. A leaking or RUPTURED SG is FAULTED outside of containment.

FPB Table 9-F-3 CNMT Barrier Loss 1.A A

1. There is a Potential Loss or Loss of the RCS Barrier due to a leaking or RUPTURED SG.

AND

2. The leaking or RUPTURED SG is FAULTED outside of containment.

Revised the threshold to clearly state that the SG leakage or RUPTURE condition must be associated with RCS leakage meeting the threshold for either RCS Barrier Loss 1.A or RCS Barrier Potential Loss 1.A. It was always the intent that the RCS leakage must be to a leaking or RUPTURED SG before an SAE is warranted, but now the expectation is explicit.

FPB Table 9-F-3 CNMT Barrier Potential Loss 2.A A. 1. (Site-specific criteria for entry into core cooling restoration procedure)

AND

2. Restoration procedure not effective within 15 minutes.

FPB Table 9-F-3 CNMT Barrier Potential Loss 2.A A.

1. (Site-specific criteria for entry into core cooling restoration procedure)

AND

2. Restoration procedure not effective within 15 minutes.

No change to the threshold.

FPB Table 9-F-3 CNMT Barrier A. Containment radiation monitor reading greater than (site-specific value).

N/A None - deleted.

This threshold was replaced with Containment Barrier Potential Loss 4.C and 4.D; see discussion below for these new thresholds. The containment radiation monitor reading is

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Rev. 7 Wording Change Summary/Basis Potential Loss 3.A calculated using an assumed source term and instantaneous dispersal of the RCS inventory into containment. These assumptions may not be aligned with an actual event, thus affecting the accuracy of the threshold assessment. In addition, the actual containment monitor readings are influenced by in-containment conditions such as the use of sprays, natural deposition/plateout, atmospheric leakage, natural and forced convection, and filters. The existing threshold also makes possible the declaration of a General Emergency and issuance of an evacuation PAR even though there is no (and may never be) radiological release that exceeds EPA PAGs offsite. The replacement thresholds, Containment Barrier Potential Loss 4.C and 4.D, will result in more appropriate emergency classifications and PARs.

FPB Table 9-F-3 CNMT Barrier Loss 4.A A. Containment isolation is required AND EITHER of the following:

1. Containment integrity has been lost based on Emergency Director judgment.

OR

2. UNISOLABLE pathway from the containment to the environment exists.

FPB Table 9-F-3 CNMT Barrier Loss 4.A A. Containment isolation is required AND EITHER of the following:

1. Containment integrity has been lost based on Emergency Director judgment.

OR

2. UNISOLABLE pathway from the containment atmosphere to the environment exists.

Added the word atmosphere to improve clarity; this was a non-intent change. The releases of interest are sourced from gaseous radioactivity in the containment atmosphere.

FPB Table 9-F-3 B. Indications of RCS leakage outside of containment.

FPB Table 9-F-3 B. 1. There is a Potential Loss or Loss of the RCS Barrier due to UNISOLABLE RCS Revised the threshold to clearly state that the leakage outside containment condition must be associated with RCS leakage meeting the threshold for either RCS Barrier Loss 1.A or RCS

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Rev. 7 Wording Change Summary/Basis CNMT Barrier Loss 4.B CNMT Barrier Loss 4.B leakage.

AND

2. The leakage is to a location outside of containment.

Barrier Potential Loss 1.A. It was always the intent that the leak path must be from the RCS to a location outside containment before an SAE is warranted, but now the expectation is explicit.

FPB Table 9-F-3 CNMT Barrier Potential Loss 4.A A. Containment pressure greater than (site-specific value)

No change to the threshold.

FPB Table 9-F-3 CNMT Barrier Potential Loss 4.B B. Explosive mixture exists inside containment FPB Table 9-F-3 CNMT Barrier Potential Loss 4.B B. Flammable mixture in containment atmosphere Changed explosive to flammable as this is the term used for the mixture of concern in PWR EOPs/SAMGs. Revised the Basis accordingly.

FPB Table 9-F-3 CNMT Barrier Potential Loss 4.C C. 1. Containment pressure greater than (site-specific pressure setpoint)

AND

2. Less than one full train of (site-specific system or equipment) is operating per design for 15 minutes or longer.

FPB Table 9-F-3 CNMT Barrier Potential Loss 4.C C. Dose assessment using actual meteorology indicates doses greater than 750 mrem TEDE at or beyond (site-specific dose receptor point).

There are two parts to this change:

Part 1 - The old threshold was removed because it may lead to a General Emergency declaration and PARs during conditions when there is no (or a minimal) release in progress and no immediate challenge to containment integrity. Atmospheric pressure-related challenges to containment integrity are best bounded by Containment Barrier Potential Loss threshold 4.A, Containment pressure greater than (site-specific value),

where the value is the containment design pressure.

Part 2 - The new threshold is a replacement for the existing radiation monitor reading threshold in Containment Barrier Potential Loss 3.A. It is set to 75% of the lower limit of the EPA

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Rev. 7 Wording Change Summary/Basis PAG for sheltering-in-place or evacuation of the public.

Releases of this magnitude are far greater than normal containment leakage and, when combined with the loss of the fuel clad and RCS barriers, warrant the declaration of a General Emergency. The resulting PARs, which may include an evacuation, will be more appropriate given that a significant release is in progress (i.e., one that is well in excess what would be expected from normal containment leakage).

N/A N/A FPB Table 9-F-3 CNMT Barrier Potential Loss 4.D D. Field survey results indicate closed window dose rates greater than 750 mR/hr at or beyond (site-specific dose receptor point) that are expected to continue for 60 minutes or longer.

This is a companion threshold for Containment Barrier Potential Loss 4.C; see above discussion for Part 2.

IC HU1 EAL #1 EAL #2 EAL #3 Confirmed SECURITY CONDITION or threat.

No change to the IC or EALs. Added a basis statement to clarify that a site ISFSI is also within the scope of the IC.

Deleted a paragraph in the Basis section because it duplicated a paragraph in the Developer Notes section; the information is actually for developer usage.

IC HU2 EAL #1 Seismic event greater than OBE levels.

(1) Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:

(site-specific indication that a seismic event met or IC HU2 EAL #1 EAL #2 Seismic event greater than OBE levels.

(1) Seismic event greater than Operating Basis Earthquake (OBE) as indicated by:

(site-specific indication that a seismic event met or The IC was revised to add a second EAL (#2). This EAL is used when the sites seismic monitoring instrumentation is out-of-service (i.e., a backup EAL). Use of a backup seismic event EAL was discussed in NEI 99-01, Revision 6, but a decision was made to take the information from the Developer Notes and turn it into a separate EAL.

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Rev. 7 Wording Change Summary/Basis exceeded OBE limits) exceeded OBE limits)

OR (2) a. Seismic monitoring instrumentation is unavailable to the extent that an OBE cannot be determined (e.g., out-of-service for testing or maintenance).

AND

b. Control Room personnel feel an actual or potential seismic event.

AND

c. The occurrence of a seismic event is confirmed in manner deemed appropriate by the Shift Manager or Emergency Director.

IC HU3 EAL #1 EAL #2 EAL #3 EAL #4 EAL #5 Hazardous Event (1) A tornado strike within the PROTECTED AREA.

(1) Internal room or area flooding of a magnitude sufficient to require manual or automatic N/A None - deleted.

Deleted IC - EALs #1, #3, #4, and #5 are unnecessary as the covered events present a very low safety risk to the public.

Sites have sufficient procedures and capabilities to respond to these events without the need to activate an emergency plan (e.g., use of protocols and resources for responding to severe weather or industrial accidents). In particular, a site would be able to perform a post-event damage assessment, and identify and implement the necessary corrective/ compensatory

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Rev. 7 Wording Change Summary/Basis electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.

(3) Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g.,

an offsite chemical spill or toxic gas release).

(4) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.

(5) (Site-specific list of natural or technological hazard events) measures without mobilizing the ERO. Depending on the circumstances of the event, some plant response actions may also be required by Technical Specifications. Should the event have a more than minor impact, it would result in a report to the NRC in accordance with 10 CFR 50.72 or an emergency declaration under another IC.

EAL #2 was retained but relocated to Recognition Categories C and S as IC CU6 and IC SU7, respectively. The new locations were determined to be a more logical fit.

N/A N/A IC HU3 EAL #1 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.

This IC is a relocation of IC HA5. See discussion below for IC HA5.

IC HU4 EAL #1 EAL #2 EAL #3 FIRE potentially degrading the level of safety of the plant.

IC HU4 EAL #1 EAL #2 FIRE potentially degrading the level of safety of the plant.

EAL #1 (old EAL #3)

EAL #2 (old EAL #4)

EALs #1 and #2 were deleted; these EALs are unnecessary as the covered events present a very low safety risk to the public.

Sites have sufficient procedures and capabilities to respond to these events without the need to activate an emergency plan (e.g., use of protocols and equipment described in the site Fire

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Rev. 7 Wording Change Summary/Basis EAL #4 Protection Program). In particular, a site would be able to perform firefighting and a post-event damage assessment, and identify and implement the necessary corrective/

compensatory measures, without mobilizing the ERO.

Depending on the circumstances of the event, some plant response actions may also be required by Technical Specifications. Should the event have a more than minor impact, it would result in a report to the NRC in accordance with 10 CFR 50.72 or an emergency declaration under another IC.

EALs #3 and #4 were retained and renumbered as EAL #1 and EAL #2.

IC HU7 EAL #1 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a (NO)UE.

IC HU5 EAL #1 Other conditions exist which in the judgment of the Shift Manager/ Emergency Director warrant declaration of a (NO)UE.

Renumbered the IC based on other changes. Added Shift Manager for clarity.

IC HA1 EAL #1 EAL #2 EAL #3 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

No change to the IC or EALs. Pulled the definition of Owner Controlled Area into the Developer Notes (from Appendix B) based on user feedback.

Deleted a paragraph in the Basis section because it duplicated a paragraph in the Developer Notes section; the information is actually for developer usage.

IC HA5 EAL #1 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.

IC HU3 EAL #1 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.

This IC and EAL were relocated from the Alert level to the Unusual Event level; no changes were made to the IC or EAL wording. The change was made based on a reassessment of the potential impact of the event and associated operating experience. Sites have plans and resources for responding to a

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Rev. 7 Wording Change Summary/Basis hazardous materials event (e.g., those needed to meet OSHA or State requirements). A hazardous materials response does not require a full activation of the site ERO, which would occur following an Alert declaration. The declaration of an Unusual Event would ensure that key ERO managers are made aware of the event and available to support the response if needed.

Should the event have significant operational or radiological consequences, enough to warrant an Alert or higher classification, then the emergency declaration would be based on another IC.

IC HA6 EAL #1 Control Room evacuation resulting in transfer of plant control to alternate locations.

(1) An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

IC CA7 EAL #1 IC SA3 EAL #1 Control Room evacuation resulting in transfer of plant control to alternate locations.

(1) An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

No change to the IC or EAL; however, the IC and EAL were relocated from the H Recognition Category to the C and S Recognition Categories. The new locations were determined to be a more logical fit.

IC HA7 EAL #1 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.

IC HA5 EAL #1 Other conditions exist which in the judgment of the Shift Manager/ Emergency Director warrant declaration of an Alert.

Renumbered the IC based on other changes. Added Shift Manager for clarity.

IC HS1 EAL #1 HOSTILE ACTION within the PROTECTED AREA.

(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the IC HS1 EAL #1 HOSTILE ACTION within the PROTECTED AREA.

(1) A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the No change to the IC or EAL. Deleted a paragraph in the Basis section because it duplicated a paragraph in the Developer Notes section; the information is actually for developer usage.

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Rev. 7 Wording Change Summary/Basis (site-specific security shift supervision).

(site-specific security shift supervision).

IC HS6 EAL #1 Inability to control a key safety function from outside the Control Room.

(1) a. An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

AND

b. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes).

Reactivity control Core cooling [PWR] /

RPV water level

[BWR]

RCS heat removal IC CS7 EAL #1 Challenge to core cooling safety function with Control Room evacuated.

(1) a. Plant control has been transferred to locations outside the Control Room.

AND

b. EITHER of the following Initiating Conditions is met.

IC CA1, Loss of (reactor vessel/RCS

[PWR] or RPV [BWR])

inventory IC CA3, Inability to maintain the plant in cold shutdown This IC and EAL were relocated from the H Recognition Category to the C and S Recognition Categories. The new locations were determined to be a more logical fit.

Simplified the wording in EAL 1.a; no change to the intent.

Changed EAL 1.b to provide escalation criteria that reflects the intent of the previous criteria but is more appropriate for shutdown conditions. If IC CA1 or CA3 are met, then there is a challenge to removing heat from the RCS, and an Alert would be declared. Should this condition exist with the Control Room evacuated, then there may be additional challenges to controlling plant safety functions/equipment and escalation to a Site Area Emergency is appropriate.

IC HS6 EAL #1 Inability to control a key safety function from outside the Control Room.

(1) a. An event has resulted in plant control being IC SS3 EAL #1 Challenge to a fission product barrier with Control Room evacuated.

(1) a. Plant control has been transferred to locations This IC and EAL were relocated from the H Recognition Category to the C and S Recognition Categories. The new locations were determined to be a more logical fit.

Simplified the wording in EAL 1.a; no change to the intent.

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Rev. 7 Wording Change Summary/Basis transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

AND

b. Control of ANY of the following key safety functions is not reestablished within (site-specific number of minutes).

Reactivity control Core cooling [PWR] /

RPV water level

[BWR]

RCS heat removal outside the Control Room.

AND

b. ANY of the following conditions exist:

The reactor is not shutdown with adequate shutdown margin verified A loss or potential loss of Fuel Clad Barrier (per the Fission Product Barrier Table)

A loss or potential loss of RCS Barrier (per the Fission Product Barrier Table)

Changed EAL 1.b to provide escalation criteria that reflects the intent of the previous criteria but is more clearly defined. The new wording also promotes timely and accurate emergency declarations since operators will already be monitoring the status of the fission product barrier table thresholds and associated indications.

First bullet - a reactivity control problem is indicated if the The reactor is not shutdown with adequate shutdown margin verified.

Second bullet - if the Fuel Clad Barrier is potentially lost or lost, then there is a challenge to core cooling.

Third bullet - if the RCS Barrier is potentially lost or lost, then there is a challenge to RCS heat removal.

If either the Fuel Clad or RCS Barrier is lost, then an Alert would be declared. Should this condition exist with the Control Room evacuated, then there may be additional challenges to controlling plant safety functions/equipment and escalation to a Site Area Emergency is appropriate.

IC HS7 EAL #1 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency.

IC HS5 EAL #1 Other conditions exist which in the judgment of the Shift Manager/ Emergency Director warrant declaration of a Site Area Emergency.

Renumbered the IC based on other changes. Added Shift Manager for clarity.

IC HG1 EAL #1 HOSTILE ACTION resulting in loss of physical control of the facility.

N/A N/A This IC and EAL were deleted based on the resolution of EPFAQ 2015-13 (ML16166A366). This EPFAQ addressed the application of lessons learned from the first cycle of Hostile Action-Based (HAB) drills and exercises to IC HG1. NEI and the

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Rev. 7 Wording Change Summary/Basis industry had an opportunity to comment on the EPFAQ, and a public meeting was held to discuss and agree upon the resolution. The key point from the EPFAQ resolution is:

Based on these considerations, and given the confusion these redundant EALs had on EAL decision-making at the GE level, consideration can be given to not include EAL HG1 in a site-specific EAL scheme. However, EALs AA2, AS2, AG2, AS1, AG1, HS1, HS6, HS7, and HG7 shall be as provided in NEI 99-01, Revision 6 (ADAMS Accession No. ML12326A805) to ensure the intended event is appropriately bound at the correct ECL.

Although some were renumbered, all the cited EALs have been retained.

IC HG7 EAL #1 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.

IC HG5 EAL #1 Other conditions exist which in the judgment of the Shift Manager/ Emergency Director warrant declaration of a General Emergency.

Renumbered the IC based on other changes. Added Shift Manager for clarity.

IC SU1 EAL #1 Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

(1) Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 minutes or longer.

IC SU1 EAL #1 Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.

(1) Loss of ALL offsite AC power capability to (site-specific emergency buses) for 15 minutes or longer.

No change to the IC or EAL.

IC SU2 EAL #1 UNPLANNED loss of Control Room indications for 15 minutes or longer.

N/A None - deleted.

This IC and the associated EAL are unnecessary as the covered condition presents a very low safety risk to the public. Sites have sufficient procedures and capabilities to respond to this condition without the need to activate an emergency plan (e.g., use of protocols and resources for responding to a loss of

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Rev. 7 Wording Change Summary/Basis operationally significant indications). In particular, a site would be able to assess the equipment failure(s), and identify and implement any necessary corrective/compensatory measures without mobilizing the ERO. Some plant response actions may also be required by Technical Specifications. This condition would lead to a report to the NRC in accordance with 10 CFR 50.72 and, depending on concurrent events or resulting impacts, may necessitate an emergency declaration under another IC. Should this condition occur in conjunction with a reactor trip or ECCS (SI) actuation, then an Alert would be declared in accordance with IC SA2.

IC SU3 EAL #1 EAL #2 Reactor coolant activity greater than Technical Specification allowable limits.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered conditions present a very low safety risk to the public. Sites have sufficient capabilities to respond to this condition without the need to activate an emergency plan (e.g., procedures and resources described in Operations, Radiation Protection and Chemistry Programs). In particular, a site would be able to take the necessary actions to either lower RCS activity or shutdown the plant without mobilizing the ERO. These actions would be driven by requirements in the sites Technical Specifications.

Should the activity exceedance require a shutdown, then the condition would also lead to a report to the NRC in accordance with 10 CFR 50.72.

IC SU4 EAL #1 EAL #2 EAL #3 RCS leakage for 15 minutes or longer.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered conditions present a very low safety risk to the public. Sites have sufficient capabilities to respond to this condition without the need to activate an emergency plan (e.g., procedures and resources described in Operations, Radiation Protection and Chemistry Programs). In particular, a site would be able to take the necessary actions to either isolate the RCS leakage or shutdown the plant without mobilizing the ERO. These actions

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Rev. 7 Wording Change Summary/Basis would be driven by requirements in the sites Technical Specifications. Should the RCS leakage require a shutdown, then the condition would also lead to a report to the NRC in accordance with 10 CFR 50.72.

IC SU5 EAL #1 EAL #2 Automatic or manual (trip

[PWR] / scram [BWR]) fails to shutdown the reactor.

N/A None - deleted.

This IC and the associated EALs are unnecessary as the covered condition presents a very low safety risk to the public. Sites have sufficient procedures and capabilities to respond to an unsuccessful reactor trip/scram without the need to activate an emergency plan. It is worth noting that LWR power facilities are required to have ATWS mitigation equipment and strategies per 10 CFR 50.62 (which maintain very low event risk to the public), and that the associated mitigation equipment is subject to the maintenance requirements in 10 CFR 50.65 (thus ensuring high reliability). For this IC, although there was an issue with the RPS, the reactor was promptly shutdown following the initial trip/scram failure (through an alternative method) and no fission product barrier was challenged. The RPS issue would be addressed by the stations corrective action program. In addition, some plant response actions would be required by Technical Specifications and the site would make a report to the NRC in accordance with 10 CFR 50.72. Finally, this condition would not impact the ability of the site to implement the Emergency Plan or Security Plan, or require ERO mobilization or offsite support to assess and correct.

IC SU6 EAL #1 EAL #2 EAL #3 Loss of all onsite or offsite communications capabilities.

IC SU4 EAL #1 EAL #2 EAL #3 Loss of all onsite or offsite communications capabilities.

No change to IC or EALs. Renumbered the IC based on other changes. Added Developer Note guidance to address operating experience with electronic/internet-based notification methods (e.g., ROP FAQ 20-04).

IC SU7 Failure to isolate containment IC SU5 Failure to isolate containment No change to IC or EALs. Renumbered the IC based on other

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Rev. 7 Wording Change Summary/Basis EAL #1 EAL #2 or loss of containment pressure control. [PWR]

EAL #1 EAL #2 or loss of containment pressure control. [PWR]

changes.

N/A N/A IC SU7 EAL #1 Internal flooding affecting a SAFETY SYSTEM component required for the current operating mode.

This IC is the relocated EAL #2 from IC HU3, which was replaced with a new IC and EAL.

N/A N/A IC SU8 EAL #1 Automatic or manual (trip

[PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

This is the relocated IC and EAL #1 from IC SA5; see change description below for IC SA5.

IC SA1 EAL #1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

(1) a. AC power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer.

AND

b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.

IC SA1 EAL #1 Loss of all but one AC power source to emergency buses for 15 minutes or longer.

(1) Only one power source listed in Table SA1-1 is available to supply power to (site-specific emergency buses) for 15 minutes or longer.

Table SA1-1: AC Power Sources Offsite Source #1 Source #2, etc.

Onsite No change to IC statement. Revised EAL #1 to simplify the wording; no change to the intent (i.e., the EALs are functionally equivalent). Also added a provision to list credited power sources in the EAL (in Table SA1-1) per EPFAQ 2015-15.

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IC SA2 EAL #1 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.

ANY of the following transient events in progress.

Automatic or manual runback greater than 25%

thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] /

trip [PWR]

ECCS (SI) actuation Thermal power oscillations greater than IC SA2 EAL #1 UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. [PWR]

a. One or more of the following parameters cannot be determined from within the Control Room for 15 minutes or longer due to an UNPLANNED event.

[BWR]

ANY of the following transient events in progress.

Reactor scram [BWR] /

trip [PWR]

Added alternative EAL 1.a with a provision for BWRs to use the term determined per EPFAQ 2019-04.

Added provision for developers to specify the number of steam generators for which auxiliary or emergency feed water flow must be available. This allows the EAL to be more closely aligned with plant EOP requirements.

Deleted three of the listed transient events because their occurrence is not risk-significant enough to warrant an Alert declaration. These events would become sufficiently risk-significant if they lead to a reactor scram [BWR] / trip [PWR] or an ECCS (SI) actuation - these are the two transient events that have been retained. In addition, the three deleted events can challenge a Control Room staffs ability to determine the start time of the event. In many cases, a detailed review of computer logs or analog recorders would be required; these reviews could likely not be completed in time to support a required emergency declaration and notification.

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ECCS (SI) actuation N/A N/A IC SA3 EAL #1 Control Room evacuation resulting in transfer of plant control to alternate locations.

(1) An event has resulted in plant control being transferred from the Control Room to (site-specific remote shutdown panels and local control stations).

This IC and EAL were relocated from IC HA6. No change to IC or EAL.

IC SA5 EAL #1 Automatic or manual (trip

[PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

IC SU5 EAL #1 Automatic or manual (trip

[PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

This IC and EAL were relocated from the Alert level to the Unusual Event level; no changes were made to the IC or EAL wording. The change was made based on a reassessment of the potential event risk and consequences, and associated operating experience. Sites have procedures and capabilities to respond to an unsuccessful reactor trip/scram (e.g., strategies and equipment to meet 10 CFR 50.62), including the use of alternative measures to shut down the reactor before a fission product barrier is challenged (e.g., local opening of reactor trip breakers). In addition, some plant response actions would be required by Technical Specifications and the site would make a report to the NRC in accordance with 10 CFR 50.72. Further, this condition does not require full ERO mobilization or any offsite support to assess and correct. Should the event lead to a challenge of either the Fuel Clad Barrier or RCS Barrier, then an Alert classification would be made in accordance with the thresholds in the Fission Product Barrier Tables. Absent such a challenge, an Unusual Event declaration is appropriate.

IC SA9 Hazardous event affecting a SAFETY SYSTEM needed for IC SA7 Hazardous event affecting two or more SAFETY SYSTEM The IC and EAL were revised to incorporate concepts first captured in EPFAQs 2016-02 and 2018-04. Although these

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Rev. 7 Wording Change Summary/Basis EAL #1 the current operating mode.

(1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Internal or external flooding event High winds or tornado strike FIRE EXPLOSION (site-specific hazards)

Other events with similar hazard characteristics as determined by the Shift Manager AND

b. EITHER of the following:

1. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.

EAL #1 trains.

(1) a. The occurrence of ANY of the following hazardous events:

Seismic event (earthquake)

Internal or external flooding event High winds or tornado strike FIRE EXPLOSION (site-specific hazards)

Other events with similar hazard characteristics as determined by the Shift Manager AND

b. The event has resulted in BOTH of the following:

1. Indications of degraded performance on a SAFETY SYSTEM train.

AND

2.

EITHER of the following:

EPFAQs were a starting point, the information in both were significantly evolved during the development of Revision 7 to address lessons learned from operating experience and comments from the NRC staff. The key point is that an event would need to impact two or more safety system trains to be considered an actual or potential substantial degradation of the level of safety of the plant (i.e., an Alert).

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2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.

a)

VISIBLE DAMAGE to a second SAFETY SYSTEM train.

OR b) Indications of degraded performance to a second SAFETY SYSTEM train.

None See change discussion above for:

FPB Table 9-F-2, Fuel Clad Barrier Loss 1.A FPB Table 9-F-3, Fuel Clad Barrier Loss 3.B IC SA9 EAL #1 Reactor coolant activity > 2%

fuel clad failure.

The FPB matrix thresholds for RCS activity indicative of a fuel clad loss were relocated to this new IC. NEI 99-01, Revision 6, states that reactor coolant activity above 300 Ci/gm dose equivalent I-131 corresponds to an approximate range of 2%

to 5% fuel clad damage. To promote greater consistency in the development of the EAL for this IC (i.e., eliminate the ~3%

range), the 300 Ci/gm dose equivalent I-131 indication basis was replaced with 2% fuel clad failure. This value is indicative of a fuel clad barrier loss and, as a standard, will have licensees determine EAL indications based on the same level of fuel clad damage.

IC SS1 EAL #1 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.

No change to IC or EAL. Added a note and basis information to allow credit for non-safety-related power sources; this addition addressed EPFAQ 2015-15.

N/A N/A IC SS3 Challenge to a fission product barrier with Control Room evacuated.

This IC and EAL were relocated from IC HS6. See discussion above for HS6.

IC SS5 Inability to shutdown the reactor causing a challenge to N/A None - deleted.

This IC and the associated EALs are unnecessary as the classification of this condition is adequately addressed by the

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thresholds in the Fission Product Barrier (FPB) Tables. The two bulleted conditions in EAL statement (1).c entail a Potential Loss or Loss of both the Fuel Clad Barrier and the RCS Barrier; this condition would lead to a Site Area Emergency declaration under a FPB Table, regardless of the ATWS. Removing IC SS5 simplifies the emergency classification process.

IC SS8 Loss of all Vital DC power for 15 minutes or longer.

(1) Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer.

IC SS6 Loss of all Vital DC power for 15 minutes or longer.

(1) Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer.

Renumbered the IC based on other changes. No change to IC or EAL.

Deleted Developer Note on battery voltage - information was judged to be unnecessary since site-specific values should be considered.

IC SG1 EAL #1 Prolonged loss of all offsite and all onsite AC power to emergency buses.

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).

AND

b. EITHER of the following:

Restoration of at least one AC emergency bus in less than (site-specific hours) is not likely.

IC SG1 EAL #1 Extended loss of all AC power to emergency buses.

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses).

AND

b. (Site-specific indication of inadequate core cooling)

This IC and EAL were revised to remove to the loss of AC power coping time assessment as it is no longer relevant given the requirements in 10 CFR 50.155 (and the associated capabilities at each site). The new wording places the focus on indications of potential or actual core damage (i.e., inadequate core cooling). This condition challenges the RCS and Fuel Clad Barriers and, if further mitigation actions are unsuccessful, the Containment Barrier.

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(Site-specific indication of an inability to adequately remove heat from the core)

IC SG8 Loss of all AC and Vital DC power sources for 15 minutes or longer.

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 minutes or longer.

AND

b. Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer.

IC SG6 Loss of all AC and Vital DC power sources for 15 minutes or longer.

(1) a. Loss of ALL offsite and ALL onsite AC power to (site-specific emergency buses) for 15 minutes or longer.

AND

b. Indicated voltage is less than (site-specific bus voltage value) on ALL (site-specific Vital DC busses) for 15 minutes or longer.

Renumbered the IC based on other changes. No change to IC or EAL.

Added a note and basis information to address credit for non-safety-related power sources; this addition addressed EPFAQ 2015-15.

Deleted Developer Note on battery voltage - information was judged to be unnecessary since site-specific values should be considered.

Appendix A Acronyms and Abbreviations Appendix A Acronyms and Abbreviations Added a few new abbreviations.

Appendix B Definitions Appendix B Definitions Deleted the term CONFINEMENT BOUNDARY since it is no longer used in the scheme.

Revised the term IMMINENT based on operating experience and to improve clarity, Deleted the term NORMAL LEVELS since it is no longer used in the scheme.

Moved the term OWNER CONTROLLED AREA to the Developer Notes of IC HA1 where it is used and can be

Summary of IC and EAL Changes in NEI 99-01 Revision 7 Page 52 of 52 Rev. 6 IC and EAL#

Rev. 6 Wording Rev. 7 IC and EAL#

Rev. 7 Wording Change Summary/Basis more easily referenced.

Revised the term PROJECTILE to incorporate the NRCs definition.

Updated the term UNISOLABLE to incorporate EPFAQ 2018-01.

Revised the term VISIBLE DAMAGE to better align with the revised text in IC CA6 or SA7 Added a provision for BWR licensees to include definitions of cannot be maintained above/below and cannot be restored above/below, from EPG/SAG, Revision 4, to their emergency classification scheme, if those definitions appear in the site-specific EOPs and/or controlling development procedures. This change addressed information in EPFAQ 2019-04.

Appendix C All ICs and EAL in Recognition Category PD, Permanently Defueled N/A None - deleted.

This Recognition Category was deleted. Licensees can continue to follow the decommissioning scheme guidance in NEI 99-01, Revision 6, which is endorsed in Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors, Revision 6. Future changes to guidance on decommissioning schemes will be addressed in DG-1346, Emergency Planning for Decommissioning Nuclear Power Reactors [proposed new Regulatory Guide 1.235].

N/A N/A Appendix C See new Appendix C, Guidance for Radiation Effluent Monitor EALs.

The new Appendix C addresses the development of EALs based on calculated effluent radiation monitor readings per the Developer Notes in AA1, AS1, and AG2. Also adds information from EPFAQ 2015-09.

ML22195A169

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