Text

Response to Request for Additional Information (RAI) Regarding the License Amendment Request to Revise the Emergency Action Level (EAL) Scheme for the Emergency Plan
	ML15170A324
Person / Time
Site:	Fermi
Issue date:	06/18/2015
From:	Colonnello W; DTE Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC-15-0061
	Download: ML15170A324 (690)
	v • d • e

{{#Wiki_filter:Vito A. Kaminskas Site Vice President DTE Energy Company 6400 N. Dixie Highway, Newport, MI 48166 Tel: 734.586.6515 Fax: 734.586.4172 Email: kaminskasv@dteenergy.com

                                                             '        DTE Energy June 18, 2015                                                           10 CFR 50.90 NRC-15-0061 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) DTE Electric Company Letter to the NRC, "License Amendment Request to Revise the Emergency Action Level Scheme for the Emergency Plan," NRC-14-0055, dated October 21, 2014 (ML14295A078)

3) NRC Email to DTE Electric Company, "Fermi 2 - Request for Additional Information Regarding the License Amendment Request to Adopt NEI 99-01, Revision 6 (MF5048)", dated May 6, 2015 (ML15127A212)

Subject:

Response to Request for Additional Information (RAI) Regarding the License Amendment Request to Revise the Emergency Action Level (EAL) Scheme for the Emergency Plan In Reference 2, DTE Electric Company (DTE) submitted a license amendment request for Fermi 2 to revise the Emergency Action Level (EAL) scheme for the Emergency Plan with the scheme described in Nuclear Energy Institute (NEI) 99-01, Revision 6. In Reference 3, the NRC staff requested additional information to complete the review of Reference 2. provides DTE's response to the NRC staff request for additional information (RAI). Enclosure 2 provides the marked up EAL Technical Bases which incorporates the RAI responses in Enclosure 1. Enclosure 3 provides the EAL Technical Bases clean pages. Enclosure 4 provides the Radiological Emergency Response Preparedness (RERP) calculation referenced in Enclosure 1, RAI 6. This letter contains no new regulatory commitments.

USNRC NRC-15-0061 Page 2 Should you have any questions or require additional information, please contact Mr. Christopher R. Robinson of my staff at (734) 586-5076. I declare under penalty of perjury that the foregoing is true and correct. Executed on June 18, 2015, Wayne A. Colonnello Director - Nuclear Work Management for Vito A. Kaminskas

Enclosures:

1) Response to Request for Additional Information (RAI) Regarding the License Amendment Request to Revise the Emergency Action Level (EAL) Scheme for the Emergency Plan

2) Markup of Fermi 2 EAL Technical Bases

3) Revised (Clean) Fermi 2 EAL Technical Bases

4) Radiological Emergency Response Preparedness Calculation cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Michigan Public Service Commission Regulated Energy Division (kindschlamichigan.gov)

Enclosure 1 to NRC-15-0061 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Response to Request for Additional Information (RAI) Regarding the License Amendment Request to Revise the Emergency Action Level (EAL) Scheme for the Emergency Plan to NRC-15-0061 Page 1 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL For all the EALs, a plant-specific A review of all plant-specific bases, basis section was added in addition including the changes made in response to the NEI 99-01, Revision 6 basis to the NRC request for additional information, called generic in the information (RAI), indicates that there is package. While the staff does not no conflict in intent between the generic object to information being added to NEI 99-01, Revision 6 bases and the the EAL basis to ensure associated plant-specific bases. understanding of the particular EAL, it must be noted that when there is a potential conflict between the plant-specific and generic General Various basis sections, that the information in the generic basis in NEI 99-01, Revision 6 will be given precedence by the staff as it is an NRC-endorsed methodology for the development of EALs. To avoid any potential misunderstandings, please review all the EALs to ensure that information developed for the plant-specific sections are not in conflict with the generic sections, and revise accordingly, if necessary. In Section 1.0, Purpose of the Added the following to Section 1.0, Fermi 2 EAL Technical Bases Introduction, of the Emergency Action please incorporate the relevant Level (EAL) Technical Bases document: language from NEI 99-01, Because the information in a basis Revision 6, Section 4.6, related to document can affect emergency 1 1.0 the applicability of 10 CFR classification decision-making (e.g., the 50.54(q) to this document, or Emergency Director refers to it during provide justification for not an event), the NRC staff expects that incorporating. changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). The endorsed version of NEI 99-01, Revised the referenced ADAMS Revision 6 can be found at Accession No. to ML13091A209. ADAMS Accession No. ML13091A209. For clarity, please 2 2.1, 4.1 reference the correct ADAMS accession number of the endorsed version of NEI 99-01, Revision 6, on Pages 4 and 17. Section 3.1.2, Valid Indications, Added the following to Section 3.1.2, (1st paragraph) of the Fermi 2 EAL Valid Indications, of the EAL Technical 3 3.1.2 Technical Bases is derived from Bases document: NEI 99-01, Revision 6, Section 5.1 The validation of indications should be (3rd paragraph). The last sentence to NRC-15-0061 Page 2 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL from NEI 99-01, Revision 6, which completed in a manner that supports states, "The validation of timely emergency declaration. " indications should be completed in a manner that supports timely emergency declaration," was not included. While other areas of the Fermi 2 EAL Technical Bases document imply this, to ensure a consistent understanding, please add the quoted sentence from NEI 99-01, Revision 6, or provide further justification for not incorporating. NEI 99-01, Revision 6, Section 4.3, a. Fermi 2 has validated the "InstrumentationUsed for EALs," instrumentation value references used states "...scheme developers should in submitted EALs are within the ensure that specific values used as calibrated range of the referenced EAL setpoints are within the instrumentation and the resolution is calibrated range of the referenced appropriate for the setpoint/indication. instrumentation.. .EAL setpoint values should not use terms such as b. In all instances where the term Max

                      'off-scale low' or 'off-scale high'        Normal is used to quantify a set of since that type of reading may not          instrumentation, the instrumentation be readily differentiated from an           is suitable as a threshold for timely instrument failure."                       EAL declaration for the specific EAL or Fission Barrier Threshold.

a. Please confirm that all setpoints and indications used In lieu of using the established set of in the EAL scheme are within Max Safe instrumentation/values, an the calibrated range(s) of the alternate table with appropriate stated instrumentation and that thresholds values was developed and 4 PC-Loss 2_A the resolution is appropriate is provided as Table F-2. This table for the setpoint/indication. includes available radiation monitor

b. Where the term "Max Safe" is instrumentation in the Reactor used, the plant-specific Building that can be read in the information states that there is Control Room/Relay Room.

only one available instrument and that all other "Max Safe" Refer to RAI 29 response for values must be derived from additional information. surveys. This is not the staff's expectation as surveys do not provide for timely EAL declaration, and are only acceptable based on the NRC-endorsed methodology when determining radiation levels for the Central Alarm Station used in EAL RA3.1. For all instances where the terms "Max Safe" or "Max Normal" are used to quantify a set of to NRC-15-0061 Page 3 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL instrumentation, please confirm the suitability of this instrumentation as a threshold for timely EAL declaration for the specific EAL or Fission Barrier Threshold, and if necessary, develop EALs using specific instrumentation and values rather than this established set of "Max Safe" instruments/values. In Section 5.0, "Definitions," for Added the following definitions to consistency with the approved Section 5.0, Definitions, Acronyms & guidance, please include the Abbreviations, of the EAL Technical following definitions from NEI 99- Bases document: 01, Revision 6, or provide " Independent Spent Fuel Storage justification for not incorporating: Independent SF lt, Installation (ISFSI),

                            " Independent Spent Fuel Storage      " Unusual Event, Installation (ISFSI),              e Alert,
                            " (Notification of) Unusual           e Site Area Emergency, Event,                             . General Emergency,
                            " Alert,                              e  Initiating Condition,
                            "  Site Area Emergency,               . Emergency Action Level;
                            "  General Emergency,                 . Fission Product Barrier Threshold, and
                            "  Initiating Condition,              " Emergency Classification Level
                            "  Emergency Action Level;
                            "  Fission Product Barrier          The Owner Controlled Area definition 5            5.0              Threshold, and                   has been aligned with the definition
                            " Emergency Classification          provided in EP-101 Emergency Level.                           Classification as follows:

In addition, for emergency planning Owner ControlledArea - The company purposes, please confirm that the property immediately surroundingthe site-specific definitions of Owner PROTECTEDAREA security fence. Controlled Area (OCA) and Site Access is normally limited to people on Boundary are correctly applied official business. throughout the document, as The Site Boundary definition is correct as typically the OCA is the area provided. beyond which the licensee does not own, lease, or otherwise control, and the Site Boundary is a sub-area of the OCA. Please revise, or elaborate in the document, as appropriate to ensure understanding on intended application. For EALs RG1.1, RS1.1 and The bases for the current Fermi 2 effluent 6 RUl.1, RS1.1, RG1.1 RU1.1, please provide justification monitor thresholds is provided in the for the difference in radiation Additional Information section of the to NRC-15-0061 Page 4 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL monitor threshold values from the respective AU1, AS1 and AG1 bases in values currently provided in EP- EP-101 Classification of Emergencies 101, dated May 7, 2013 (ADAMS Enclosure A, Tab A. Accession No. ML13212A154), or The proposed Table R-1 effluent revise accordingly. thresholds are based on calculation EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values, Revision 0, developed in support of the NEI 99-01, Revision 6, EAL scheme change. The calculation methodology is in alignment with the respective AU1, AS1 and AG1 NEI 99-01, Revision 6, developer notes. Calculation EP-EALCALC-FERMI-1401 is provided in Enclosure 4. For EAL RA1.1, the Plant-Specific Revised plant-specific bases information. Basis information describes the Changed UE to Alert. Unusual Event basis, when this 7 RA1.1 EAL is for the Alert. Please provide justification for this apparent misalignment, or revise accordingly. NRC regulations in Appendix E, All the listed instrumentation in RU2.1 Section IV.C.2, to 10 CFR Part 50 and RA2.2 is available to be read within require the licensee to establish and the Control Room. maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an EAL has been exceeded and to promptly 8 RU2.1, RA2.2 declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. For EALs RU2.1 and RA2.2, please confirm that all the stated instrumentation is available within the Control Room, and if not, identify and justify the impact on timely declaration of this EAL. Category AA2 (2nd paragraph) from The following has been added to the NEI 99-01, Revision 6 states, This RA2.1 generic bases: IC applies to irradiated fuel that is 9 RA2.1 This EAL applies to irradiated fuel that licensed for dry storage up to the is licensed for dry storage up to the point point that the loaded storage cask is that the loaded storage cask is sealed. sealed. Once sealed, damage to a loaded cask to NRC-15-0061 Page 5 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL For EAL RA2.1, please explain causing loss of the CONFINEMENT why discussion of the applicability BOUNDARY is classified in accordance of this EAL until spent fuel is with EU1.1. sealed in a licensed dry cask, was not incorporated, or revise accordingly. Category CU3 from NEI 99-01, Deleted Visual observation of reactor Revision 6, states that this EAL cavity level from the Refueling Floor reflects a condition where there from CU3.2 bases. has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. For EAL CU3.2, the bulleted list of 10 CU3.2 available instruments for monitoring reactor pressure vessel (RPV) level includes visual observation of reactor cavity level from the refueling floor. The intent of this EAL is to declare an event upon a loss of instrumentation; therefore the inclusion of this bullet would be inconsistent with EAL intent. Please revise accordingly, or provide further justification for this criterion. For EALs CU5.1 and SU7.1, the Removed Michigan (MI) State Radios use of the Michigan (MI) State (800 MHz) from the onsite Radios (800 MHz) is listed as a communication method column in Tables communication method; however, in CU5.1 and SU7.1. they are not discussed in the Plant-11 CU5.1, SU7.1 Specific description. Please explain how the MI State Radios (800 MHz) work for in-the-plant communications, or remove it from consideration as an onsite communication method. For EAL HU2.1, the staff expects An operator is dispatched immediately licensees to use the EAL method upon receipt of the associated annunciator 12 HU2.1 described in Category HU2 of NEI 6D69. The operator reviews the printout, 99-01, Revision 6 when checking for peak acceleration data and instrumentation is not available in determines any error status. Once the Control Room to directly determined, they immediately contact the to NRC-15-0061 Page 6 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL support the EAL. Please explain in Control Room with results. The readout is greater detail: (1) how these located in the Control Room envelope specific seismic values are within the Relay Room. During multiple determined; and (2) the timing of drills conducted involving seismic this determination after the receipt activity, the time to dispatch an operator, of the Seismic System obtain the data, and communicate to the Event/Trouble annunciator. Control Room is <5 minutes. For EALs HU4.1 and HU4.2, there Deleted the plant-specific bases related to is a potential for misunderstanding classification clock start from both HU4.1 between the information contained and HU4.2 plant-specific bases. in the plant-specific section and the generic section as to when the 13 HU4.1, HU4.2 declaration clock starts. To ensure consistent understanding, please use the NEI 99-01, Revision 6 endorsed wording, or provide further justification to support staff evaluation of an alternative. Category HU4 in NEI 99-01, The Fermi 2 ISFSI is physically located Revision 6 includes the ISFSI for inside the Plant Protected Area. plants with the ISFSI outside of the Therefore, the ISFSI is not referenced in Protected Area. However, the staff either EAL. could not decipher from the referenced UFSAR Figure 1.2-5 if the ISFSI is located within the 14 HU4.3, HU4.4 Protected Area. For EALs HU4.3 and HU4.4, please clarify whether the ISFSI is located within the protected area. If the ISFSI is not considered to be part of the Protected Area, add the ISFSI to the EALs, or provide justification for not incorporating. Category HU4 in NEI 99-01, Added the omitted Appendix R bases Revision 6 specifies basis related discussion to EALs HU4.1 and HU4.2. requirements from 10 CFR 50, HU4.3 and HU4.4 address fires outside of Appendix R. those areas that contain structures, For EALs HU4.1, HU4.2, HU4.3 system, and components important to 15 HU4.1, HU4.2 and HU4.4, there does not appear to achieving and maintaining safe shutdown be any Appendix R discussion. conditions as outlined in Table H-1. Please include the information from Those areas are specifically covered by the NEI 99-01, Revision 6 guidance HU4.1 and HU4.2. Therefore, the related to Appendix R and the reference to the Appendix R basis is not timing, or provide justification for applicable for HU4.3 and HU4.4. not incorporating to NRC-15-0061 Page 7 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL For EAL HA5.1, the plant- Added Control Room to HA5.1 Table specific basis states that the H-2 applicable under all modes. Control Room is not included as it is addressed in RA3.1; yet, EAL RA3.1 does not bound atmospheric gasses. Typically a well-designed Control Room has a ventilation system capable of protecting the 16 HA5.1 operations staff from the consequences of an external gaseous release and, therefore, need not be included in EAL HA5.1. Please provide further justification on why the Control Room is excluded from the applicability of this EAL, or revise accordingly. The plant-specific basis in EALs The cited bases is intended to define HA6.1 and HS6.1 contain a bullet control relative to the safety functions, stating, RPV level and pressure not the point at which classification is are being controlled from the assessed. Dedicated or Remote Shutdown However, the following statement was Panels. This implies that the added to the HA6.1 plant-specific bases licensee will wait until this bullet is for clarity: met to declare the EAL when, in fact, the staff expects the For the purpose of this EAL the 15 17 HA6.1 declaration start clock for this EAL minute classification clock starts when to be when the last licensed the last licensed operator leaves the operator leaves the Control Room Control Room. (as reflected in first sentence of the HS6.1 plant-specific basis). Please explain in more detail why the staff should consider the addition of the above statement, or revise accordingly. For EAL HS6.1, please explain HS6.1 has been revised to reflect the why operating mode specificity to modified mode applicability as described the key safety functions list was not below. addressed, as the shutdown For BWRs, the specified safety control operating modes may not need functions are reactivity, Reactor Pressure control of these safety functions as Vessel (RPV) water level, and Reactor 18 HS6.1 quickly as the hot operating modes. Coolant System (RCS) heat removal. None of these safety control functions are necessary in the Defueled Mode. Further, once the reactor enters Mode 3, the reactor is, by definition, shutdown under all conditions without boron. Therefore, reactivity control would only be to NRC-15-0061 Page 8 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL applicable in Modes 1 and 2. Note 10 in EALs SA1.1 and SS1.1 Added Note 10 to SU1.1 and SG1.1. state, Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 minutes. However, Note 10 was 19 SU1.1, SG1.1 not included in EALs SU1.1 and SG1.1. Please explain why Note 10 was not added to EALs SU1.1 and SG1.1 to reflect the restrictions when considering the combustion turbine generators (CTGs), or revise accordingly. For EALs SU4.1 and SU4.2, a Added a new Note 11 to SU4.1 that reads: comment was added to the plant-Consistent with Technical Specification specific basis section to limit the 3.7.5, this EAL is applicable at all times consideration of this EAL when in while in Mode 1, Mode 2 or in Mode 3 Operating Mode 3 (Hot Shutdown). with any main steam line not isolated and While the staff has no technical steam jet air ejector in operation. 20 SU4.1, SU4.2 issue with this restriction, this should be incorporated as a note to Added a new Note 12 to SU4.2 that reads: the EAL and included on the EAL Wallboard, as it changes the Consistent with Technical Specification 3.4.7, this EAL is applicable at all times applicability of the EAL. Please revise accordingly, or remove the while in Mode 1, Mode 2 or in Mode 3 restriction, as applicable. with any main steam line not isolated. For EALs SU6.1, SU6.2, SA6.1 The method used to determine the reactor and SS6.1, please note that it is not is shutdown following a reactor scram for the staff's expectation that licensees the purposes of emergency classification include a specific power level to is consistent with the Fermi 2 EOPs; denote shutdown, as licensed reactor power indicating <3% (Average operators are trained to determine if Power Range Monitor (APRM) the plant is shutdown, or not, using downscale). As specified in the generic power as well as other plant developers guidance, Developers may parameters. Please explain the include site-specific EOP criteria SU6.1, SU6.2, SA6.1, 21 basis for including a specific power indicative of a successful reactor SS6.1 level. shutdown in an EAL statement, the Basis or both (e.g., a reactor power level). In addition, the staff notes that the Reactor power < 3% is therefore the Startup Operating Mode is not plant-specific indication of a successful included in Category SU5 in NEI reactor scram. Per the generic SU5, SA5 99-01, Revision 6. Please explain and SS5 developer notes: in further detail the basis for the addition of the Startup operating This IC is applicable in any Mode in mode to EALs SU6.1, SU6.2, which the actual reactor power level SA6.1 and SS6.1, or revise could exceed the power level at which to NRC-15-0061 Page 9 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL accordingly. the reactor is considered shutdown. A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability. For example, if the reactor is considered to be shutdown at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. For BWRs, including Fermi 2, the plant operating mode is defined by the position of the Mode Switch. Startup mode is defined as Mode Switch in Startup/Hot Standby or Mode Switch Refuel with all reactor vessel head closure bolts fully tensioned. By procedure during reactor startup, the Mode Switch is not repositioned from the Startup/Hot Standby position until reactor power is between 5% and 10% (above the APRM downscale value of 3%). Therefore, consistent with the generic developer note, these EALs are applicable in both Power Operation (Mode 1) and Startup (Mode 2) operating modes. The basis section of Category SS5 The cited bases statement is in alignment in NEI 99-01 is intended to reflect with the generic IC and guidance. that this EAL addresses a condition Generic SS5 IC is: when all means of achieving shutdown are unsuccessful, and Inability to shutdown the reactor continued power generation is causing a challenge to (core cooling challenging the capability to [PWR] / RPV water level [BWR]) or remove heat from the core and/or RCS heat removal RCS. The generic bases for SS5 states: The plant-specific section of EAL 22 SS6.1 This IC addresses a failure of the RPS SS6.1 states, For this Site Area to initiate or complete an automatic or Emergency EAL, reactor shutdown achieved by injection of boron or manual reactor (trip [PWR] / scram use of the alternate control rod [BWR]) that results in a reactor insertion methods of 29.ESP.03 is shutdown, [emphasis added] all also credited provided [emphasis subsequent operator actions to added] reactor power can be manually shutdown the reactor are unsuccessful, and continued power reduced below the APRM generation is challenging the capability downscale trip set point before to adequately remove heat from the core indications of an extreme and/or the RCS. challenge to either core cooling or to NRC-15-0061 Page 10 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL heat removal exists." Please Thus, if any action can be taken to explain the addition of the above shutdown the reactor (reactor power sentence as it appears to be <3%) before either of the bulleted misaligned with the initiating condition exist, the EAL is not exceeded. conditions described in NEI 99-01, Revision 6. For EAL SA8.1, please explain the Deleted the following paragraph from the information added to the "plant- SA8.1 and CA6.1 plant-specific bases: specific" basis section (14'paragraph) that potentially "The term "required' as used in this conflicts with the 3 rd paragraph of EAL is defined as the number of operable the "generic" basis section with systems required by Technical respect to the operating mode. In Specificationsfor the present operating 23 SA8.1, CA6.1 addition, please explain the value- mode. Therefore, damage to systems that added to the EAL with the does not affect the required number of inclusion of this "plant-specific" systems required to meet Technical information, particularly when it is Specificationsfor the current mode different than the "generic" (NEI would not require classification." 99-01, Revision 6 information), or The paragraph was deleted from CA6.1 revise accordingly. because it is similar to SA8.1, with the only difference being mode applicability. Category E-HU1 in NEI 99-01, Revised EU1.1 to the following: Revision 6 provides an example EAL which states, "Damage to a "Damage to a loaded cask loaded cask cak COFINEENTCONFINEMENT loadd BOUNDARY as CONFINEMENT indicated by an on-contact radiation BOUNDARY as indicated by an reading greater than [emphasis added] on-contact radiation reading greater EITHER of the following on the surface than (2 times the site-specific cask E R o thelowin onteaurac specific technical specification ofthe spentfuel cask (overpack): allowable radiation level) on the . 60 mrem/hr (r + n) on the top of surface of the spent fuel cask." the overpack 24 EU.l1 For EAL EU1.1, it is unclear that OR either of the two bullets listed will suffice for making this declaration. " 600 mrem/hr (r + n) on the side Please explain further, or clarify of the overpack excluding inlet accordingly (e.g., including a logic and outlet ducts"

                        'or' statement).                      Corrected the identified typo.

Also note the misspelling of 'multi-purpose canister' (milti-purpose canister), when describing the confinement boundary in the "plant-specific" section. For EAL EU1.1, the last sentence Deleted the cited statement from the of the 1t paragraph developed for category description. 25 EU 1.1 CATEGORY E the Category E discussion states, "Formaloffsite planning is not requiredbecause the postulated to NRC-15-0061 Page 11 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL worst-case accident involving an ISFSI has insignificant consequences to the public health and safety. This statement is not applicable for use in an EAL scheme for a licensed and operating reactor site. Please provide justification for including this statement, or revise accordingly. Table 9-F-2, BWR EAL Fission Table F-1, Fission Product Barrier (FPB) Product Barrier Table, in NEI 99- Threshold Matrix has been modified to 01, Revision 6 provides logic for reflect the logic presented in NEI 99-01, the thresholds with multiple Revision 6. considerations (e.g., any, or). 26 TABLE F-1 Table F-1, Fission Product Barrier Threshold Matrix, in the licensees submittal differs in logic with that in NEI 99-01, Revision 6. Please clarify the expected logic to be used and/or explain any deviations from the NEI 99-01, Revision 6. Under the Fission Barrier Matrix As part of the Fukushima driven for FC-Loss 1.A and PC-Potential Emergency Operation Procedure Loss (PLoss) 1.A, please provide (EOP)/Severe Accident Guideline (SAG) additional justification as to why strategy changes, BWROG Emergency criterion SAG entry is required Procedure Guideline (EPG)/SAG, should be considered, in lieu of NEI Revision 3, was developed by the 99-01, Revision 6 barrier threshold BWROG and is being implemented by of "Primary containment flooding Fermi 2. In Revision 3, the transition to required," or revise accordingly the SAGs is no longer Primary Containment Flooding is required but is now signaled by conditions indicative of a loss of adequate core cooling which FC-Loss 1.A, PC- then requires exit from all EOPs and entry 27 PLoss 1.A to the SAGs. This is because the need for the primary containment flooding strategy is now a discretionary decision that would likely only be made for a very limited set of conditions once the SAGs are entered. The FC-Loss 1.A and PC-PLoss 1.A threshold and bases must be revised to reflect this change in strategy and the disconnect between primary containment flooding and SAG entry associated with implementation of BWROG EPG/SAG Revision 3. Both FC-Loss 1.A and PC-PLoss 1.A thresholds have been revised to NRC-15-0061 Page 12 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL as follows:

                                                                 "Inadequatecore cooling as indicated by ANY of thefollowing:

1. RPV water level cannot be restored and maintained> -48 in. with > 5725 gpm Core Spray loop flow OR

2. RPV water level cannot be restored and maintained> -25 in with

                                                                         < 5725 gpm Core Spray loop flow OR

3. RPV water level cannot be determined and core damage is occurring" The threshold bases have been revised to reflect the above change.

The NEI 99-01, Revision 6 basis Revised RCS-Loss 2.A, RCS-PLoss 2.A, (for RC-Loss 2.A), as shown in the PC-Loss 2.A, and PC-Loss 5.A Plant-

                       "generic" basis section of RC-Loss     Specific bases to include the following:

2.A) states, "If it is determined that is the rupturedline cannot beleak promptly isolatedfrom the isolated to preclude EAL declaration [emphasis added] Control Room, must occur within the 15-minute the RCS BarrierLoss threshold is assessment period." RCS-Loss 2.A, met. " 28 RCS-PLoss 2.A, The basis for RC-Loss 2.A and RC-PC-Loss 2.A, PLoss 2.A define the term PC-Loss 5.A "Unisolable" as, "An open or breachedsystem line that cannot be isolated,[emphasis added] remotely or locally." Please provide further justification to support the ability to promptly isolate locally, or revise accordingly consistent with NEI 99-01, Revision 6. RCS-PLoss 2.A, For RC-PLoss 2.A and PC-Loss a. Secondary Containment Control Max 29 PC-Loss 2.A 2.A, please address the following: Normal Operating Temperatures

a. Please verify that all of the (EOP Table 12) and Max Normal to NRC-15-0061 Page 13 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL "Max Normal" or "Max Safe" Area Radiation Levels (EOP Table instruments are able to be used 14) can be determined from the in the time requirements Control Room/Relay Room in the expected of an EAL, or revise time requirements expected of an to list specific instrumentation EAL.

that would meet these requirements. Secondary Containment Control Max re ants. basis Safe Operating Temperatures (EOP information excludes all but Table 12) can be determined from the one radiation monitor for use, Control Room/Relay Room in the and states that all other values time requirements expected of an will be determined by survey. EAL. This is not consistent with the Based on the available radiation guidance in NEI 99-01, monitor instrumentation in the Revision 6 as: (1) Reactor Building that can be read in exclusionary information the Control Room/Relay Room, an should be part of the alternate table with appropriate threshold, not hidden within thresholds has been developed. The the basis information; and (2) threshold values are provided in Table only the Central Alarm Station F-2. is acceptable for surveys due to the timing associated with Revised Table F-1 Fission Product getting them done. Please Barrier Threshold Matrix and PC-provide further justification Loss 2.A to reflect the addition of for staff evaluation, or revise Table F-2. accordingly. "UNISOLABLE primary system

c. The guidance in NEI 99-01, leakage into Secondary Containment Revision 6 is that the EAL that results in exceeding EITHER of declaration clock starts when the following:

the "Max Normal" or "Max Safe" alarm/indication occurs, 1. One or more Secondary and that determination of Containment ControlMax Safe whether the leak is isolated to Operating Temperatures (EOP preclude EAL declaration Table 12) must occur within 15-minutes. Please explain how proposed OR EALs will ensure that this is the understanding of the EAL 2. One or more Secondary decision-makers, or add Containmentarearadiationlevels guidance as applicable. (Table F-2)

d. Please explain why RC-PLoss b. See response to RAI 29.a above.

2.A describes "Max Safe" conditions in the "plant- c. As a result of the changes to the specific" basis section, when definition of Unisolable and the this threshold is for "Max development of Table F-2, the Normal", or revise declaration timing requirements will be accordingly. met. See responses to RAIs 28 and 29.a for additional information.

d. Deleted the paragraph describing "Max Safe" conditions in the RCS-to NRC-15-0061 Page 14 RAI # TECHNICAL BASES NRC RAI Fermi 2 RAI Response SECTION/EAL PLoss 2.A plant-specific bases.

The threshold for RC-Loss 3.A Deleted the following from the RCS-Loss states, Drywell pressure > 1.68 3.A plant-specific bases: psig due to RCS leakage. Loss of drywell cooling that results in In the plant-specific basis section pressure greater than 1.68 psig should for RC-Loss 3.A, the sentence, not be considered an RCS barrier loss 30 RCS-Loss 3.A Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS Barrier Loss, was added. Please explain why this sentence was added, or revise the plant-specific basis section accordingly Under the plant-specific basis for Deleted the cited last paragraph from PC-PC-PLoss 3.B, the last paragraph PLoss 3.B plant-specific bases. regarding if the hydrogen or oxygen Equipment operability issues are monitor is unavailable requires addressed under the Fermi Equipment significantly more justification to Important to Emergency Response support staffs evaluation. It is (EITER) program. expected that the ability to monitor drywell and suppression pool hydrogen and oxygen 31 PC-PLoss 3.B concentrations is maintained during the operating modes applicable to the Fission Barrier Matrix. If this ability is compromised, then the licensee is required to compensate for this loss and timely restore functionality of the instrumentation. Please provide further justification, or revise accordingly. Under the Fission Barrier Matrix Deleted the cited sentence from the PC-for PC-PLoss 3.C, please explain PLoss 3.C plant-specific bases. why the plant-specific basis states, This threshold should be considered when EOP Primary Containment Control Step TWT-5 is reached and emergency RPV depressurization is required. This 32 PC-PLoss 3.C seems to imply that if the Heat Capacity Limit (HCL) is reached by any other emergency operating procedure, it would not result in classification. Please clarify if the statement above is the only instance where the threshold would be met, or revise accordingly.

Enclosure 2 to NRC-15-0061 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Markup of Emergency Action Level Technical Bases to be incorporated into Implementing Procedure EP-101, Classification of Emergencies to NRC-15-0061 Page 1 Fermi 2 Emergency Action Level Technical Bases TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ................................................................................................... 3 2.0 DISCUSSION .............................................................................................. 3 2.1 Background ....................................................................................... 3 2.2 Fission Product Barrier Thresholds ................................................... 4 2.3 Fission Product Barrier Classification Criteria ................................... 5 2.4 EAL Organization .............................................................................. 5 2.5 Technical Bases Information ............................................................. 8 2.6 Operating Mode Applicability ............................................................ 9 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ................... 11 3.1 General Considerations .................................................................... 11 3.1.1 Classification Timeliness ..................................................... 11 3.1.2 Valid Indications .................................................................. 11 3.1.3 Imminent Conditions ............................................................ 12 3.1.4 Planned vs. Unplanned Events............................................ 12 3.1.5 Classification Based on Analysis ......................................... 12 3.1.6 Emergency Director Judgment ............................................ 13 3.2 Classification Methodology ............................................................... 13 3.2.1 Classification of Multiple Events and Conditions ................. 13 3.2.2 Consideration of Mode Changes During Classification........ 14 3.2.3 Classification of Imminent Conditions .................................. 14 3.2.4 Emergency Classification Level Upgrading and Downgrading ............................................................... 15 3.2.5 Classification of Short-Lived Events .................................... 15 3.2.6 Classification of Transient Conditions.................................. 15 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition ............................................................................. 16 3.2.8 Retraction of an Emergency Declaration ............................. 17 to NRC-15-0061 Page 2 Fermi 2 Emergency Action Level Technical Bases TABLE OF CONTENTS SECTION TITLE PAGE

4.0 REFERENCES

............................................................................................ 17 4.1 Developmental .................................................................................. 17 4.2 Implementing .................................................................................... 18 4.3 Commitments .................................................................................... 18 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...................................... 19 5.1 Definitions ......................................................................................... 19 5.2 Acronyms & Abbreviations ................................................................ 25 6.0 FERMI-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .......................... 28 7.0 ATTACHMENTS .......................................................................................... 32 7.1 Attachment 1 - Emergency Action Level Technical Bases ............... 33 Category R Abnormal Rad Levels / Rad Effluent ............................. 34 Category C Cold Shutdown / Refueling System Malfunction ........... 83 Category H Hazards and Other Conditions Affecting Plant Safety .. 135 Category S System Malfunction....................................................... 182 Category F Fission Product Barrier Degradation ............................. 239 Category E ISFSI ............................................................................. 246 7.2 Attachment 2 - Fission Product Barrier Matrix and Bases ................. 249 to NRC-15-0061 Page 3 Fermi 2 Emergency Action Level Technical Bases 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for the Enrico Fermi Unit 2 Power Plant (Fermi 2). It should be used to facilitate review of the Fermi 2 EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EP-101, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification. Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the Fermi 2 Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 Methodology for Development of Emergency Action Levels as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: to NRC-15-0061 Page 4 Fermi 2 Emergency Action Level Technical Bases Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs). Simplifying the fission product barrier EAL threshold for a Site Area Emergency. Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ADAMS Accession Number ML110240324 ML13091A209) (Ref. 4.1.1), Fermi 2 conducted an EAL implementation upgrade project that produced the EALs discussed herein. 2.2 Fission Product Barrier Thresholds Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier. The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. to NRC-15-0061 Page 5 Fermi 2 Emergency Action Level Technical Bases B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. C. Primary Containment (PC): The Primary Containment Barrier includes the drywell, the suppression pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier 2.4 EAL Organization The Fermi 2 EAL scheme includes the following features: Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode. to NRC-15-0061 Page 6 Fermi 2 Emergency Action Level Technical Bases o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency. Within each group, assignment of EALs to categories and subcategories: o Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The Fermi 2 EAL categories are aligned to and represent the NEI 99-01Recognition Categories. o Subcategories are used in the Fermi 2 scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The Fermi 2 EAL categories and subcategories are listed below. to NRC-15-0061 Page 7 Fermi 2 Emergency Action Level Technical Bases EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: R - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - ED Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI) Hot Conditions: S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown / Refueling System 1 - RPV Level Malfunction 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems to NRC-15-0061 Page 8 Fermi 2 Emergency Action Level Technical Bases The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information. 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, F or E)

2. Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1). If to NRC-15-0061 Page 9 Fermi 2 Emergency Action Level Technical Bases a category does not have a subcategory, this character is assigned the number one (1).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: A Plant-Specific basis section that provides Fermi-relevant information concerning the EAL. This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Fermi Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (Ref. 4.1.7) 1 Power Operation Mode Switch in Run to NRC-15-0061 Page 10 Fermi 2 Emergency Action Level Technical Bases 2 Startup Mode Switch in Startup/Hot Standby or Mode Switch Refuel with all reactor vessel head closure bolts fully tensioned. 3 Hot Shutdown Mode Switch in Shutdown and RCS temperature > 200°F 4 Cold Shutdown Mode Switch in Shutdown and RCS temperature 200°F. 5 Refueling Mode Switch in Shutdown or Refuel with one or more reactor vessel head closure bolts less then fully tensioned. D Defueled All nuclear fuel removed from reactor vessel (i.e., full core off load during refueling or extended outage). The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred. to NRC-15-0061 Page 11 Fermi 2 Emergency Action Level Technical Bases 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds. 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (Ref. 4.1.12). 3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment. to NRC-15-0061 Page 12 Fermi 2 Emergency Action Level Technical Bases 3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (Ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). to NRC-15-0061 Page 13 Fermi 2 Emergency Action Level Technical Bases 3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started. When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (Ref. 4.1.14). 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared. to NRC-15-0061 Page 14 Fermi 2 Emergency Action Level Technical Bases There is no additive effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (Ref. 4.1.2). 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. to NRC-15-0061 Page 15 Fermi 2 Emergency Action Level Technical Bases 3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (Ref. 4.1.2). 3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram. 3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the to NRC-15-0061 Page 16 Fermi 2 Emergency Action Level Technical Bases applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (Ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (Ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. to NRC-15-0061 Page 17 Fermi 2 Emergency Action Level Technical Bases 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (Ref. 4.1.3).

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML110240324 ML13091A209 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Fermi 2 Offsite Dose Calculation Manual (ODCM) TRM Volume III Figure 3.0-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents 4.1.7 Technical Specifications Table 1.1-1 Modes 4.1.8 Fermi 2 Radiological Emergency Response Preparedness Plan Figure J-1 Owner-Controlled AreaEP-101 Classification of Emergencies, Rev. 39 4.1.9 Technical Specifications Section 3.6 Containment Systems 4.1.10 UFSAR Figure 1.2-5 Site Plot Plan 4.1.11 WG-001 Fermi Writers Guide 4.1.12 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants to NRC-15-0061 Page 18 Fermi 2 Emergency Action Level Technical Bases 4.1.13 Fermi Certificate of Compliance No. 114 Appendix A Technical Specifications for the HI-STORM 100 Cask System 4.2 Implementing 4.2.1 EP-101Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to Fermi EAL Comparison Matrix 4.2.3 Fermi EAL Matrix 4.3 Commitments 4.3.1 None to NRC-15-0061 Page 19 Fermi 2 Emergency Action Level Technical Bases 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (Ref. 4.1.1 except as noted) Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. The Multi-Purpose Canister (MPC) serves as the Confinement Boundary for contained radioactive materials (Ref. 4.1.13) Containment Closure The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). Emergency Action Level A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual to NRC-15-0061 Page 20 Fermi 2 Emergency Action Level Technical Bases effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Unusual Event (UE), Alert, Site Area Emergency (SAE) and General Emergency (GE). EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires Fermi 2 to recommend protective actions for the general public to offsite planning agencies. Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE to NRC-15-0061 Page 21 Fermi 2 Emergency Action Level Technical Bases ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area) (Ref. 4.1.8). Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. to NRC-15-0061 Page 22 Fermi 2 Emergency Action Level Technical Bases Initiating Condition An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force. Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value. Owner Controlled Area Area depicted in the Fermi 2 Radiological Emergency Response Preparedness Plan Figure J-1 Owner-Controlled Area (Ref. 4.1.8). The company property immediately surrounding the PROTECTED AREA security fence. Access is normally limited to people on official business (Ref. 4.1.8). Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 4.1.10). RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway to NRC-15-0061 Page 23 Fermi 2 Emergency Action Level Technical Bases The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway. Safety System A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary. Site Boundary SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. (Ref. 4.1.6). Unisolable to NRC-15-0061 Page 24 Fermi 2 Emergency Action Level Technical Bases An open or breached system line that cannot be isolated, remotely or locally. Unplanned A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment. Visible Damage Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. to NRC-15-0061 Page 25 Fermi 2 Emergency Action Level Technical Bases 5.2 Acronyms & Abbreviations °F ........................................................................................................Degrees Fahrenheit ° ............................................................................................................................Degrees

................................................................................................................ Feet or Minutes
........................................................................................................... Inches or Seconds
........................................................................................................................... Gamma n ........................................................................................................................... Neutron AC ....................................................................................................... Alternating Current AOP .................................................................................. Abnormal Operating Procedure APRM ..................................................................... Average Power Range Meter Monitor ATWS ...................................................................... Anticipated Transient Without Scram BWR ............................................................................................... Boiling Water Reactor BWROG ................................................................. Boiling Water Reactor Owners Group CAHRRM............................................. Containment Area High Range Radiation Monitor CDE ....................................................................................... Committed Dose Equivalent CFR ...................................................................................... Code of Federal Regulations CS ................................................................................................................... Core Spray CW ......................................................................................................... Circulating Water DBA ................................................................................................Design Basis Accident DC ............................................................................................................... Direct Current EAL .............................................................................................Emergency Action Level ECCS ........................................................................... Emergency Core Cooling System ECL ................................................................................. Emergency Classification Level ED ......................................................................................................Emergency Director EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPIP ................................................................ Emergency Plan Implementing Procedure EPRI ............................................................................. Electric Power Research Institute ERG ............................................................................... Emergency Response Guideline ESF ................................................................................ Engineered Safeguards Feature FAQ ........................................................................................ Frequently Asked Question FEMA .............................................................. Federal Emergency Management Agency FSAR .................................................................................... Final Safety Analysis Report GE ..................................................................................................... General Emergency HCTL ............................................................................. Heat Capacity Temperature Limit HPCI ................................................................................ High Pressure Coolant Injection to NRC-15-0061 Page 26 Fermi 2 Emergency Action Level Technical Bases HSI ............................................................................................. Human System Interface IC .......................................................................................................... Initiating Condition ID .............................................................................................................. Inside Diameter IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI ........................................................... Independent Spent Fuel Storage Installation Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LOCA .........................................................................................Loss of Coolant Accident LPCI ................................................................................. Low Pressure Coolant Injection MCR ................................................................................................... Main Control Room MPC ................................... Maximum Permissible Concentration/Multi-Purpose Canister MPH .......................................................................................................... Miles Per Hour MSIV ...................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line mR, mRem, mrem, mREM ............................................... milli-Roentgen Equivalent Man MW .....................................................................................................................Megawatt NEI .............................................................................................. Nuclear Energy Institute NPP ...................................................................................................Nuclear Power Plant NRC ............................................................................... Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD .................................................. North American Aerospace Defense Command (NO)UE............................................................................. (Notification Of) Unusual Event NUMARC1 ................................................. Nuclear Management and Resources Council OBE ...................................................................................... Operating Basis Earthquake OCA .............................................................................................. Owner Controlled Area ODCM/ODAM.......................................... Offsite Dose Calculation (Assessment) Manual ORO ................................................................................ Off-site Response Organization PA ............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PRA/PSA ...................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PS ........................................................................................................ Protection System PSIG ................................................................................ Pounds per Square Inch Gauge R .........................................................................................................................Roentgen RB ........................................................................................................... Reactor Building RCC ........................................................................................... Reactor Control Console 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI). to NRC-15-0061 Page 27 Fermi 2 Emergency Action Level Technical Bases RCIC ................................................................................ Reactor Core Isolation Cooling RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RETS ......................................................... Radiological Effluent Technical Specifications RPS ......................................................................................... Reactor Protection System RPV ............................................................................................Reactor Pressure Vessel RSVR ................................................................................................................. Reservoir RW .................................................................................................................... Radwaste RWCU .......................................................................................... Reactor Water Cleanup SAR ............................................................................................... Safety Analysis Report SAS ......................................................................................... Safety Automation System SBO ......................................................................................................... Station Blackout SCBA ...................................................................... Self-Contained Breathing Apparatus SPDS ........................................................................... Safety Parameter Display System SRO ........................................................................................... Senior Reactor Operator TAF ...................................................................................................... Top of Active Fuel TB ............................................................................................................ Turbine Building TEDE ............................................................................... Total Effective Dose Equivalent TOAF .................................................................................................... Top of Active Fuel TSC ...........................................................................................Technical Support Center UFSAR Updated Final Safety Analysis Report to NRC-15-0061 Page 28 Fermi 2 Emergency Action Level Technical Bases 6.0 FERMI-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Fermi 2 EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Fermi 2 EALs based on the NEI guidance can be found in the EAL Comparison Matrix. Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 to NRC-15-0061 Page 29 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1 CA3.2 CA3 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2 3 to NRC-15-0061 Page 30 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 to NRC-15-0061 Page 31 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 to NRC-15-0061 Page 32 Fermi 2 Emergency Action Level Technical Bases 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis to NRC-15-0061 Page 33 Fermi 2 Emergency Action Level Technical Bases ATTACHMENT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES to NRC-15-0061 Page 34 Fermi 2 Emergency Action Level Technical Bases Category R - Abnormal Rad Levels / Rad Effluent EAL Group: ALL (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

to NRC-15-0061 Page 35 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. EAL: RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 36 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific Liquid Releases Fermi does not perform continuous radioactive liquid releases and no longer performs periodic batch radioactive liquid releases, per administrative controls. However, to provide EALs consistent with the template scheme, a liquid effluent EAL threshold has been developed. (Ref. 2) Per ODCM Figure 6.0-1, all sources of liquid effluent converge at a common discharge point prior to reaching the environment (Ref. 1). The D11-K604 Radiation Monitor on the liquid radwaste effluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 MPC at the discharge to Lake Erie. The monitor is located upstream of the Isolation Valve (G11-F733) on the liquid radwaste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir decant flow (Ref. 2). The Circulating Water Reservoir (CWR) Decant Line Radiation Monitor (D11-N402) and recorder (D11-R806) provides indication of the concentration of radioactive material in diluted radioactive liquid releases just before discharge to Lake Erie; and thus being the final monitor in the liquid discharge line is the liquid monitor used to address this EAL threshold. The value shown in Table R-1 column UE represents 2 times the ODCM limit of 1 MPC (Ref. 2). Gaseous Releases The column UE gaseous release values in Table R-1 represent 2 times the calculated release values associated with the ODCM limit for total body (the skin limit requires a higher release rate value) plus background. (Ref. 1, 2). For this initiating condition, the applicable effluent monitors are RB SPING, SGTS I SPING, SGTS II SPING, RW SPING and TB SPING (Ref. 2). to NRC-15-0061 Page 37 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. Escalation of the emergency classification level would be via IC RA1. Fermi Basis Reference(s):

1. Fermi Offsite Dose Calculation Manual

2. EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values Rev. 0

3. NEI 99-01 AU1 to NRC-15-0061 Page 38 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): None Basis: Site Specific None Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional to NRC-15-0061 Page 39 Fermi 2 Emergency Action Level Technical Bases releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RA1. Fermi Basis Reference(s):

1. Fermi Offsite Dose Calculation Manual

2. NEI 99-01 AU1 to NRC-15-0061 Page 40 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.1 Alert In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 41 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific Liquid Releases The RA1 IC addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Per ODCM Figure 6.0-1, all sources of liquid effluent converge at a common discharge point prior to reaching the environment (Ref. 1). The D11-K604 Radiation Monitor on the liquid radwaste effluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 MPC at the discharge to Lake Erie. The monitor is located upstream of the Isolation Valve (G11-F733) on the liquid radwaste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir decant flow (Ref. 2). The Circulating Water Reservoir (CWR) Decant Line Radiation Monitor (D11-N402) and recorder (D11-R806) provides indication of the concentration of radioactive material in diluted radioactive liquid releases just before discharge to Lake Erie; and thus being the final monitor in the liquid discharge line is the liquid monitor used to address this EAL threshold (Ref. 1, 2). The value shown in Table R-1 column UE Alert for liquid releases represents 10 mRem for one hour of exposure (Ref. 2). Gaseous Releases For gaseous releases, the preferred method for classification is by means of the computerized dose assessment program incorporating actual meteorology. This method is preferred since it eliminates uncertainty associated with assumed meteorology and source term data. For this initiating condition, the applicable gaseous effluent monitors are the Division I and II AXMs. to NRC-15-0061 Page 42 Fermi 2 Emergency Action Level Technical Bases The column ALERT gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 10 mRem TEDE or 50 mRem thyroid CDE (1% of the EPA PAGs) (Ref. 2). Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1. Fermi Basis Reference(s):

1. Fermi Offsite Dose Calculation Manual

2. EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values Rev. 0

3. NEI 99-01 AA1 to NRC-15-0061 Page 43 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.2 Alert Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4) Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific Calculated dose from airborne sources using computerized dose assessment model incorporating current meteorology indicates greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY (Ref. 1). Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). to NRC-15-0061 Page 44 Fermi 2 Emergency Action Level Technical Bases Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1. Fermi Reference(s):

1. EP-542 Computer-Based Offsite Dose Assessment - Airborne Release, Rev. 11

2. NEI 99-01 AA1 to NRC-15-0061 Page 45 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.3 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific Dose assessments based on liquid releases are manual calculations performed per Offsite Dose Calculation Manual (Ref. 1). Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and to NRC-15-0061 Page 46 Fermi 2 Emergency Action Level Technical Bases conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1. Fermi Reference(s):

1. Fermi 2 Offsite Dose Calculation Manual

2. NEI 99-01 AA1 to NRC-15-0061 Page 47 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.4 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 10 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific EP-220, Personnel Monitoring and Radiological Emergency Teams, provides guidance for emergency or post-accident radiological environmental monitoring (Ref. 1). Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits to NRC-15-0061 Page 48 Fermi 2 Emergency Action Level Technical Bases (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC RS1. Fermi Reference(s):

1. EP-220, Personnel Monitoring and Radiological Emergency Teams, Rev. 19

2. NEI 99-01 AA1 to NRC-15-0061 Page 49 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.1 Site Area Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "SAE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 50 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific For gaseous releases, the preferred method for classification is by means of the computerized dose assessment program incorporating actual meteorology and effluent monitor readings. This method is preferred since it eliminates uncertainty associated with assumed meteorology and source term data. For this initiating condition, the applicable gaseous effluent monitors are the Division I and II AXMs. The column SAE gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 100 mRem TEDE or 500 mRem thyroid CDE (10% of the EPA PAGs) (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RG1. to NRC-15-0061 Page 51 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values Rev. 0

2. NEI 99-01 AS1 to NRC-15-0061 Page 52 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.2 Site Area Emergency Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4) Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific Calculated dose from airborne sources using computerized dose assessment model incorporating current meteorology indicates greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the SITE BOUNDARY (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions to NRC-15-0061 Page 53 Fermi 2 Emergency Action Level Technical Bases alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RG1. Fermi Reference(s):

1. EP-542 Computer-Based Offsite Dose Assessment - Airborne Release, Rev. 11

2. NEI 99-01 AS1 to NRC-15-0061 Page 54 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE EAL:

RS1.3 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 100 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 500 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific EP-220, Personnel Monitoring and Radiological Emergency Teams, provides guidance for emergency or post-accident radiological environmental monitoring (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and to NRC-15-0061 Page 55 Fermi 2 Emergency Action Level Technical Bases conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC RG1. Fermi Reference(s):

1. EP-220, Personnel Monitoring and Radiological Emergency Teams, Rev. 19

2. NEI 99-01 AS1 to NRC-15-0061 Page 56 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.1 General Emergency In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "GE" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 57 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific For gaseous releases, the preferred method for classification is by means of the computerized dose assessment program incorporating actual meteorology and effluent monitor readings. This method is preferred since it eliminates uncertainty associated with assumed meteorology and source term data. For this initiating condition, the applicable gaseous effluent monitors are the Division I and II AXMs. The column GE gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 1000 mRem TEDE or 5000 mRem thyroid CDE (100% of the EPA PAGs) (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. to NRC-15-0061 Page 58 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values Rev. 0

2. NEI 99-01 AG1 to NRC-15-0061 Page 59 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.2 General Emergency Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the SITE BOUNDARY (Notes 3, 4) Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific Calculated dose from airborne sources using computerized dose assessment model incorporating current meteorology indicates greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions to NRC-15-0061 Page 60 Fermi 2 Emergency Action Level Technical Bases alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Fermi Reference(s):

1. EP-542 Computer-Based Offsite Dose Assessment - Airborne Release, Rev. 11

2. NEI 99-01 AG1 to NRC-15-0061 Page 61 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE EAL:

RG1.3 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY: Closed window dose rates > 1,000 mR/hr expected to continue for 60 min. Analyses of field survey samples indicate thyroid CDE > 5,000 mrem for 60 min. of inhalation. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific EP-220, Personnel Monitoring and Radiological Emergency Teams, provides guidance for emergency or post-accident radiological environmental monitoring (Ref. 1). Generic This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and to NRC-15-0061 Page 62 Fermi 2 Emergency Action Level Technical Bases conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Fermi Reference(s):

1. EP-220, Personnel Monitoring and Radiological Emergency Teams, Rev. 19

2. NEI 99-01 AG1 to NRC-15-0061 Page 63 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by any of the following: 2D1, FUEL POOL WATER LEVEL LOW alarm Floodup Level Transmitter (when in service) Visual observation AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors: RB5 Spent Fuel Pool ARM (Ch. 15) RB5 Refuel Floor Lo Range ARM (Ch. 17) RB5 Refuel Floor Hi Range ARM (Ch. 18) Mode Applicability: All Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway. Basis: Plant-Specific Indications of decreasing level include: Alarm 2D1, FUEL POOL WATER LEVEL LOW (Ref. 1, 2) Floodup Level Transmitter (when in service) to NRC-15-0061 Page 64 Fermi 2 Emergency Action Level Technical Bases During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. There is a Low Reactor Vessel/Fuel Pool Water Level Alarm that can be connected to these transmitters as well as a 5th Floor Alarm Unit that can be used to warn of the loss of shielding. (Ref. 3) Visual observation of reactor cavity and spent fuel pool level from the Refueling Floor Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment. Technical Specification LCO 3.7.7 requires at least 22 ft of water above irradiated fuel assemblies seated in the spent fuel pool storage racks. Technical Specification LCO 3.9.6 requires at least 20 ft 6 in. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations. This maintains sufficient water level in the fuel transfer canal, refueling cavity, and spent fuel pool to retain iodine fission product activity in the water in the event of a fuel handling accident (Ref. 4, 5). The spent fuel pool low level alarm is actuated by level switch G41-N001B four inches below normal level. (Ref. 2) Radiation monitors that may indicate a loss of shielding above irradiated fuel include (Ref. 6): RB5 Spent Fuel Pool ARM (Ch. 15) RB5 Fuel Floor Lo Range ARM (Ch. 17) RB5 Fuel Floor Hi Range ARM (Ch. 18) Generic This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. to NRC-15-0061 Page 65 Fermi 2 Emergency Action Level Technical Bases A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2. Fermi Reference(s):

1. AOP 20.708 Loss of FPCCU

2. ARP 2D1 Fuel Pool Water Level Low

3. SOP 23.800.06 Rev. 10 Reactor Vessel Water Level Monitoring During Refueling Operations

4. Technical Specifications LCO 3.7.7 Spent Fuel Pool water Level

5. Technical Specifications LCO 3.9.6 RPV Water Level

6. ARP 16D1 RB REFUELING FIFTH FLOOR HIGH RADN

7. NEI 99-01 AU2 to NRC-15-0061 Page 66 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above or damage to irradiated fuel EAL:

RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability: All Definition(s): REFUELING PATHWAY-. The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway. Basis: Plant-Specific Indications of decreasing water level with the potential to uncover irradiated fuel include: Floodup Level Transmitter (when in service) During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. Theres a Low Reactor Vessel/Fuel Pool Water Level Alarm that can be connected to these transmitters as well as a 5th Floor Alarm Unit that can be used to warn of the loss of shielding. (Ref. 1) Visual observation of reactor cavity and/or spent fuel pool level from the Refueling Floor Technical Specification LCO 3.7.7 requires at least 22 ft of water above irradiated fuel assemblies seated in the spent fuel pool storage racks. Technical Specification LCO 3.9.6 requires at least 20 ft 6 in. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations. This maintains sufficient water level in the fuel to NRC-15-0061 Page 67 Fermi 2 Emergency Action Level Technical Bases transfer canal, refueling cavity, and spent fuel pool to retain iodine fission product activity in the water in the event of a fuel handling accident. (Ref. 2, 3) Allowing level to decrease could result in spent fuel being uncovered, reducing spent fuel decay heat removal and creating an extremely hazardous radiation environment. Generic This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1.1. This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. to NRC-15-0061 Page 68 Fermi 2 Emergency Action Level Technical Bases A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RS1. Fermi Reference(s):

1. SOP 23.800.06 Rev. 10 Reactor Vessel Water Level Monitoring During Refueling Operations

2. Technical Specifications LCO 3.7.7 Spent Fuel Pool water Level

3. Technical Specifications LCO 3.9.6 RPV Water Level

4. NEI 99-01 AA2 to NRC-15-0061 Page 69 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above or damage to irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND Any of the following radiation monitor indications: RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm RBHVAC Vent Exhaust Radiation Monitor > 16,000 cpm Fuel Pool Vent Exhaust Radiation Monitor > 5 mR/hr Mode Applicability: All Definition(s): None Basis: Plant-Specific When considering escalation, information may come from: Radiation monitor readings Sampling and surveys Dose projections/calculations Reports from the scene regarding the extent of damage (e.g., refueling crew, radiation protection technicians) Radiation monitors and associated indications specified in this EAL are: RB5 Fuel Floor Hi Range ARM (Ch. 18) alarm: This high range area radiation monitor alarms at 1,000 mr/hr and provides confirming indication of possible damage to irradiated fuel (Ref. 1, 2, 3) to NRC-15-0061 Page 70 Fermi 2 Emergency Action Level Technical Bases RBHVAC Vent Exhaust Radiation Monitor > 16,000 cpm: This monitor provides indication of the release of radioactive fission products to the Reactor Building atmosphere as a result of damaged irradiated fuel. A reading of

       > 16,000 cpm requires entry into the Secondary Containment Control EOP (Ref. 4).

Fuel Pool Vent Exhaust Radiation Monitor > 5 mR/hr: This monitor also provides indication of the release of radioactive fission products to the Refueling Floor atmosphere (Reactor Building) as a result of damaged irradiated fuel. A reading of > 5 mR/hr requires entry into the Secondary Containment Control EOP (Ref. 4). Plant procedures require termination of fuel and core component movements and evacuation of the Reactor Building if elevated radiation levels are detected. All core alternations are stopped and transient fuel assemblies and core components are placed in a safe position in the reactor vessel, Spent Fuel Pool or fuel transfer canal to the extent practicable (Ref. 2, 3). Generic This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Escalation of the emergency would be based on either Recognition Category R or C ICs. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Escalation of the emergency classification level would be via IC RS1. to NRC-15-0061 Page 71 Fermi 2 Emergency Action Level Technical Bases Fermi Reference(s):

1. ARP 16D1 RB REFUELING FIFTH FLOOR HIGH RADN

2. AOP 20.710.01 Refueling Floor High Radiation 3 AOP 20.000.02 Abnormal Release of Radioactive Material

4. EOP 29.100.01 SH5 Secondary Containment and Rad Release

5. NEI 99-01 AA2 to NRC-15-0061 Page 72 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above or damage to irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to Level 2 as indicated by level < 33 ft. on G41R601A/B. Mode Applicability: All Definition(s): None Basis: Plant-Specific The Fermi SFP is located on the Reactor Building 5th floor. The surface of the water is normally maintained at plant elevation 683.5 ft. (Level 1) by scuppers that act as skimmers and wave suppressors. This results in a minimum water depth of 7 ft. above the top of the fuel while it is being moved above storage racks. Pool water level indication is painted on the north and east walls of the pool starting at 18 ft. above the stored fuel assemblies (Ref. 1). Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level at which radiation level would still allow personnel access near the pool, 18 ft. above the top of the fuel racks (Level 2 or ele. 679 ft. 1/8 in.) and SFP level at the top of the fuel racks (Level 3 or ele. 661 ft. 1/8 in.). Remote SFP level indication is available in the control room on level indicator G41R601A Panel H11P601. An indicated level of 33 ft. corresponds to the Level 2 setpoint (Ref. 2). Generic This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These to NRC-15-0061 Page 73 Fermi 2 Emergency Action Level Technical Bases events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC RS1. Fermi Reference(s):

1. USFSAR Section 9.1.2.2.1

2. Engineering Design Package (EDP) 37088

3. NEI 99-01 AA2 to NRC-15-0061 Page 74 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

EAL: RA3.1 Alert Dose rate > 15 mR/hr in EITHER of the following areas: AB3 Control Room (ARM Channel 6) Central Alarm Station (by survey) Mode Applicability: All Definition(s): None Basis: Plant-Specific ARM Channel 6 (D21-N106) is the permanently installed Control Room area radiation monitor and, along with local radiation surveys, may be used to assess this EAL threshold (Ref. 1). Permanently installed area radiation monitoring is not installed in the CAS and, therefore, radiation levels in this area must be assessed with local radiation survey techniques Generic This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. to NRC-15-0061 Page 75 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. SOP 23.611 Area Radiation Monitoring System

2. NEI 99-01 AA3 to NRC-15-0061 Page 76 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 3 - Area Radiation Levels Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.

EAL: RA3.2 Alert An UNPLANNED event results in radiation levels that prohibit or impede access to any Table R-2 rooms or areas (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table R-2 Safe Shutdown Rooms/Areas Room/Area Mode Applicability Relay Room All AB3-DC MCC Area Mode 3 RB1-F17 Mode 3 RB1-F11 Mode 3 Mode Applicability: All Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific The rooms/areas and associated mode applicability specified in Table R-2 are those that contain equipment which require a manual/local action as specified in operating procedures used for normal operation, cooldown and shutdown. This table excludes rooms/areas that may have procedurally directed actions that are of an administrative nature (normal rounds or routine inspections) or that are not crucial to the conduct of safe operation, cooldown and shutdown. to NRC-15-0061 Page 77 Fermi 2 Emergency Action Level Technical Bases Specifically: Control Room & Relay Room in all modes (Control Room is not included as it is addressed in RA3.1) AB3-DC MCC Area - Access is required when in Mode 3 to install power fuses and close the MCC for E1150-F008 which must be performed to align shutdown cooling suction path. RB1-F17 - Access is required when in Mode 3 to place the permissive switch in OPERATE for E11F610A if Div 1 RHR is being placed in shutdown cooling. This step must be performed to warmup the shutdown cooling piping. RB1-F11 - Access is required when in Mode 3 to place the permissive switch in OPERATE for E11F610B if Div 2 RHR is being placed in shutdown cooling. This step must be performed to warmup the shutdown cooling piping. (Ref. 1, 2, 3). Generic This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable. For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply. to NRC-15-0061 Page 78 Fermi 2 Emergency Action Level Technical Bases The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4. The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.). The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections). The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. Fermi Basis Reference(s):

1. Operating Procedures (normal plant operations, cooldown or shutdown), manual / local actions:

a. 22.000.03 - Power Operation 25% to 100% to 25%

b. 22.000.04 - Plant Shutdown From 25% Power

c. 22.000.05 - Pressure/Temp Monitoring During Heatup and Cooldown

d. 23.202 - High Pressure Coolant Injection System

e. 23.205 - Residual Heat Removal System

f. 23.206 - Reactor Core Isolation Cooling System

g. 23.427 - Primary Containment Isolation System

h. 23.610 - Reactor Protection System (RPS)

i. MGA03 - Procedure Use and Adherence

2. GOP 22.000.03 Power Operation 25% to 100% to 25%

3. GOP 22.000.04 Plant Shutdown from 25% Power

42. NEI 99-01 AA3 to NRC-15-0061 Page 79 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to Level 3 as indicated by level < 0 ft. 3 in. on G41R601A/B. Mode Applicability: All Definition(s): None Basis: Plant-Specific The Fermi SFP is located on the Reactor Building 5th floor. The surface of the water is normally maintained at plant elevation 683.5 ft. (Level 1) by scuppers that act as skimmers and wave suppressors. This results in a minimum water depth of 7 ft. above the top of the fuel while it is being moved above storage racks. Pool water level indication is painted on the north and east walls of the pool starting at 18 ft. above the stored fuel assemblies (Ref. 1). Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level at which radiation level would still allow personnel access near the pool, 18 ft. above the top of the fuel racks (Level 2 or ele. 679 ft. 1/8 in.) and SFP level at the top of the fuel racks (Level 3 or ele. 661 ft. 1/8 in.). Remote SFP level indication is available in the control room on level indicator G41R601A Panel H11P601. An indicated level of 0 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 2). Generic This IC addresses a significant loss of spent fuel pool inventory control and makeup to NRC-15-0061 Page 80 Fermi 2 Emergency Action Level Technical Bases capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RG1 or RG2. Fermi Reference(s):

1. USFSAR Section 9.1.2.2.1

2. Engineering Design Package (EDP) 37088

3. NEI 99-01 AS2 to NRC-15-0061 Page 81 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least Level 3 as indicated by level > 0 ft. 3 in. on G41R601A/B for > 60 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: All Definition(s): None Basis: Plant-Specific The Fermi SFP is located on the Reactor Building 5th floor. The surface of the water is normally maintained at plant elevation 683.5 ft. (Level 1) by scuppers that act as skimmers and wave suppressors. This results in a minimum water depth of 7 ft. above the top of the fuel while it is being moved above storage racks. Pool water level indication is painted on the north and east walls of the pool starting at 18 ft. above the stored fuel assemblies (Ref. 1). Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level at which radiation level would still allow personnel access near the pool, 18 ft. above the top of the fuel racks (Level 2 or ele. 679 ft. 1/8 in.) and SFP level at the top of the fuel racks (Level 3 or ele. 661 ft. 1/8 in.). Remote SFP level indication is available in the control room on level indicator G41R601A Panel H11P601. An indicated level of 0 ft. 3 in. corresponds to the Level 3 setpoint (Ref. 2). to NRC-15-0061 Page 82 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. Fermi Reference(s):

1. USFSAR Section 9.1.2.2.1

2. Engineering Design Package (EDP) 37088

3. NEI 99-01 AG2 to NRC-15-0061 Page 83 Fermi 2 Emergency Action Level Technical Bases Category C - Cold Shutdown / Refueling System Malfunction EAL Group: Cold Conditions (RCS temperature 200ºF);

EALs in this category are applicable only in one or more cold operating modes. Category C EALs are directly associated with cold shutdown or refueling system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and refueling system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, containment closure, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refueling, D - Defueled). The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure Vessel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity.

2. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4160 VAC essential buses 64B/64C and 65E/65F.

3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power to NRC-15-0061 Page 84 Fermi 2 Emergency Action Level Technical Bases Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 130 VDC ESF buses.

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

6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification.

to NRC-15-0061 Page 85 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer. EAL: CU1.1 Unusual Event UNPLANNED loss of reactor coolant results in RPV water level below the established control band for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific With the plant in Cold Shutdown, RPV water level is required to be maintained above 214 in. and normally maintained in a level band of 220 to 255 in. above TAF (Ref. 1). However, if RPV level is being controlled below the normal band, or if level is being maintained in a designated band in the reactor vessel it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. With the plant in Refueling mode, RPV water level is normally maintained at or above the reactor vessel flange (Technical Specification LCO 3.9.6 requires at least 20 ft 6 in. of water above the top of the reactor vessel flange in the refueling cavity during refueling operations) (Ref. 4). The reactor vessel flange mating surface is 379 in. above TAF (Ref. 2). RPV level can be monitored by one or more of the following (Ref. 3): Flood-up Level indicator LIEB21-R605 (+160 in. to +560 in.) to NRC-15-0061 Page 86 Fermi 2 Emergency Action Level Technical Bases Wide Range Level indicators LIB21-R604A/B (+10 in. to +220 in.) Narrow Range Level indicators LIC32-R606A/B/C/D (+160 in. to +220 in.) Floodup Level Transmitter (when in service) During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. (Ref. 2) Visual observation of reactor cavity level from the Refueling Floor or by remote video display (when available) Regardless of where RPV level is intentionally being controlled, either above or below the reactor vessel flange, as in Cold Shutdown, it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. Generic This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band). This condition is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. to NRC-15-0061 Page 87 Fermi 2 Emergency Action Level Technical Bases The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3. Fermi Basis Reference(s):

1. GOP 22.000.04 Plant Shutdown From 25% Power

2. SOP 23.800.06 Reactor Vessel Water Level Monitoring During Refueling Operations

3. UFSAR Section 7 Instrumentation and Controls Table 7.5-1 Control Room Level Indication

4. Technical Specification LCO 3.9.6 RPV Water Level

5. NEI 99-01 CU1 to NRC-15-0061 Page 88 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer EAL:

CU1.2 Unusual Event RPV water level cannot be monitored AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Table C-1 Sumps & Tanks Drywell Floor Drain Sump Drywell Equipment Drain Sump RB Floor Drain Sumps RB Equipment Drain Sumps Torus Visual Observation Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific RPV level can be monitored by one or more of the following (Ref. 1): Flood-up Level indicator B21LIE-R605 (+160 in. to +560 in.) Wide Range Level indicators B21LI-R604A/B (+10 in. to +220 in.) Narrow Range Level indicators C32LI-R606A/B/C/D (+160 in. to +220 in.) to NRC-15-0061 Page 89 Fermi 2 Emergency Action Level Technical Bases Floodup Level Transmitter (when in service) During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. (Ref. 3) Visual observation of reactor cavity level from the Refueling Floor In this EAL, all water level indication is unavailable, and the RCS inventory loss must be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess RCS leakage (Ref. 2). Generic This IC addresses a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. This condition is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3. Fermi Basis Reference(s):

1. UFSAR Section 7 Instrumentation and Controls Table 7.5-1 Control Room Level Indication

2. UFSAR Section 5.2.7 Reactor Coolant Pressure Boundary Leak Detection System to NRC-15-0061 Page 90 Fermi 2 Emergency Action Level Technical Bases

3. SOP 23.800.06 Rev. 10 Reactor Vessel Water Level Monitoring During Refueling Operations

4. NEI 99-01 CU1 to NRC-15-0061 Page 91 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA1.1 Alert Loss of RPV inventory as indicated by RPV water level < 111 in. above TAF (Level 2) Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: Plant-Specific When reactor vessel water level drops to 110.8 in. (rounded to 111in.) above TAF high pressure steam-driven injection sources HPCI (ECCS) and RCIC actuate (Ref. 1). Although these systems cannot restore RCS inventory in the cold condition, the Low-Low (Level 2) ECCS actuation setpoint is operationally significant and is indicative of a loss of RCS inventory significantly below the normally established control band specified in CU1.1. Generic This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, a lowering of water level below 111 in above TAF indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. to NRC-15-0061 Page 92 Fermi 2 Emergency Action Level Technical Bases Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3. If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Fermi Basis Reference(s):

1. TRM Table TR3.3.5.1-1 EmegrnecyEmergency Core Cooling System Instrumentation

2. NEI 99-01 CA1 to NRC-15-0061 Page 93 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory EAL:

CA1.2 Alert RPV water level cannot be monitored for 15 min. (Note 1) AND UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-1 Sumps & Tanks Drywell Floor Drain Sump Drywell Equipment Drain Sump RB Floor Drain Sumps RB Equipment Drain Sumps Torus Visual Observation Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific RPV level can be monitored by one or more of the following (Ref. 1): Flood-up Level indicator LIEB21-R605 (+160 in. to +560 in.) to NRC-15-0061 Page 94 Fermi 2 Emergency Action Level Technical Bases Wide Range Level indicators LIB21-R604A/B (+10 in. to +220 in.) Narrow Range Level indicators LIC32-R606A/B/C/D (+160 in. to +220 in.) Floodup Level Transmitter (when in service) During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. (Ref. 3) Visual observation of reactor cavity level from the Refueling Floor In this EAL, all water level indication is unavailable for greater then 15 minutes, and the RCS inventory loss must be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess RCS leakage (Ref. 2). Generic This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1. If the RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1. Fermi Basis Reference(s):

1. UFSAR Section 7 Instrumentation and Controls Table 7.5-1 Control Room Level Indication to NRC-15-0061 Page 95 Fermi 2 Emergency Action Level Technical Bases

2. UFSAR Section 5.2.7 Reactor Coolant Pressure Boundary Leak Detection System

3. SOP 23.800.06 Rev. 10 Reactor Vessel Water Level Monitoring During Refueling Operations

4. NEI 99-01 CA1 to NRC-15-0061 Page 96 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1.1 Site Area Emergency CONTAINMENT CLOSURE not established AND RPV water level < 32 in. above TAF (Level 1) Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). Basis: Plant-Specific When reactor vessel water level drops to 31.8 in. (rounded to 32 in.) above TAF low pressure ECCS such as LPCI and Core Spray actuate (Ref. 1). The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and potential loss of the Fuel Clad barrier. to NRC-15-0061 Page 97 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RPV levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1. Fermi Basis Reference(s):

1. TRM Table TR3.3.5.1-1 EmegrnecyEmergency Core Cooling System Instrumentation

2. NEI 99-01 CS1 to NRC-15-0061 Page 98 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1.2 Site Area Emergency CONTAINMENT CLOSURE established AND RPV water level < 0 in. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). Basis: Plant-Specific When RPV water level drops to 0 in. (TAF) core uncovery is about to occur (Ref. 1). Generic This IC addresses a significant and prolonged loss of RPV level control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. to NRC-15-0061 Page 99 Fermi 2 Emergency Action Level Technical Bases Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of CS1.1 and CS1.2 reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. NEI 99-01 CS1 to NRC-15-0061 Page 100 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RCS inventory affecting core decay heat removal capability EAL:

CS1.3 Site Area Emergency RPV water level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by EITHER of the following: RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-1 Sumps & Tanks Drywell Floor Drain Sump Drywell Equipment Drain Sump RB Floor Drain Sumps RB Equipment Drain Sumps Torus Visual Observation Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: to NRC-15-0061 Page 101 Fermi 2 Emergency Action Level Technical Bases Plant-Specific If all means of level monitoring are not available, the RCS inventory loss may be detected by the Fuel Floor area radiation monitors or indication or sump/tank level increases: In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. RB5 Fuel Floor Hi Range ARM (Ch. 18) is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred. D11-N443A/B are the Containment High Range Radiation Monitors but they are not located in the Containment with sufficient line-of-sight to the irradiated fuel in the reactor vessel to be of use in detecting loss of inventory above the core (Ref. 1, 2). If water level monitoring capability is unavailable, the reactor vessel inventory loss may be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess RCS leakage (Ref. 3). Generic This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to to NRC-15-0061 Page 102 Fermi 2 Emergency Action Level Technical Bases account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1. Fermi Basis Reference(s):

1. ARP 16D1 RB REFUELING FIFTH FLOOR HIGH RADN

2. AOP 20.710.01 Refueling Floor High Radiation

3. UFSAR Section 5.2.7 Reactor Coolant Pressure Boundary Leak Detection System

4. NEI 99-01 CS1 to NRC-15-0061 Page 103 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:

CG1.1 General Emergency RPV water level < 0 in. for > 30 min. (Note 1) AND Any of the following indications of containment challenge: CONTAINMENT CLOSURE not established (Note 6) Primary Containment hydrogen concentration > 6% UNPLANNED increase in Primary Containment pressure Exceeding one or more Secondary Containment Control Max Safe Operating Area Radiation Levels (EOP Table 14) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). UNPLANNED-. A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. to NRC-15-0061 Page 104 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific When RPV water level drops to 0 in. (TAF) core uncovery is about to occur (Ref. 1). Four indications are associated with a challenge to Containment: CONTAINMENT CLOSURE is not established. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6% by volume in the presence of oxygen (>5%) (Ref. 1). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. An unplanned pressurization that can breach the containment barrier signifies a challenge to the Primary Containment pressure retaining capability which is dependent on the status of either containment integrity or CONTAINMENT CLOSURE. If containment integrity is established for full power operation, a breach could occur if the design Primary Containment pressure is exceeded (62 psig) (Ref. 2). For this condition, a small unplanned pressure rise above atmospheric pressure does not challenge containment. If in refueling operations, however, a breach could occur if the unplanned pressure rise exceeded the capability of a temporary containment seal used to establish CONTAINMENT CLOSURE. This would occur at a much lower pressure than the containment design pressure. The use of secondary containment radiation monitors provides indication of increased release that may be indicative of a challenge to Primary Containment. The Secondary Containment Control EOP Max Safe area radiation values have been selected because these values are easily recognizable and have a defined basis. (Ref. 1, 3) to NRC-15-0061 Page 105 Fermi 2 Emergency Action Level Technical Bases The only Secondary Containment Maximum Safe Operating Radiation Level that can be determined remotely in the Control Room is the RBSB Torus Room on ARM Channel 14. No other Secondary Containment Maximum Safe Operating Radiation Levels (> 5 R/hr) can be determinedetermined by installed area radiation monitors due to instrument range limitations. Therefore the area radiation threshold (other than for the RBSB Torus Room) for the Primary Containment PotentalPotential Loss based on RCS leak rate must be determined via local survey. (ref. 1, 3) Generic This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as to NRC-15-0061 Page 106 Fermi 2 Emergency Action Level Technical Bases ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. UFSAR Table 6.2-1 Containment Parameters

3. EOP 29.100.01 Sheet 5 Secondary Containment and Rad Release

4. NEI 99-01 CG1 to NRC-15-0061 Page 107 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged EAL:

CG1.2 General Emergency RPV water level cannot be monitored for 30 min. (Note 1) AND Core uncovery is indicated by EITHER of the following: RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory AND Any of the following indications of containment challenge: CONTAINMENT CLOSURE not established (Note 6) Primary Containment hydrogen concentration > 6% UNPLANNED increase in Primary Containment pressure Exceeding one or more Secondary Containment Control Max Safe Operating Area Radiation Levels (EOP Table 14) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required. Table C-1 Sumps & Tanks Drywell Floor Drain Sump Drywell Equipment Drain Sump RB Floor Drain Sumps RB Equipment Drain Sumps Torus Visual Observation Mode Applicability: 4 - Cold Shutdown, 5 - Refueling to NRC-15-0061 Page 108 Fermi 2 Emergency Action Level Technical Bases Definition(s): CONTAINMENT CLOSURE - The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific If all means of level monitoring are not available, the RCS inventory loss may be detected by the Fuel Floor area radiation monitors or indication or sump/tank level increases: In the Refueling Mode, as water level in the reactor vessel lowers, the dose rate above the core will increase. The dose rate due to this core shine should result in indications on installed area radiation monitors. RB5 Fuel Floor Hi Range ARM (Ch. 18) is located on the Refuel Floor in the Reactor Building and is designed to provide monitoring of radiation due to a fuel handling event or loss of shielding during refueling operations. If this radiation monitor reaches and exceeds the alarm setpoint of 1,000 mr/hr, a loss of inventory with potential to uncover the core is likely to have occurred. D11-N443A/B are the Containment High Range Radiation Monitors but they are not located in the Containment with sufficient line-of-sight to the irradiated fuel in the reactor vessel to be of use in detecting loss of inventory above the core (Ref. 1, 2). If water level monitoring capability is unavailable, the reactor vessel inventory loss may be detected by sump or tank level changes (Table C-1). Plant design and procedures provide the capability to detect and assess RCS leakage (Ref. 3). to NRC-15-0061 Page 109 Fermi 2 Emergency Action Level Technical Bases Four indications are associated with a challenge to Containment: CONTAINMENT CLOSURE is not established. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gases in the Primary Containment. However, Primary Containment monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists. An explosive mixture can be formed when hydrogen gas concentration in the Primary Containment atmosphere is greater than 6% by volume in the presence of oxygen (>5%) (Ref. 4). In Cold Shutdown and Refueling modes it is assumed that the Primary Containment is de-inerted. An unplanned pressurization that can breach the containment barrier signifies a challenge to the Primary Containment pressure retaining capability which is dependent on the status of either containment integrity or CONTAINMENT CLOSURE. If containment integrity is established for full power operation, a breach could occur if the design Primary Containment pressure is exceeded (62 psig) (Ref. 5). For this condition, a small unplanned pressure rise above atmospheric pressure does not challenge containment. If in refueling operations, however, a breach could occur if the unplanned pressure rise exceeded the capability of a temporary containment seal used to establish CONTAINMENT CLOSURE. This would occur at a much lower pressure than the containment design pressure. The use of secondary containment radiation monitors provides indication of increased release that may be indicative of a challenge to Primary Containment. The Secondary Containment Control EOP Max Safe area radiation values have been selected because these values are easily recognizable and have a defined basis. (Ref. 4, 6) The only Secondary Containment Maximum Safe Operating Radiation Level that can be determined remotely in the Control Room is the RBSB Torus Room on ARM Channel 14. No other Secondary Containment Maximum Safe Operating Radiation Levels (> 5 R/hr) to NRC-15-0061 Page 110 Fermi 2 Emergency Action Level Technical Bases can be determinedetermined by installed area radiation monitors due to instrument range limitations. Therefore the area radiation threshold (other than for the RBSB Torus Room) for the Primary Containment PotentalPotential Loss based on RCS leak rate must be determined via local survey. Generic This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. to NRC-15-0061 Page 111 Fermi 2 Emergency Action Level Technical Bases The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Fermi Basis Reference(s):

1. ARP 16D1 RB REFUELING FIFTH FLOOR HIGH RADN

2. AOP 20.710.01 Refueling Floor High Radiation

3. UFSAR Section 5.2.7 Reactor Coolant Pressure Boundary Leak Detection System

4. EOP Support Documentation Section 1 Plant Specific Technical Guideline

5. UFSAR Section 6.2.1.2.1 Primary Containment

6. EOP 29.100.01 Sheet 5 Secondary Containment and Rad Release

7. NEI 99-01 CG1 to NRC-15-0061 Page 112 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer.

EAL: CU2.1 Unusual Event AC power capability to 4160V essential Division I (64B/64C) and Division II (65E/65F) reduced to a single power source (Table C-2) for 15 min. (Note 1, 10) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table C-2 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; to NRC-15-0061 Page 113 Fermi 2 Emergency Action Level Technical Bases (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Plant-Specific Table C-2 lists AC sources capable of powering essential AC buses. For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it. This EAL is indicated by the loss of all but one AC power source to 4160V essential buses 64B/64C (DiviaionDivision 1) and 65E/65F (Division II) for greater then or equal to 15 minutes such that a loss of any additional source will result in a complete loss of AC power to essential busses. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) as an onsite AC power supply only if it is already aligned to and capable of powering one of the essential 4160 V divisions within the 15 minute time criteria (Ref. 2). to NRC-15-0061 Page 114 Fermi 2 Emergency Action Level Technical Bases The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of essential bus power is not restored within 15 minutes, an Unusual Event is declared under this EAL. This cold condition EAL is equivalent to the hot condition loss of all offsite AC power EAL SA1.1. Generic This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An AC power source is a source recognized in AOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of one division of essential power sources (e.g., onsite diesel generators). A loss of essential power sources (e.g., onsite diesel generators) with a single division of essential buses being back-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. NEI 99-01 CU2 to NRC-15-0061 Page 115 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 2 - Loss of Essential AC Power Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer.

EAL: CA2.1 Alert Loss of all offsite and all onsite AC power capability (Table C-2) to 4160V essential Division I (64B/64C) and Division II (65E/65F) for 15 min. (Note 1, 10) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table C-2 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled Definition(s): None Basis: Plant-Specific Table C-2 lists AC sources capable of powering essential AC divisions. For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it. to NRC-15-0061 Page 116 Fermi 2 Emergency Action Level Technical Bases This EAL is indicated by the loss of all offsite and onsite AC power capability to 4160V essential buses 64B/64C (Division I) and 65E/65F (Division II) for greater then or equal to 15 minutes. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) as an onsite AC power supply only if it is already aligned to and capable of powering one of the essential 4160 V divisions within the 15 minute time criteria (Ref. 2). This EAL is the cold condition equivalent of the hot condition loss of all AC power EAL SS1.1. Generic This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. to NRC-15-0061 Page 117 Fermi 2 Emergency Action Level Technical Bases When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an essential bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or RS1. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. NEI 99-01 CA2 to NRC-15-0061 Page 118 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature.

EAL: CU3.1 Unusual Event UNPLANNED increase in RCS temperature to > 200°F Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific Several instruments are capable of providing indication of RCS temperature (Ref. 2) with respect to the Technical Specification cold shutdown temperature limit (200°F, Ref. 1): Primary: Recirc Loop A(B) Suction Temperature - B31-R650 (H11-P603) Reactor Vessel Shell Temperature - B21-R007 (H11-P603) Reactor Vessel Bottomhead Drain - G33-N601 (H11-P602) (drain line must have flow) Alternate: RHR A/B HX Inlet - E11-R601A/B Reactor Vessel Shell Temperature - B21-R007 (H11-P603) Generic This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limitandlimit and represents a potential to NRC-15-0061 Page 119 Fermi 2 Emergency Action Level Technical Bases degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Fermi Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. GOP 22.000.05 Pressure/Temperature Monitoring During Heatup and Cooldown

3. NEI 99-01 CU3 to NRC-15-0061 Page 120 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature.

EAL: CU3.2 Unusual Event Loss of all RCS temperature and RPV level indication for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: Plant-Specific Several instruments are capable of providing indication of RCS temperature (Ref. 2) with respect to the Technical Specification cold shutdown temperature limit (200°F, Ref. 1): Primary: o Recirc Loop A(B) Suction Temperature - B31-R650 (H11-P603) o Reactor Vessel Shell Temperature - B21-R007 (H11-P603) o Reactor Vessel Bottomhead Drain - G33-N601 (H11-P602) (drain line must have flow) Alternate: o RHR A/B HX Inlet - E11-R601A/B o Reactor Vessel Shell Temperature - B21-R007 (H11-P603) RPV level can be monitored by one or more of the following (Ref. 3): Flood-up Level indicator LIEB21-R605 (+160 in. to +560 in.) Wide Range Level indicators LIB21-R604A/B (+10 in. to +220 in.) to NRC-15-0061 Page 121 Fermi 2 Emergency Action Level Technical Bases Narrow Range Level indicators LIC32-R606A/B/C/D (+160 in. to +220 in.) Floodup Level Transmitter (when in service) During refueling operations with the fuel pool gates removed, the RPV floodup level instrumentation (B21-N027) and the Rx Vessel Core Plate dp transmitter (B21-N032) are capable of displaying the common level of the reactor cavity and the spent fuel pool. (Ref. 4) Visual observation of reactor cavity level from the Refueling Floor Generic This EAL addresses the inability to determine RCS temperature and RPV level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria. Fermi Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. GOP 22.000.05 Pressure/Temperature Monitoring During Heatup and Cooldown

3. UFSAR Section 7 Instrumentation and Controls Table 7.5-1 Control Room Level Indication

4. SOP 23.800.06 Rev. 10 Reactor Vessel Water Level Monitoring During Refueling Operations

5. NEI 99-01 CU3 to NRC-15-0061 Page 122 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain the plant in cold shutdown.

EAL: CA3.1 Alert UNPLANNED increase in RCS temperature to > 200°F for > Table C-3 duration (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table C-3: RCS Heat-up Duration Thresholds Containment Closure RCS Status Heat-up Duration Status Intact N/A 60 min.* established 20 min.* Not intact not established 0 min.

If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): CONTAINMENT CLOSURE - The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. to NRC-15-0061 Page 123 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Several instruments are capable of providing indication of RCS temperature (Ref. 2) with respect to the Technical Specification cold shutdown temperature limit (200°F, Ref. 1): Primary: Recirc Loop A(B) Suction Temperature - B31-R650 (H11-P603) Reactor Vessel Shell Temperature - B21-R007 (H11-P603) Reactor Vessel Bottomhead Drain - G33-N601 (H11-P602) (drain line must have flow) Alternate: RHR A/B HX Inlet - E11-R601A/B Reactor Vessel Shell Temperature - B21-R007 (H11-P603) Generic This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact. The 20-minute to NRC-15-0061 Page 124 Fermi 2 Emergency Action Level Technical Bases criterion was included to allow time for operator action to address the temperature increase. The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Primary Containment or Reactor Building atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. Escalation of the emergency classification level would be via IC CS1 or RS1. Fermi Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. GOP 22.000.05 Pressure/Temperature Monitoring During Heatup and Cooldown

3. NEI 99-01 CA3 to NRC-15-0061 Page 125 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: Inability to maintain the plant in cold shutdown.

EAL: CA3.2 Alert UNPLANNED RPV pressure increase > 10 psig Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): UNPLANNED- A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific A 10 psig RPV pressure increase can be monitored on various indicators such as C32-R609 (H11-P603) (Ref. 1). Generic This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. This EAL provides a pressure-based indication of RCS heat-up. Escalation of the emergency classification level would be via IC CS1 or RS1. Fermi Basis Reference(s):

1. GOP 22.000.05 Pressure/Temperature Monitoring During Heatup and Cooldown

2. NEI 99-01 CA3 to NRC-15-0061 Page 126 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of Vital DC power for 15 minutes or longer.

EAL: CU4.1 Unusual Event Degraded voltage (< 105 VDC) on required 130 VDC system vital buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: Plant-Specific As used in this EAL, required means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. The fifteen minute interval is intended to exclude transient or momentary power losses. At Fermi 2, the vital 260/130 VDC System ensures power is available for the reactor to be shutdown safely and maintained in a safe condition. The vital DC system is divided into two independent divisions - Division I and Division II - with separate DC power supplies. These power supplies consist of two separate 260/130V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI. The system provides sufficient capacity, via each of the Class 1E DC batteries, to power all required loads for 4 hours following a loss of AC power supply (Ref. 1). Based on Technical Specifications Bases Section B.3.8.4, the 130 VDC battery minimum design voltage limit is 105 VDC. (Ref. 2). This EAL is the cold condition equivalent of the hot condition loss of DC power EAL SS2.1. to NRC-15-0061 Page 127 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, required means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Division I is out-of-service (inoperable) for scheduled outage maintenance work and Division II is in-service (operable), then a loss of Vital DC power affecting Division II would require the declaration of an Unusual Event. A loss of Vital DC power to Division I would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R. Fermi Basis Reference(s):

1. Design Bases Document R32-00 DC Electrical System

2. Technical Specifications Bases Section B.3.8.4 DC Sources - Operating

3. NEI 99-01 CU4 to NRC-15-0061 Page 128 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities.

EAL: CU5.1 Unusual Event Loss of all Table C-4 onsite communication methods OR Loss of all Table C-4 offsite communication methods OR Loss of all Table C-4 NRC communication methods Table C-4 Communication Methods System Onsite Offsite NRC Administrative Telephones X X X RERP Emergency Telephones X X X Satellite Phones X X Federal Telephone System (ENS) X X Automatic Ring Lines X MI State Radios (800 MHz) X X Plant Radio System X Hi-Com (PA System) X Mode Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled Definition(s): None to NRC-15-0061 Page 129 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific The Table C-4 list for onsite communications loss encompasses the loss of all means of routine communications (e.g., administrative and internal telephones, plant page [Hi-Com] and plant radios) (Ref. 1, 2). The Table C-4 list for offsite communications loss encompasses the loss of all means of communications with offsite authorities. This includes the RERP telephone dedicated ring lines, backup phone systems administrative telephone lines, satellite, and FTS (ENS) which can be utilized as a regular telephone (Ref. 1, 2). The Table C-4 list for NRC communications loss encompasses the loss of all means of communications with the NRC. This includes the FTS (ENS), backup phone systems (administrative telephone lines, RERP phones and satellite) (Ref. 1, 2). The communications methods used at Fermi 2 are described in the RERP Plan (Ref. 1). The radio network at Fermi 2 involves several radio systems to effect communicaitonscommunications within the plant with damage control teams, rescue teams, fire brigade, radiological monitoring teams, and security personnel as well as provide backup comminicationscommunications to essential Offsite Emergency Response Organizations (OROs) in the event of telephone equipment malfunction. There are two radio consoles normally used in the Control Room. One is installed in panel H11-P700 to establish communications using plant radio zone 1 (control room group) to hand-held portable radios (OPS channel 1 or 2) via the plant radio repeater system. An additional radio console is located in panel H11-P703 to allow for backup communications to hand-held protableportable radios on various other user groups via plant radio zone 1 repeater system or backup repeaters (zone 2). Maintenance channels 1, 2, or 3 can also be selected at this station. This console also provodesprovides a backup radio communication selection into security zone 3 that provides another two repeaters for radio operation. to NRC-15-0061 Page 130 Fermi 2 Emergency Action Level Technical Bases The availability of one method of ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible This EAL is the cold condition equivalent of the hot condition EAL SU7.1. Generic This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Monroe and Wayne County EOCs. The third EAL condition addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. Fermi Basis Reference(s):

1. Fermi Emergency Plan Section F Emergency Communications

2. EP-580 Equipment Important to Emergency Response

3. NEI 99-01 CU5 to NRC-15-0061 Page 131 Fermi 2 Emergency Action Level Technical Bases Category: C - Cold Shutdown / Refueling System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

EAL: CA6.1 Alert The occurrence of any Table C-5 hazardous event AND EITHER of the following: Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required for the current operating mode Table C-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability: 4 - Cold Shutdown, 5 - Refueling Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, to NRC-15-0061 Page 132 Fermi 2 Emergency Action Level Technical Bases grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Basis: Plant-Specific The term required as used in this EAL is defined as the number of operable systems required by Technical Specifications for the present operating mode. Therefore, damage to systems that does not affect the required number of systems required to meet Technical Specifications for the current mode would not require classification. The significance of seismic events are discussed under EAL HU2.1 (Ref. 1). to NRC-15-0061 Page 133 Fermi 2 Emergency Action Level Technical Bases Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (Ref. 2, 3). Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 90 mph (sustained). (Ref. 4). Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (Ref. 5). An explosion that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first conditional addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second conditional addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS1 or RS1. Fermi Basis Reference(s):

1. AOP 20.000.01 Acts of Nature

2. AOP 20.000.03 Turbine Building Flooding

3. PLG-0849 Fermi 2 Internal Flooding Analysis to NRC-15-0061 Page 134 Fermi 2 Emergency Action Level Technical Bases

4. UFSAR Section 3.3.3.1 Design Wind Speed

5. AOP 20.000.22 Plant Fires

6. NEI 99-01 CA6 to NRC-15-0061 Page 135 Fermi 2 Emergency Action Level Technical Bases Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ALL (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety. The events of this category pertain to the following subcategories:

1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or Technological Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification.

4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown

5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation Events that are indicative of loss of Control Room habitability. If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

to NRC-15-0061 Page 136 Fermi 2 Emergency Action Level Technical Bases

7. ED Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

to NRC-15-0061 Page 137 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Security Initiating Condition: Confirmed SECURITY CONDITION or threat. EAL: HU1.1 Unusual Event A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by Security Shift Supervisor OR Notification of a credible security threat directed at the site OR A validated notification from the NRC providing information of an aircraft threat Mode Applicability: All Definition(s): SECURITY CONDITION - Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). Basis: Plant-Specific If the Security Shift Supervisor determines that a threat notification is credible, the Security Shift Supervisor will notify the Shift Manager that a Credible Threat condition exists for to NRC-15-0061 Page 138 Fermi 2 Emergency Action Level Technical Bases Fermi 2. The three main criteria for determining credibility are: technical feasibility, operational feasibility, and resolve. For Fermi 2, a validated notification delivered by the FBI, NRC or similar agency is treated as credible (Ref. 1, 2). Generic This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1. Timely and accurate communications between the Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan. The first threshold references the Security Shift Supervisor because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Fermi Safeguards Contingency Plan (Ref. 1). The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the Fermi Safeguards Contingency Plan (Ref. 1). Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may to NRC-15-0061 Page 139 Fermi 2 Emergency Action Level Technical Bases be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Fermi Safeguards Contingency Plan (Ref. 1). Escalation of the emergency classification level would be via IC HA1. Fermi Basis Reference(s):

1. Fermi Safeguards Contingency Plan

2. NEI 99-01 HU1 to NRC-15-0061 Page 140 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

EAL: HA1.1 Alert A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor OR A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). OWNER CONTROLLED AREA - Area depicted in the Fermi 2 Radiological Emergency Response Preparedness Plan Figure J-1 Owner-Controlled Area (Ref. 1) The company property immediately surrounding the PROTECTED AREA security fence. Access is normally limited to people on official business. Basis: Plant-Specific The Owner Controlled Area is depicted in the Fermi 2 Radiological Emergency Response Preparedness Plan Figure J-1 Owner-Controlled Area (Ref. 1). Generic to NRC-15-0061 Page 141 Fermi 2 Emergency Action Level Technical Bases This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between the Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security procedures. to NRC-15-0061 Page 142 Fermi 2 Emergency Action Level Technical Bases The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Fermi Safeguards Contingency Plan (Ref. 2). Escalation of the emergency classification level would be via IC HS1. Fermi Basis Reference(s):

1. Fermi 2 Radiological Emergency Response Preparedness Plan Figure J-1 Owner-Controlled Area

2. Fermi Safeguards Contingency Plan

3. NEI 99-01 HA1 to NRC-15-0061 Page 143 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION within the PROTECTED AREA EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). Basis: Plant-Specific None Generic This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. to NRC-15-0061 Page 144 Fermi 2 Emergency Action Level Technical Bases Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Fermi Safeguards Contingency Plan (Ref. 2). Escalation of the emergency classification level would be via IC HG1. Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. Fermi Safeguards Contingency Plan

3. NEI 99-01 HS1 to NRC-15-0061 Page 145 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 1 - Security Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility EAL:

HG1.1 General Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Supervisor AND EITHER of the following has occurred: Any of the following safety functions cannot be controlled or maintained Reactivity RPV water level RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). to NRC-15-0061 Page 146 Fermi 2 Emergency Action Level Technical Bases IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Plant-Specific None Generic This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions. It also addresses a HOSTILE ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps, heat exchangers, controls, etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between Security Shift Supervisor and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Fermi Safeguards Contingency Plan (Ref. 2). Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. Fermi Safeguards Contingency Plan

3. NEI 99-01 HG1 to NRC-15-0061 Page 147 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event greater than OBE levels EAL:

HU2.1 Unusual Event Seismic event greater than Operating Basis Earthquake (OBE) as indicated by peak accelerations > 0.05g vertical or > 0.08g horizontal on D30-R800 Active Seismic Playback Printer Mode Applicability: All Definition(s): None Basis: Plant-Specific The instrumentation used to indicate a seismic event includes the Triaxial Seismic Switch and the Triaxial Response Spectrum Recorder. Annunciator, ARP 6D69 (SEISMIC SYSTEM EVENT/TROUBLE), is sounded in the Control Room whenever the Triaxial Seismic Switch senses ground acceleration in excess of 0.01g (Ref. 1, 2, 3, 4). The Fermi 2 seismic instrumentation supports readily assessable (within 15 minutes) OBE indications (> 0.08g horizintalhorizontal, > 0.05g vertical acceleration). An offsite agency (USGS, National Earthquake Information Center) can confirm that an earthquake has occurred in the area of the plant. However, such confirmation should not preclude timely emergency declaration. Provide the analyst with the following Fermi 2 coordinates: 41º 57' 48" north latitude, 83º 15' 31" west longitude (Ref. 5). Generic This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant to NRC-15-0061 Page 148 Fermi 2 Emergency Action Level Technical Bases impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. Fermi Basis Reference(s):

1. UFSAR Section 3.7.4 Seismic Instrumentation Program

2. AOP 20.000.01 Acts of Nature

3. ARP 6D69 Seismic System Event/Trouble

4. SOP 23.612 Seismic Monitoring

5. UFSAR Section 1.2.2.1 Location and Size of Site

6. NEI 99-01 HU2 to NRC-15-0061 Page 149 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event A tornado strike within the PROTECTED AREA Mode Applicability: All Definition(s): PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). Basis: Plant-Specific None Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. AOP 20.000.01 Acts of Nature

3. NEI 99-01 HU3 to NRC-15-0061 Page 150 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component required for the current operating mode Mode Applicability: All Definition(s): FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Plant-Specific The term required as used in this EAL is defined as the number of operable systems required by Technical Specifications for the present operating mode. Therefore, isolation of components that do not affect the required number of systems required to meet Technical Specifications for the current mode would not require classification. to NRC-15-0061 Page 151 Fermi 2 Emergency Action Level Technical Bases Refer to Fermi 2 "Internal Flooding Analysis" to identify susceptible internal Flooding Areas (Ref. 2). Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. Fermi Basis Reference(s):

1. AOP 20.000.03 Turbine Building Flooding

2. PLG-0849 Fermi 2 Internal Flooding Analysis

2. NEI 99-01 HU3 to NRC-15-0061 Page 152 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event 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) Mode Applicability: All Definition(s): PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). Basis: Plant-Specific As used here, the term "offsite" is meant to be areas external to the Fermi 2 PROTECTED AREA. AOP 20.000.30, "OffisteOffsite Release of Toxic/Flammable Gas", provides additional information on hazardous substances and spills. Potential sources of toxic gases are chlorine and anahydrous ammonia (Ref. 2). Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. to NRC-15-0061 Page 153 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. AOP 20.000.30 OffisteOffsite Release of Toxic/Flammable Gas

3. NEI 99-01 HU3 to NRC-15-0061 Page 154 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technological Hazard Initiating Condition: Hazardous event EAL:

HU3.4 Unusual Event A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7) Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents. Mode Applicability: All Definition(s): None Basis: Plant-Specific None Generic This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site FLOODING caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. to NRC-15-0061 Page 155 Fermi 2 Emergency Action Level Technical Bases Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C. Fermi Basis Reference(s):

1. NEI 99-01 HU3 to NRC-15-0061 Page 156 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1): Report from the field (i.e., visual observation) Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND The FIRE is located within any Table H-1 area Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table H-1 Fire Areas Reactor Building Auxiliary Building Reactor Building Steam Tunnel RHR Complex 4160V Ductbanks between RHR Complex and Auxiliary Building Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. to NRC-15-0061 Page 157 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific The 15 minute requirement begins with a credible notification that a fire is occurring, or receipt of multiple valid fire detection system alarms or field validation of a single fire alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 15 minute time limit or a classification must be made Table H-1 Fire Areas are based on UFSAR Section 3.2 Classification of Structures, Components and Systems. Category I structures are those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (Ref. 1). Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc. Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report. Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." to NRC-15-0061 Page 158 Fermi 2 Emergency Action Level Technical Bases When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 15-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. Fermi Basis Reference(s):

1. UFSAR Section 3.2 Classification of Structures, Components and Systems

2. NEI 99-01 HU4 to NRC-15-0061 Page 159 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE) AND The fire alarm is indicating a FIRE within any Table H-1 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table H-1 Fire Areas Reactor Building Auxiliary Building Reactor Building Steam Tunnel RHR Complex 4160V Ductbanks between RHR Complex and Auxiliary Building Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. to NRC-15-0061 Page 160 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific The 30 minute requirement begins upon receipt of a single valid fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1. Table H-1 Fire Areas are based on UFSAR Section 3.2 Classification of Structures, Components and Systems. Category I structures are those structures containing functions and systems required for safe shutdown of the plant (SAFETY SYSTEMS) (Ref. 1). Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. Basis-Related Requirements from Appendix R to NRC-15-0061 Page 161 Fermi 2 Emergency Action Level Technical Bases Appendix R to 10 CFR 50, states in part: Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions." When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off. Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents. In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. Fermi Basis Reference(s):

1. UFSAR Section 3.2 Classification of Structures, Components and Systems

2. NEI 99-01 HU4 to NRC-15-0061 Page 162 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event A FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). Basis: Plant-Specific None Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. to NRC-15-0061 Page 163 Fermi 2 Emergency Action Level Technical Bases Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. NEI 99-01 HU4 to NRC-15-0061 Page 164 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4 - Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.4 Unusual Event A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Mode Applicability: All Definition(s): FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. PROTECTED AREA - An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 1). Basis: Plant-Specific None Generic This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is to NRC-15-0061 Page 165 Fermi 2 Emergency Action Level Technical Bases not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8. Fermi Basis Reference(s):

1. UFSAR Figure 1.2-5 Site Plot Plan

2. NEI 99-01 HU4 to NRC-15-0061 Page 166 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms or areas AND Entry into the room or area is prohibited or impeded (Note 5) Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. Table H-2 Safe Shutdown Rooms/Areas Room/Area Mode Applicability Control Room All Relay Room All AB3-DC MCC Area Mode 3 RB1-F17 Mode 3 RB1-F11 Mode 3 Mode Applicability: All Definition(s): None Basis: Plant-Specific The rooms/areas and associated mode applicability specified in Table H-2 are those that contain equipment which require a manual/local action as specified in operating procedures used for normal operation, cooldown and shutdown. This table excludes rooms/areas that may have procedurally directed actions that are of an administrative nature (normal rounds or routine inspections) or that are not crucial to the conduct of safe operation, cooldown and shutdown. to NRC-15-0061 Page 167 Fermi 2 Emergency Action Level Technical Bases Specifically: Control Room & Relay Room in all modes (Control Room is not included as it is addressed in RA3.1) AB3-DC MCC Area - Access is required when in Mode 3 to install power fuses and close the MCC for E1150-F008 which must be performed to align shutdown cooling suction path. RB1-F17 - Access is required when in Mode 3 to place the permissive switch in OPERATE for E11F610A if Div 1 RHR is being placed in shutdown cooling. This step must be performed to warmup the shutdown cooling piping. RB1-F11 - Access is required when in Mode 3 to place the permissive switch in OPERATE for E11F610B if Div 2 RHR is being placed in shutdown cooling. This step must be performed to warmup the shutdown cooling piping. (Ref. 1, 2, 3). Generic This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Directors judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary to NRC-15-0061 Page 168 Fermi 2 Emergency Action Level Technical Bases measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply. The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3. The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing). The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections). The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action. If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs. to NRC-15-0061 Page 169 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. Operating Procedures (normal plant operations, cooldown or shutdown), manual / local actions:

a. 22.000.03 - Power Operation 25% to 100% to 25%

b. 22.000.04 - Plant Shutdown From 25% Power

c. 22.000.05 - Pressure/Temp Monitoring During Heatup and Cooldown

d. 23.202 - High Pressure Coolant Injection System

e. 23.205 - Residual Heat Removal System

f. 23.206 - Reactor Core Isolation Cooling System

g. 23.427 - Primary Containment Isolation System

h. 23.610 - Reactor Protection System (RPS)

i. MGA03 - Procedure Use and Adherence

2. GOP 22.000.03 Power Operation 25% to 100% to 25%

3. GOP 22.000.04 Plant Shutdown from 25% Power

4. NEI 99-01 HA5 to NRC-15-0061 Page 170 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Dedicated or Remote Shutdown Panels Mode Applicability: All Definition(s): None Basis: Plant-Specific For the purpose of this EAL the 15 minute classification clock starts when the last licensed operator leaves the Control Room. Per AOP 20.000.18, "Control of the Plant from the Dedicated Shutdown Panel", (Ref. 1) and AOP 20.000.19, "Shutdown from Outside the Control Room", (Ref. 2) plant control is established at the Dedicated or Remote Shutdown Panels when: Initial Control Room actions are complete All available Operators have reported to the Dedication or Remote Shutdown Panels RPV level and pressure are being controlled from the Dedicated or Remote Shutdown Panels Generic This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control to NRC-15-0061 Page 171 Fermi 2 Emergency Action Level Technical Bases the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the emergency classification level would be via IC HS6. Fermi Basis Reference(s):

1. AOP 20.000.18 Control of the Plant from the Dedicated Shutdown Panel

2. AOP 20.000.19 Shutdown from Outside the Control Room

3. UFSAR Section 7.5.1.5.5 Procedure for Reactor Shutdown from Outside the Main Control Room

4. NEI 99-01 HA6 to NRC-15-0061 Page 172 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Dedicated or Remote Shutdown Panels AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1): Reactivity (Mode 1 and 2 only) RPV water level RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: All1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling Definition(s): None Basis: Plant-Specific For the purpose of this EAL the 15 minute clock starts when the last licensed operator leaves the Control Room. Per AOP 20.000.18, "Control of the Plant from the Dedicated Shutdown Panel", (Ref. 1) and AOP 20.000.19, "Shutdown from Outside the Control Room" (Ref. 2) plant control is established at the Dedicated or Remote Shutdown Panels when: Initial Control Room actions are complete All available Operators have reported to the Dedicated or Remote Shutdown Panels to NRC-15-0061 Page 173 Fermi 2 Emergency Action Level Technical Bases RPV level and pressure are being controlled from the Dedicated or Remote Shutdown Panels Generic This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not control is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Escalation of the emergency classification level would be via IC FG1 or CG1 Fermi Basis Reference(s):

1. AOP 20.000.18 Control of the Plant from the Dedicated Shutdown Panel

2. AOP 20.000.19 Shutdown from Outside the Control Room

3. UFSAR Section 7.5.1.5.5 Procedure for Reactor Shutdown from Outside the Main Control Room

4. NEI 99-01 HS6 to NRC-15-0061 Page 174 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs. Mode Applicability: All Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Plant-Specific None to NRC-15-0061 Page 175 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Unusual Event. Fermi Basis Reference(s):

1. NEI 99-01 HU7 to NRC-15-0061 Page 176 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Judgment Initiating Condition: Other conditions exist that in the judgment of the Emergency Director warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels. Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). Basis: Plant-Specific None Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert. to NRC-15-0061 Page 177 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. NEI 99-01 HA7 to NRC-15-0061 Page 178 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Judgment Initiating Condition: Other conditions existing that in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY. Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area) SITE BOUNDARY - That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. Basis: Plant-Specific None to NRC-15-0061 Page 179 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency. Fermi Reference(s):

1. NEI 99-01 HS7 to NRC-15-0061 Page 180 Fermi 2 Emergency Action Level Technical Bases Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Mode Applicability: All Definition(s): HOSTILE ACTION - An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). IMMINENT - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Basis: Plant-Specific None to NRC-15-0061 Page 181 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency. Fermi Basis Reference(s):

1. NEI 99-01 HG7 to NRC-15-0061 Page 182 Fermi 2 Emergency Action Level Technical Bases Category S - System Malfunction EAL Group: Hot Conditions (RCS temperature > 200ºF);

EALs in this category are applicable only in one or more hot operating modes. Numerous system-related equipment failure events that warrant emergency classification have been identified in this category. They may pose actual or potential threats to plant safety. The events of this category pertain to the following subcategories:

1. Loss of Essential AC Power Loss of essential plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4160 VAC essential buses 64B/64C and 65E/65F.

2. Loss of Vital DC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the 130 VDC ESF buses.

3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification. Losses of indicators are in this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5% clad failures) is indicative of fuel failures and is covered under Category F, Fission Product Barrier Degradation. However, lesser to NRC-15-0061 Page 183 Fermi 2 Emergency Action Level Technical Bases amounts of clad damage may result in coolant activity exceeding Technical Specification limits. These fission products will be circulated with the reactor coolant and can be detected by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Primary Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, ATWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Primary Containment integrity.

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

8. Hazardous Event Affecting Safety Systems Various natural and technological events that result in degraded plant safety system performance or significant visible damage warrant emergency classification under this subcategory.

to NRC-15-0061 Page 184 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite AC power capability to essential buses for 15 minutes or longer. EAL: SU1.1 Unusual Event Loss of all offsite AC power capability (Table S-1) to 4160V essential Division I (64B/64C) and Division II (65E/65F) for 15 min. (Note 1, 10) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table S-1 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Table S-1 lists AC sources capable of powering essential buses 64B/64C (Division 1) and 65E/65F (Division 2). For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it. to NRC-15-0061 Page 185 Fermi 2 Emergency Action Level Technical Bases This EAL is indicated by the loss of capability of all offsite AC power sources to power 4160V essential Division I (64B/64C) and Division II (65E/65F) for greater then or equal to 15 minutes. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) as an onsite AC power supply only if it is already aligned to and capable of powering one of the essential 4160 V divisions within the 15 minute time criteria (Ref. 2). The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. Generic This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC essential buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, capability means that an offsite AC power source(s) is available to the essential buses, whether or not the buses are powered from it. to NRC-15-0061 Page 186 Fermi 2 Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. Escalation of the emergency classification level would be via IC SA1. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. NEI 99-01 SU1 to NRC-15-0061 Page 187 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all but one AC power source to essential buses for 15 minutes or longer.

EAL: SA1.1 Alert AC power capability to 4160V essential Division I (64B/64C) and Division II (65E/65F) reduced to a single power source (Table S-1) for 15 min. (Note 1, 10) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table S-1 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; to NRC-15-0061 Page 188 Fermi 2 Emergency Action Level Technical Bases (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Basis: Plant-Specific Table S-1 lists AC sources capable of powering essential buses. For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it. This EAL is indicated by the loss of all but one AC power source to 4160V essential buses 64B/64C (Division I) and 65E/65F (Division II) for greater then or equal to 15 minutes such that a loss of any additional source will result in a complete loss of AC power to essential busses. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 (alternatively CTGs 11-2, 11-3 or to NRC-15-0061 Page 189 Fermi 2 Emergency Action Level Technical Bases 11-4) as an onsite AC power supply only if it is already aligned to and capable of powering one of the essential 4160 V divisions within the 15 minute time criteria (Ref. 2). The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. If the capability of a second source of essential bus power is not restored within 15 minutes, an Alert is declared under this EAL. Generic This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An AC power source is a source recognized in AOPs, and capable of supplying required power to an essential bus. Some examples of this condition are presented below. A loss of all offsite power with a concurrent failure of one essential power source (e.g., an onsite diesel generators). A loss of essential power sources (e.g., onsite diesel generators) with a single division of essential buses being back-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. NEI 99-01 SA1 to NRC-15-0061 Page 190 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Loss of all offsite and all onsite AC power to essential buses for 15 minutes or longer.

EAL: SS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability (Table S-1) to 4160V essential Division I (64B/64C) and Division II (65E/65F) for 15 min. (Note 1, 10) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table S-1 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Table S-1 lists AC sources capable of powering essential Division I and Division II AC buses. For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it. to NRC-15-0061 Page 191 Fermi 2 Emergency Action Level Technical Bases This EAL is indicated by the loss of all offsite and onsite AC power capability to 4160V essential Division I buses 64B/64C and Division II 65E/65F for greater then or equal to 15 minutes. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) as an onsite AC power supply only if it is already aligned to and capable of powering one of the essential 4160 V divisions within the 15 minute time criteria (Ref. 2). The 15-minute interval was selected as a threshold to exclude transient or momentary power losses. This EAL is the hot condition equivalent of the cold condition loss of all AC power EAL CA1.1. When in Cold Shutdown, Refueling, or Defueled mode, the event can be classified as an Alert because of the significantly reduced decay heat, lower temperature and pressure, increasing the time to restore one of the essential buses, relative to that existing when in hot conditions. to NRC-15-0061 Page 192 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. NEI 99-01 SS1 to NRC-15-0061 Page 193 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to essential buses OR loss of all essential AC and vital DC power sources for 15 minutes or longer.

EAL: SG1.1 General Emergency Loss of all offsite and all onsite AC power capability (Table S-1) to 4160V essential Division I (64B/64C) and Division II (65E/65F) (Note 10) AND EITHER of the following: Restoration of at least one essential division within 4 hours is not likely (Note 1) RPV water level CANNOT be restored and maintained > -25 in. Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table S-1 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific to NRC-15-0061 Page 194 Fermi 2 Emergency Action Level Technical Bases Table S-1 lists AC sources capable of powering essential AC buses. For emergency classification purposes, capability means that an AC power source is available to the essential divisional buses, whether or not the buses are currently powered from it (Ref. 1, 2). This EAL is indicated by the extended loss of offsite and onsite AC power capability to 4160V essential Division I buses 64B/64C and Division II buses 65E/65F either for greater then the Fermi 2 Station Blackout (SBO) coping analysis time (4 hrs.) (Ref. 3) or that has resulted in indications of an actual loss of adequate core cooling. Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (Ref. 4). The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 (alternatively CTGs 11-2, 11-3 or 11-4) provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division (Division 1 unless cross-tied) within 1 hour (Ref. 2). Generic to NRC-15-0061 Page 195 Fermi 2 Emergency Action Level Technical Bases This IC addresses a prolonged loss of all power sources to AC essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one 4160V essential bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers. The estimate for restoring at least one 4160V essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public. The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. UFSAR Section 8.4.2.1 SBO Coping Duration

4. EOP Support Documentation Section 1 Plant Specific Technical GuidlineGuideline (PSTG)

5. NEI 99-01 SG1 to NRC-15-0061 Page 196 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 1 - Loss of Essential AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to essential buses OR loss of all essential AC and vital DC power sources for 15 minutes or longer.

EAL: SG1.2 General Emergency Loss of all offsite and all onsite AC power capability (Table S-1) to 4160V essential Division I (64B/64C) and Division II (65E/65F) for 15 min. (Note 1, 10) AND Degraded voltage (< 105 VDC) on both 130 VDC system vital buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min. Table S-1 AC Power Sources Offsite: System Service Transformer 64 (Div I) System Service Transformer 65 (Div II) Onsite: EDGs 11/12 (Div I) EDGs 13/14 (Div II) CTG 11-1, 11-2, 11-3 or 11-4 Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Table S-1 lists AC sources capable of powering essential AC divisions. For emergency classification purposes, capability means that an AC power source is available to the to NRC-15-0061 Page 197 Fermi 2 Emergency Action Level Technical Bases essential divisional buses, whether or not the buses are currently powered from it (Ref. 1, 2). This EAL is indicated by the loss of all offsite and onsite essential AC power capability to 4160V essential Division I (64B/64C) and Division II (65E/65F) for greater than 15 minutes in combination with degraded vital DC power voltage. This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi. The safety-related essential electrical AC system consists of two physically and electrically independent and redundant power trains, Division I and Division II, supplying electrical power to safety-related equipment. Either Division I or Division II has the capability and the capacity to supply the essential AC power loads in the respective division. Each division is equipped with a split bus, radially fed network including four 4160-volt buses, four 480-volt buses and associated transformers. For the purpose of emergency classification, the loss of ability to power either split bus in a division (64B or C OR 65E or F) is considered an inability to power the respective division. Similarly, a loss of either EDG in a division is considered a loss of ability to power that divisions essential buses. Division I is normally supplied with offsite power from the 120-kV switchyard, whereas Division II is normally supplied from the 345-kV switchyard, thus providing two physically and electrically independent offsite sources (Ref. 1). CTG 11-1 provides a 120 kV AC line to a 13.8 kV/4160 V transformer to the essential buses. This Alternate AC (AAC) Power Supply has the capacity to supply only one fully loaded essential division within 1 hour (Ref. 2). Credit can be taken for CTG 11-1 as an onsite AC power supply only if it is already aligned to one of the essential 4160 V divisions. At Fermi 2, the vital 260/130 VDC System ensures power is available for the reactor to be shutdown safely and maintained in a safe condition. The vital DC system is divided into two independent divisions - Division I and Division II - with separate DC power supplies. These power supplies consist of two separate 260/130V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI. The system provides sufficient capacity, via each of the Class 1E DC batteries, to power all required loads for 4 hours following a loss of AC power supply (Ref. 3). to NRC-15-0061 Page 198 Fermi 2 Emergency Action Level Technical Bases Based on Technical Specifications Bases Section B.3.8.4, the 130 VDC battery minimum design voltage limit is 105 VDC. (Ref. 4). This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU1.2. Generic This IC addresses a concurrent and prolonged loss of both essential AC and Vital DC power. A loss of all essential AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both essential AC and vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. Fermi Basis Reference(s):

1. Design Basis Document RXX-00 ESS Electrical AC Systems

2. UFSAR Chapter 8 Electrical Power

3. Design Bases Document R32-00 DC Electrical System

4. Technical Specifications Bases Section B.3.8.4 DC Sources - Operating

5. NEI 99-01 SG8 to NRC-15-0061 Page 199 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer.

EAL: SS2.1 Site Area Emergency Degraded voltage (< 105 VDC) on both 130 VDC system vital buses for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific At Fermi 2, the vital 260/130 VDC System ensures power is available for the reactor to be shutdown safely and maintained in a safe condition. The vital DCsystemDC system is divided into two independent divisions - Division I and Division II - with separate DC power supplies. These power supplies consist of two separate 260/130V batteries and chargers serving systems such as RCIC, RHR, EDGs, and HPCI. The system provides sufficient capacity, via each of the Class 1E DC batteries, to power all required loads for 4 hours following a loss of AC power supply (Ref. 1). Based on Technical Specifications Bases Section B.3.8.4, the 130 VDC battery minimum design voltage limit is 105 VDC. (Ref. 2). Generic This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. to NRC-15-0061 Page 200 Fermi 2 Emergency Action Level Technical Bases Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via ICs RG1, FG1 or SG1. Fermi Basis Reference(s):

1. Design Bases Document R32-00 DC Electrical System

2. Technical Specifications Bases Section B.3.8.4 DC Sources - Operating

3. NEI 99-01 SS8 to NRC-15-0061 Page 201 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.

EAL: SU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table S-2 Safety System Parameters Reactor power RPV water level RPV pressure Primary containment pressure Torus water level Torus temperature Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as comuptercomputer based information systems. to NRC-15-0061 Page 202 Fermi 2 Emergency Action Level Technical Bases The Primary Containment Monitoring System is an informational system that provides indications of the Primary Containment environmental variables such as Primary Containment Pressure and Suppression Pool level and temperature (Ref. 4) The Integrated Plant Computer System (IPCS) is a computer system that provides the capability of monitoring, recording and displaying plant parameters via strategically located display devices. The IPCS is designed to be highly reliable and provide current information for selected plant variables. All realtime data displays update the current field conditions in a timely manner (Ref.1). SPDS is a function of the IPCS that provides a specific selection of emergency response information. SPDS uses data from selected plant data systems and processes the data for display on the IPCS. SPDS information can be displayed on any IPCS terminal, which includes those specifically located in the control room. The SPDS display in the control room is provided to assist the operators in assessing the safety status of the plant following an accident (Ref. 1). Generic This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency to NRC-15-0061 Page 203 Fermi 2 Emergency Action Level Technical Bases operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC SA3. Fermi Basis Reference(s):

1. UFSAR Section 7.6.1.9 Plant Computer Systems

2. AOP 20.615 Loss of Integrated Plant Computer System (IPCS)

3. ARP 3D17 IPCS Computer Trouble

4. SOP 23.408 Primary Containment Monitoring

5. NEI 99-01 SU2 to NRC-15-0061 Page 204 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

EAL: SA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table S-2 parameters from within the Control Room for 15 min. (Note 1) AND Any Table S-3 transient event in progress Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Table S-2 Safety System Parameters Reactor power RPV water level RPV pressure Primary containment pressure Torus water level Torus temperature Table S-3 Transient Events Automatic or manual runback > 25% thermal reactor power Electrical load rejection > 25% full electrical load Reactor scram ECCS actuation Thermal power oscillations > 10% Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown to NRC-15-0061 Page 205 Fermi 2 Emergency Action Level Technical Bases Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific SAFETY SYSTEM parameters listed in Table S-2 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Primary Containment Monitoring System is an informational system that provides indications of the Primary Containment environmental variables such as Primary Containment Pressure and Suppression Pool level and temperature (Ref. 4) The Integrated Plant Computer System (IPCS) is a computer system that provides the capability of monitoring, recording and displaying plant parameters via strategically located display devices. The IPCS is designed to be highly reliable and provide current information for selected plant variables. All realtime data displays update the current field conditions in a timely manner (Ref.1). SPDS is a function of the IPCS that provides a specific selection of emergency response information. SPDS uses data from selected plant data systems and processes the data for display on the IPCS. SPDS information can be displayed on any IPCS terminal, which includes those specifically located in the control room. The SPDS display in the control room is provided to assist the operators in assessing the safety status of the plant following an accident (Ref. 1). Generic This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission to NRC-15-0061 Page 206 Fermi 2 Emergency Action Level Technical Bases product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an inability to monitor means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via ICs FS1 or IC RS1. Fermi Basis Reference(s):

1. UFSAR Section 7.6.1.9 Plant Computer Systems

2. AOP 20.615 Loss of Integrated Plant Computer System (IPCS)

3. ARP 3D17 IPCS Computer Trouble to NRC-15-0061 Page 207 Fermi 2 Emergency Action Level Technical Bases

4. SOP 23.408 Primary Containment Monitoring

5. NEI 99-01 SA2 to NRC-15-0061 Page 208 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.

EAL: SU4.1 Unusual Event Offgas radiation monitor D11-K601A or D11-K601B high-high alarm (ARP 3D12) (Note 11) Note 11: Consistent with Technical Specification 3.7.5, this EAL is applicable at all times while in Mode 1, Mode 2 or in Mode 3 with any main steam line not isolated and steam jet air ejector in operation. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Elevated off-gas radiation activity is indicative of potential fuel clad failures and represents a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. The Technical Specification allowable limit is 340 mCi/sec of noble gases measured at the discharge of the 2.2 minute delay piping (Ref.3). The high-high radiation alarm setpoint is set to alert operators that the Technical Specification release limit may be exceeded (Ref. 1, 2). The high-high radiation alarm setpoint has been conservatively selected because it is operationally significant and is readily recognizable by Control Room operating staff. ConsistantConsistent with Technical Specification 3.7.5, EAL SU4.1 is applicable at all times while in Mode 1 (Power Operation) and in Mode 2 (Startup) or Mode 3 (Hot Shutdown) with any main steam line not isolated and steam jet air ejector (SJAE) in operation. (Ref. 1) Generic to NRC-15-0061 Page 209 Fermi 2 Emergency Action Level Technical Bases This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs. Fermi Basis Reference(s):

1. ARP 3D12 DIV. 1/2 OFFGAS RADN MONITOR HIGH/HIGH

2. AOP 20.000.07 Fuel Cladding Failure

3. Technical Specifications Section 3.7.5 Main Condenser Offgas

4. NEI 99-01 SU3 to NRC-15-0061 Page 210 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.

EAL: SU4.2 Unusual Event Sample analysis indicates that a reactor coolant activity value is > an allowable limit specified in Technical Specifications (Note 12) Note 12: Consistent with Technical Specification 3.4.7, this EAL is applicable at all times while in Mode 1, Mode 2 or in Mode 3 with any main steam line not isolated. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific This EAL addresses RCS specific activity exceeding the limits of Technical Specifications Section 3.4.7, which are: (1) 0.2 Ci per gram DEI-131 for more than 48 hours, or (2) 4.0 Ci per gram DEI-131. ConsistantConsistent with Technical Specification 3.4.7, EAL SU4.2 is applicable at all times while in Mode 1 (Power Operation) and in Mode 2 (Startup) or Mode 3 (Hot Shutdown) with any main steam line not isolated. (Ref. 1). Generic This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs. Fermi Basis Reference(s): to NRC-15-0061 Page 211 Fermi 2 Emergency Action Level Technical Bases

1. Technical Specifications 3.4.7 RCS Specific Activity

2. NEI 99-01 SU3 to NRC-15-0061 Page 212 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer.

EAL: SU5.1 Unusual Event RCS unidentified or pressure boundary leakage > 10 gpm for 15 min. OR RCS identified leakage > 25 gpm for 15 min. OR Leakage from the RCS to a location outside Primary Containment > 25 gpm for 15 min. (Note 1) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Unidentified leakage and identified leakage are determined by performance of the RCS water inventory balance (IPCS CHGNET, LRATE). Pressure boundary leakage would first appear as unidentified leakage and can only be positively identified by inspection (Ref. 1). Technical Specifications defines RCS leakage as follows (Ref. 1, 2): Identified leakage:

1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or to NRC-15-0061 Page 213 Fermi 2 Emergency Action Level Technical Bases

2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE.

Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall. RCS leakage outside of the Primary Containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as RCS to the Reactor Building Closed Cooling Water (RBCCW system), or systems that directly see RCS pressure outside primary containment such as Reactor Water Cleanup (RWCU), reactor water sampling system and Residual Heat Removal (RHR) system (when in the shutdown cooling mode) (Ref. 3). A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. Generic This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to unidentified leakage", "pressure boundary leakage" or "identified leakage (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the Primary Containment, or a location outside of Primary Containment. to NRC-15-0061 Page 214 Fermi 2 Emergency Action Level Technical Bases The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of Recognition Category R or F. Fermi Basis Reference(s):

1. Technical Specifications Section 1.1 Definitions - Leakage

2. Technical Specifications Section 3.4.4 RCS Operational Leakage

3. UFSAR Chapter 5 Reactor Coolant SystermSystem and Connected Systems Section 5.1 Summary Description

4. NEI 99-01 SU4 to NRC-15-0061 Page 215 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:

SU6.1 Unusual Event An automatic scram did not shut down the reactor after any RPS setpoint is exceeded AND A subsequent automatic scram or manual scram action taken at COP H11-P603 is successful in shutting down the reactor as indicated by reactor power < 3% (Note 8, 9). Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note 9: For manual scram actions, the reactor scram pushbuttons, taking the Reactor Mode Switch to Shutdown or manual initiation of ARI on COP H11-P603 are the only methods applicable to this EAL. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): None Basis: Plant-Specific Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale trip setpoint of 3% (Ref. 1). to NRC-15-0061 Page 216 Fermi 2 Emergency Action Level Technical Bases For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 3% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. For purposes of emergency classification, a successful manual reactor scram includes only those actions taken by the reactor operator in the Control Room on the reactor control console (COP H11-P603). These actions include the manual scram pushbuttons, placing the Reactor Mode Switch in Shutdown and manual initiation of ARI. These pushbuttons and controls can be rapidly manipulated from the Control Room panels. If the above described response cannot be verified, operators perform contingency actions that manually insert control rods or implement alternate control rod insertion methodologies performed either away from the reactor control console or external to the Control Room. Those actions required to be performed away from the reactor control console or outside of the Control Room to initiate rapid control rod insertion are not considered a successful manual reactor scram. In the event that the operator identifies a reactor scram is imminent and successfully initiates a manual reactor scram before the automatic trip setpoint is reached, no declaration is required. The successful manual scram of the reactor before it reaches its automatic scram setpoint or reactor scram signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. If manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 3%) following a failure of an automatic scram, the event escalates to an Alert under EAL SA6.1. The APRM downscale trip setpoint (3%) is a minimum reading on the power range scale that indicates power production (Ref. 1). At or below the APRM downscale trip setpoint, to NRC-15-0061 Page 217 Fermi 2 Emergency Action Level Technical Bases plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters (e.g., number of open SRVs, number of open main turbine bypass valves, main steam flow, RPV pressure and torus water temperature trend, etc.) can be used to determine if reactor power is greater than or equal to 3% power. By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operation or Startup), however, requires emergency classification of at least an Unusual Event. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events. Generic This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core to NRC-15-0061 Page 218 Fermi 2 Emergency Action Level Technical Bases heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles. Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted. to NRC-15-0061 Page 219 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical GuidlineGuideline

2. AOP 20.000.21 Reactor Scram

3. NEI 99-01 SU5 to NRC-15-0061 Page 220 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:

SU6.2 Unusual Event A manual scram did not shut down the reactor after any manual scram action was initiated AND A subsequent automatic scram or manual scram action taken at COP H11-P603 is successful in shutting down the reactor as indicated by reactor power < 3% (Note 8, 9). Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note 9: For manual scram actions, the reactor scram pushbuttons, taking the Reactor Mode Switch to Shutdown or manual initiation of ARI on COP H11-P603 are the only methods applicable to this EAL. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): None Basis: Plant-Specific This EAL addresses a failure of a manually initiated scram in the absence of having exceeded an automatic RPS trip setpoint and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power < 3%). Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed to NRC-15-0061 Page 221 Fermi 2 Emergency Action Level Technical Bases by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale trip setpoint of 3% (Ref. 1). For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to or below 3% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. For purposes of emergency classification, a successful manual reactor scram includes only those actions taken by the reactor operator in the Control Room on the reactor control console (COP H11-P603). These actions include the manual scram pushbuttons, placing the Reactor Mode Switch in Shutdown and manual initiation of ARI. These pushbuttons and controls can be rapidly manipulated from the Control Room panels. Taking the Mode Switch out of Run position reduces the APRM scram setpoint to 15% reactor power. If reactor power is > 15% at the time the Mode Switch is taken out of the Run position, an automatic RPS scram signal will have been generated in addition to the manual scram signal generated by taking the Mode Switch to Shutdown. Should the other immediate manual scrams not be successful in reducing reactor power to < 3%, an Alert should be declared based on exceeding EAL SA6.1. If the above described response cannot be verified, operators perform contingency actions that manually insert control rods or implement alternate control rod insertion methodologies performed either away from the reactor control console or external to the Control Room. Those actions required to be performed away from the reactor control console or outside of the Control Room to initiate rapid control rod insertion are not considered a successful manual reactor scram. to NRC-15-0061 Page 222 Fermi 2 Emergency Action Level Technical Bases If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 3%) following a failure of an initial manual scram, the event escalates to an Alert under EAL SA6.1. The APRM downscale trip setpoint (3%) is a minimum reading on the power range scale that indicates power production (Ref. 1). At or below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters (e.g., number of open SRVs, number of open main turbine bypass valves, main steam flow, RPV pressure and torus water temperature trend, etc.) can be used to determine if reactor power is greater than or equal to 3% power. By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operation or Startup), however, requires emergency classification of at least an Unusual Event. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events. Generic This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems. to NRC-15-0061 Page 223 Fermi 2 Emergency Action Level Technical Bases If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram) using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plants decay heat removal systems. A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactoreactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control consoles. Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via IC FA1. Absent the plant conditions needed to meet either IC SA6 or FA1, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated. to NRC-15-0061 Page 224 Fermi 2 Emergency Action Level Technical Bases If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical GuidlineGuideline

2. AOP 20.000.21 Reactor Scram

3. NEI 99-01 SU5 to NRC-15-0061 Page 225 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.

EAL: SA6.1 Alert An automatic or manual scram fails to shut down the reactor AND Manual scram actions taken at COP H11-P603 are not successful in shutting down the reactor as indicated by reactor power > 3% (Note 8, 9) Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies. Note 9: For manual scram actions, the reactor scram pushbuttons, taking the Reactor Mode Switch to Shutdown or manual initiation of ARI on COP H11-P603 are the only methods applicable to this EAL. Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): None Basis: Plant-Specific This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (> 3%) (Ref. 1). Following a successful reactor scram, rapid insertion of the control rods occurs. Nuclear power promptly drops to a fraction of the original power level and then decays to a level several decades less with a negative period. The reactor power drop continues until reactor power reaches the point at which the influence of source neutrons on reactor to NRC-15-0061 Page 226 Fermi 2 Emergency Action Level Technical Bases power starts to be observable. A predictable post-scram response from an automatic reactor scram signal should therefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful scram has therefore occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power below the APRM downscale trip setpoint of 3% (Ref. 1). For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power below 3% is not considered a successful automatic scram. If automatic actuation of ARI has occurred and caused reactor shutdown, the automatic RPS scram must have failed. ARI is a backup means of inserting control rods in the unlikely event that an automatic RPS scram signal exists but the reactor continues to generate significant power. However, a successful automatic initiation of ARI is an acceptable means of establishing reactor shutdown conditions relative to the EAL threshold in the absence of any required subsequent manual scram actions. For purposes of emergency classification, a successful manual reactor scram includes only those actions taken by the reactor operator in the Control Room on the reactor control console (COP H11-P603). These actions include the manual scram pushbuttons, placing the Reactor Mode Switch in Shutdown and manual initiation of ARI. These pushbuttons and controls can be rapidly manipulated from the Control Room panels. If the above described response cannot be verified, operators perform contingency actions that manually insert control rods or implement alternate control rod insertion methodologies performed either away from the reactor control console or external to the Control Room. Those actions required to be performed away from the reactor control console or outside of the Control Room to initiate rapid control rod insertion are not considered a successful manual reactor scram. If subsequent manual reactor scram actions in the Control Room fail to reduce reactor power below the power associated with the safety system design (< 3%) following a failure of an initial automatic or manual scram, the event is classifiable under this EAL. The APRM downscale trip setpoint (3%) is a minimum reading on the power range scale that indicates power production (Ref. 1). At or below the APRM downscale trip setpoint, to NRC-15-0061 Page 227 Fermi 2 Emergency Action Level Technical Bases plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM/IRM) indications or other reactor parameters (e.g., number of open SRVs, number of open main turbine bypass valves, main steam flow, RPV pressure and torus water temperature trend, etc.) can be used to determine if reactor power is greater than 3% power. By definition, an operating mode change occurs when the Mode Switch is moved from the startup or run position to the shutdown position. The plant operating mode that existed at the time the event occurs (i.e., Power Operation or Startup), however, requires emergency classification of at least an Unusual Event. The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for ATWS events. Generic This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanelsback panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be at the reactor control console. Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. to NRC-15-0061 Page 228 Fermi 2 Emergency Action Level Technical Bases The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC SS6 or FS1, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical GuidlineGuideline

2. AOP 20.000.21 Reactor Scram

3. NEI 99-01 SA5 to NRC-15-0061 Page 229 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal.

EAL: SS6.1 Site Area Emergency An automatic or manual scram fails to shut down the reactor AND All actions to shut down the reactor are not successful as indicated by reactor power > 3% AND EITHER of the following conditions exist: RPV water level cannot be restored and maintained > -25 in. Torus water temperature and RPV pressure cannot be maintained below the Heat Capacity Limit (HCL) Mode Applicability: 1 - Power Operation, 2 - Startup Definition(s): None Basis: Plant-Specific This EAL addresses the following: Any automatic or manual reactor scram signal followed by a failure of all subsequent methods to shut down the reactor, both within and external to the Control Room, to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (> 3%, Ref. 1, 2), and Indications that either core cooling is extremely challenged (RPV water level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level) or heat removal is extremely challenged (torus water temperature to NRC-15-0061 Page 230 Fermi 2 Emergency Action Level Technical Bases and RPV pressure cannot be maintained below the Heat CapcityCapacity Limit) (Ref. 1). For this Site Area EmergncyEmergency EAL, reactor shutdown achieved by injection of boron or use of the alternate control rod insertion methods of 29.ESP.03 is also credited provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist (Ref. 3). The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers. Indication that core cooling is extremely challenged is manifested by RPV water level cannot be restored and maintained above -25 in. (Ref. 1,4). The Minimum Steam Cooling RPV Water Level (MSCRWL) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F. Consistent with the EOP definition of cannot be restored and maintained, the determination that RPV water level cannot be restored and maintained above the MSCRWL may be made at, before, or after RPV water level actually decreases to this point. Indication that core heat removal is extremely challenged is manifested by the inability to maintain torus water temperature and RPV pressure below the Heat Capacity Limit (HCL). The HCL is the highest torus water temperature from which emergency RPV depressurization will not raise (Ref. 1, 5): o Torus water temperature above the design value, or o Torus pressure above Primary Containment Pressure Limit before the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCL is a function of RPV pressure and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant. Plant parameters in excess of the HCL could be a precursor of Primary Containment failure. (Ref. 1, 5, 6) to NRC-15-0061 Page 231 Fermi 2 Emergency Action Level Technical Bases The HCL is given in EOP 29.100.01 Sheet 6 Curves, Cautions and Tables (Ref. 5). This threshold is met when Emergency RPV Depressurization is required in EOP Primary ContainmnetContainment Control, Step TWT-5 (Ref. 6). This condition addresses loss of functions required for hot shutdown with the reactor at pressure and temperature. Generic This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. Escalation of the emergency classification level would be via IC RG1 or FG1. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical GuidlineGuideline

2. AOP 20.000.21 Reactor Scram

3. 29.ESP.03 AlternatAlternate Control Rod Insertion Methods

4. EOP 29.100.01 Sheet 1A RPV Control - ATWS

5. EOP 29.100.01 Sheet 6 Curves, Cautions and Tables

6. EOP 29.100.01 Sheet 2 Primary Containment Control

7. NEI 99-01 SS5 to NRC-15-0061 Page 232 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities.

EAL: SU7.1 Unusual Event Loss of all Table S-4 onsite communication methods OR Loss of all Table S-4 offsite communication methods OR Loss of all Table S-4 NRC communication methods Table S-4 Communication Methods System Onsite Offsite NRC Administrative Telephones X X X RERP Emergency Telephones X X X Satellite Phones X X Federal Telephone System (ENS) X X Automatic Ring Lines X MI State Radios (800 MHz) X X Plant Radio System X Hi-Com (PA System) X Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None to NRC-15-0061 Page 233 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific The Table S-4 list for onsite communications loss encompasses the loss of all means of routine communications (e.g., administrative and internal telephones, plant page [Hi-Com] and plant radios) (Ref. 1, 2). The Table S-4 list for offsite communications loss encompasses the loss of all means of communications with offsite authorities. This includes the RERP telephone dedicated ring lines, backup phone systems administrative telephone lines, satellite, and FTS (ENS) which can be utilized as a regular telephone (Ref. 1, 2). The Table S-4 list for NRC communications loss encompasses the loss of all means of communications with the NRC. This includes the FTS (ENS), backup phone systems (administrative telephone lines, RERP phones and satellite) (Ref. 1, 2). The communications methods used at Fermi 2 are described in the RERP Plan (Ref. 1). The radio network at Fermi 2 involves several radio systems to effect communicaitonscommunications within the plant with damage control teams, rescue teams, fire brigade, radiological monitoring teams, and security personnel as well as provide backup comminicationscommunications to essential Offsite Emergency Response Organizations (OROs) in the event of telephone equipment malfunction. There are two radio consoles normally used in the Control Room. One is installed in panel H11-P700 to establish communications using plant radio zone 1 (control room group) to hand-held portable radios (OPS channel 1 or 2) via the plant radio repeater system. An additional radio console is located in panel H11-P703 to allow for backup communications to hand-held protableportable radios on various other user groups via plant radio zone 1 repeater system or backup repeaters (zone 2). Maintenance channels 1, 2, or 3 can also be selected at this station. This console also provodesprovides a backup radio communication selection into security zone 3 that provides another two repeaters for radio operation. to NRC-15-0061 Page 234 Fermi 2 Emergency Action Level Technical Bases The availability of one method of ordinary offsite communication is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible This EAL is the hot condition equivalent of the cold condition EAL CU5.1. Generic This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the State, Monroe and Wayne County EOCs The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. Fermi Basis Reference(s):

1. Fermi Emergency Plan Section F Emergency Communications

2. EP-580 Equipment Important to Emergency Response

4. NEI 99-01 SU6 to NRC-15-0061 Page 235 Fermi 2 Emergency Action Level Technical Bases Category: S - System Malfunction Subcategory: 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM required for the current operating mode.

EAL: SA8.1 Alert The occurrence of any Table S-5 hazardous event AND EITHER of the following: Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required for the current operating mode Table S-5 Hazardous Events Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics as determined by the Shift Manager Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): EXPLOSION - A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. to NRC-15-0061 Page 236 Fermi 2 Emergency Action Level Technical Bases FIRE - Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. FLOODING - A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM - A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. VISIBLE DAMAGE - Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Basis: Plant-Specific The term required as used in this EAL is defined as the number of operable systems required by Technical Specifications for the present operating mode. Therefore, damage to systems that does not affect the required number of systems required to meet Technical Specifications for the current mode would not require classification. The significance of seismic events are discussed under EAL HU2.1 (Ref. 1). to NRC-15-0061 Page 237 Fermi 2 Emergency Action Level Technical Bases Internal FLOODING may be caused by events such as component failures, equipment misalignment, or outage activity mishaps (Ref. 2, 3). Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 90 mph. (Ref. 4). Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Zone in the fire response procedure (Ref. 5). An explosion (including a steam line explosion) that degrades the performance of a SAFETY SYSTEM train or visibly damages a SAFETY SYSTEM component or structure would be classified under this EAL. The need to classify a steam line break not considered an explosion itself is considered in fission product barrier degradation monitoring (EAL Category F). Generic This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC FS1 or RS1. to NRC-15-0061 Page 238 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. AOP 20.000.01 Acts of Nature

2. AOP 20.000.03 Turbine Building Flooding

3. PLG-0849 Fermi 2 Internal Flooding Analysis

4. UFSAR Section 3.3.3.1 Design Wind Speed

5. AOP 20.000.22 Plant Fires

6. NEI 99-01 SA9 to NRC-15-0061 Page 239 Fermi 2 Emergency Action Level Technical Bases Category F - Fission Product Barrier Degradation EAL Group: Hot Conditions (RCS temperature > 200ºF);

EALs in this category are applicable only in one or more hot operating modes. EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. C. Primary Containment (PC): The Primary Containment Barrier includes the drywell, the suppression pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency. The EALs in this category require evaluation of the loss and potential loss thresholds listed in the fission product barrier matrix of Table F-1 (Attachment 2). Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier. The number of barriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level: Alert: Any loss or any potential loss of either Fuel Clad or RCS Site Area Emergency: Loss or potential loss of any two barriers to NRC-15-0061 Page 240 Fermi 2 Emergency Action Level Technical Bases General Emergency: Loss of any two barriers and loss or potential loss of the third barrier The logic used for Category F EALs reflects the following considerations: The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Primary Containment Barrier. Unusual Event ICs associated with fission product barriers are addressed in Recognition Category S. For accident conditions involving a radiological release, evaluation of the fission product barrier thresholds will need to be performed in conjunction with dose assessments to ensure correct and timely escalation of the emergency classification. For example, an evaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for General Emergency IC RG1 has been exceeded. The fission product barrier thresholds specified within a scheme reflect plant-specific Fermi 2 design and operating characteristics. As used in this category, the term RCS leakage encompasses not just those types defined in Technical Specifications but also includes the loss of RCS mass to any location inside the primary containment, an interfacing system, or outside of the primary containment. The release of liquid or steam mass from the RCS due to the as-designed/expected operation of a relief valve is not considered to be RCS leakage. At the Site Area Emergency level, EAL users should maintain cognizance of how far present conditions are from meeting a threshold that would require a General Emergency declaration. For example, if the Fuel Clad and RCS fission product barriers were both lost, then there should be frequent assessments of containment radioactive inventory and integrity. Alternatively, if both the Fuel Clad and RCS fission product barriers were potentially lost, the Emergency Director would have more assurance that there was no immediate need to escalate to a General Emergency. to NRC-15-0061 Page 241 Fermi 2 Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL: FA1.1 Alert Any loss or any potential loss of either Fuel Clad or RCS barrier (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Primary Containment barrier. Unlike the Primary Containment barrier, loss or potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Primary Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under EAL FS1.1. Generic None Fermi Basis Reference(s):

1. NEI 99-01 FA1 to NRC-15-0061 Page 242 Fermi 2 Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1.1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the Site Area Emergency classification level, each barrier is weighted equally. A Site Area Emergency is therefore appropriate for any combination of the following conditions: One barrier loss and a second barrier loss (i.e., loss - loss) One barrier loss and a second barrier potential loss (i.e., loss - potential loss) One barrier potential loss and a second barrier potential loss (i.e., potential loss - potential loss) At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Primary Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss to NRC-15-0061 Page 243 Fermi 2 Emergency Action Level Technical Bases thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent. Generic None Fermi Basis Reference(s):

1. NEI 99-01 FS1 to NRC-15-0061 Page 244 Fermi 2 Emergency Action Level Technical Bases Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of the third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of the third barrier (Table F-1) Mode Applicability: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown Definition(s): None Basis: Plant-Specific Fuel Clad, RCS and Primary Containment comprise the fission product barriers. Table F-1 (Attachment 2) lists the fission product barrier thresholds, bases and references. At the General Emergency classification level each barrier is weighted equally. A General Emergency is therefore appropriate for any combination of the following conditions: Loss of Fuel Clad, RCS and Primary Containment barriers Loss of Fuel Clad and RCS barriers with potential loss of Primary Containment barrier Loss of RCS and Primary Containment barriers with potential loss of Fuel Clad barrier Loss of Fuel Clad and Primary Containment barriers with potential loss of RCS barrier Generic None to NRC-15-0061 Page 245 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. NEI 99-01 FG1 to NRC-15-0061 Page 246 Fermi 2 Emergency Action Level Technical Bases Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ALL (EALs in this category are applicable to any plant condition, hot or cold.) An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety. A Notification of Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask confinement boundary is damaged or violated. to NRC-15-0061 Page 247 Fermi 2 Emergency Action Level Technical Bases Category: ISFSI Subcategory: Confinement Boundary Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL: EU1.1 Unusual Event Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than EITHER of the following on the surface of the spent fuel cask (overpack): 60 mrem/hr ( + n) on the top of the overpack OR 600 mrem/hr ( + n) on the side of the overpack excluding inlet and outlet ducts Mode Applicability: All Definition(s): CONFINEMENT BOUNDARY- The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As applied to the Fermi 2 ISFSI, the CONFINEMENT BOUNDARY is defined to be the HI-STORM MiltiMulti-Purpose Canister (MPC). Basis: Plant-Specific Overpacks are the casks which receive and contain the sealed Multi-Purpose Canisters (MPCs) for interim storage on the ISFSI. They provide gamma and neutron shielding, and provide for ventilated air flow to promote heat transfer from the MPC to the environs. The term overpack does not include the transfer cask (Ref. 1). MPCs are the sealed spent nuclear fuel canisters which consist of a honeycombed fuel basket contained in a cylindrical canister shell which is welded to a baseplate, lid with welded port cover plates, and closure ring. The MPC provides the CONFINEMENT BOUNDARY for the contained radioactive materials (Ref. 1). to NRC-15-0061 Page 248 Fermi 2 Emergency Action Level Technical Bases The values shown represent 2 times the limits specified in the ISFSI Certificate of Compliance Technical Specification section 5.7.4 for radiation external to a loaded MPC overpack (Ref. 1). Generic This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of damage is determined by radiological survey. The technical specification multiple of 2 times, which is also used in Recognition Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the on-contact dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSIs are covered under ICs HU1 and HA1. Fermi Basis Reference(s):

1. Certificate of Compliance No. 1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System to NRC-15-0061 Page 249 Fermi 2 Emergency Action Level Technical Bases ATTACHMENT 2 FISSION PRODUCT BARRIER MATRIX AND BASES to NRC-15-0061 Page 250 Fermi 2 Emergency Action Level Technical Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Primary Containment).

The table is structured so that the three barriers occupy adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential Loss thresholds. The first column of the table (to the left of the Fuel Clad Barrier Loss column) lists the categories (types) of fission product barrier thresholds. The fission product barrier categories are:

1. RPV Water Level

2. RCS Leak Rate

3. Primary Containment Conditions

4. Primary Containment Radiation / RCS Activity

5. Primary Containment Integrity or Bypass

6. Emergency Director Judgment Each category occupies a row in Table F-1 thus forming a matrix defined by the category rows and the Loss/Potential Loss columns. The intersection of each category row with each Loss/Potential Loss column forms a cell in which one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word None is entered in the cell.

Thresholds are assigned letters within each Loss and Potential Loss column beginning with A. In this manner, a threshold can be identified by its category number and threshold letter. For example, the first Fuel Clad barrier Loss in Category 2 is FC Loss 2.A, the third Primary Containment barrier Potential Loss in Category 4 is PC P-Loss 4.C, etc. If a cell in Table F-1 contains more than one threshold, each of the thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessary to exceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss. to NRC-15-0061 Page 251 Fermi 2 Emergency Action Level Technical Bases Subdivision of Table F-1 by category facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a systematic approach to assessing the classification status of the fission product barriers. When equipped with knowledge of plant conditions related to the fission product barriers, the EAL-user first scans down the category column of Table F-1, locates the likely category and then reads across the row of fission product barrier Loss and Potential Loss thresholds in that category to determine if any threshold has been exceeded. If a threshold has not been exceeded in that category row, the EAL-user proceeds to the next likely category and continues review of the row of thresholds in the new category. The EAL-user must examine each of the three fission product barriers to determine if other barrier thresholds in the category are lost or potentially lost. For example, if Primary Containment radiation is sufficiently high (i.e., > 10,000 R/hr), a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the Primary Containment barrier exist. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1, FA1.1 and FU1.1 to determine the appropriate emergency classification. In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Primary Containment barrier threshold bases. In each barrier, the bases are given according to category Loss followed by category Potential Loss beginning with Category 1, then 25.

Enclosure 2 to NRC-15-0061 Page 252 Fermi 2 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A. SAG entry is requiredInadequate core A. SAG entry is requiredInadequate cooling as indicated by ANY of core cooling as indicated by ANY the following: of the following:

1. RPV water level 1. RPV water level cannot cannot be restored be restored and and maintained > -48 maintained > -48 in. with in. with 5725 gpm 5725 gpm Core Spray Core Spray loop flow loop flow 1 OR A. RPV water level cannot be restored and maintained above 0 A. RPV water level cannot be restored and maintained above 0 None None OR RPV Water 2. RPV water level in. (TAF) or cannot be determined in. (TAF) or cannot be determined 2. RPV water level cannot Level cannot be restored be restored and and maintained > -25 maintained > -25 in with in with < 5725 gpm < 5725 gpm Core Spray Core Spray loop flow loop flow OR OR 1.3. RPV water level 1.3. RPV water level cannot cannot be be determined and core determined and core damage is occurring damage is occurring A. UNISOLABLE primary system A. UNISOLABLE primary system leakage into Secondary leakage into Secondary Containment that results in Containment that results in exceeding EITHER of the A. UNISOLABLE break in any of the exceeding EITHER of the following:

following: following:

1. One or more Secondary Main Steam Line 1. One or more Secondary Containment Control HPCI Steam Line Containment Control Max Max Safe Operating 2 None None

RCIC Steam Line RWCU Normal Operating Temperatures (EOP Temperatures (EOP Table 12) None RCS Leak Rate Feedwater Table 12) OR OR OR 1.2. One or more Secondary B. Emergency RPV Depressurization 1.2. One or more Secondary Containment Control is required Containment Control Max Max Safe Operating Normal Operating Area Area Radiation Levels Radiation Levels (EOP (EOP Table 14) area Table 14) radiation levels (Table F-2) A. Primary Containment Pressure

                                                                                                                                                                                                        > 62 psig A. UNPLANNED rapid drop in Primary Containment pressure          OR following Primary Containment 3                      None                                None A. Drywell pressure > 1.68 psig due None pressure rise                      B.  > 6% H2 AND > 5% O2 in EITHER the drywell or suppression PC                                                                                  to RCS leakage                                                           OR                                     chamber Conditions B. Primary Containment pressure          OR response not consistent with LOCA conditions                    C. EOP Heat Capacity Limit (HCL) exceeded to NRC-15-0061 Page 253                                      Fermi 2 Emergency Action Level Technical Bases Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier                                               Reactor Coolant System Barrier                                                  Primary Containment Barrier Category                    Loss                        Potential Loss                                    Loss                        Potential Loss                                  Loss                       Potential Loss A. CHRRM reading > 2.25E+3 R/hr 4          OR PC Rad /                                                         None                    A. CHRRM reading > 8.72E+1 R/hr                       None                                  None                 A. CHRRM reading > 1.79E+4 R/hr RCS      B. Primary coolant activity > 300 Activity      µCi/gm DEI-131 A. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal 5                        None                                 None                                    None                                  None                      OR                                                 None PC Integrity or Bypass                                                                                                                                                              B. Intentional Primary Containment venting, irrespective of offsite radioactivity release rates, per EOPs 6       A. Any condition in the opinion of the Emergency Director that A. Any condition in the opinion of the Emergency Director that A. Any condition in the opinion of the Emergency Director that A. Any condition in the opinion of the Emergency Director that indicates A. Any condition in the opinion of the Emergency Director that A. Any condition in the opinion of the Emergency Director that indicates ED          indicates loss of the fuel clad      indicates potential loss of the fuel                                                                                 indicates loss of the Primary       potential loss of the Primary indicates loss of the RCS barrier    potential loss of the RCS barrier Judgment        barrier                              clad barrier                                                                                                         Containment barrier                 Containment barrier to NRC-15-0061 Page 254                 Fermi 2 Emergency Action Level Technical Bases Barrier:                 Fuel Clad Category:                1. RPV Water Level Degradation Threat:      Loss Threshold:

A. SAG entry is requiredInadequate core cooling as indicated by ANY of the following:

1. RPV water level cannot be restored and maintained > -48 in. with 5725 gpm Core Spray loop flow OR

2. RPV water level cannot be restored and maintained > -25 in. with

      < 5725 gpm Core Spray loop flow OR 1.3. RPV water level cannot be determined and core damage is occurring Definition(s):

None Basis: Plant-Specific Requirements for Primary Containment Flooding correspond to entry into the Severe Accident Guidelines (SAGs) and are established in EOP RPV Control, EOP RPV Control - ATWS and EOP RPV Flooding & EOP RPV Flooding - ATWS (Ref. 1, 2, 3, 4, 5). These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Direction is specified that SAG entry is required represents the requirement for Primary Containment Flooding when: EOP RPV Control - RPV water level cannot be restored and maintained above

      -48 in. with required Core Spray flow.

EOP RPV Control - RPV water level cannot be restored and maintained above -25 in. (MSCRWL) with insufficient Core Spray flow. to NRC-15-0061 Page 255 Fermi 2 Emergency Action Level Technical Bases EOP RPV Control - ATWS - RPV water level cannot be restored and maintained above -25 in. (MSCRWL). EOP Flooding & EOP Flooding - ATWS - RPV water level cannot be determined and it is determined that core damage is occurring. This threshold is also a Potential Loss of the Containment barrier (PC P-Loss 1.A). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss 1.A). SAG entry (Primary Containment Flooding), therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. Generic The Loss threshold represents the EOP requirement for SAG entry.primary containment flooding. This is identified in the BWROG EPGs/SAGs when the phrase, SAG Entry Is Required," appears. Since a site-specific RPV water level is not specified here, the Loss threshold phrase, SAG entry is required, also accommodates the EOP need to flood the primary containment when RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring. Fermi Basis Reference(s):

1. EOP 29.100.01 Sheet 1 RPV Control

2. EOP 29.100.01A Sheet 1 RPV Control - ATWS

3. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress. & Steam Cooling

4. EOP 29.100.01 Sheet 3A RPV Flooding, Emerg. Depress. & Steam Cooling - ATWS

5. EOP Support Documentation Section 1 Plant Specific Technical Guideline

6. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A to NRC-15-0061 Page 256 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 1. RPV Water Level Degradation Threat: Potential Loss Threshold:

A. RPV water level cannot be restored and maintained above 0 in. (TAF) or cannot be determined Definition(s): None Basis: Plant-Specific An RPV water level instrument reading of 0 in. indicates RPV water level is at the top of active fuel. When RPV water level is at or above the top of active fuel, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV water level is below the top of active fuel following depressurization of the RPV (automatically, manually or by failure of the RCS barrier), the uncovered portion of the core must be cooled by less reliable means (i.e., spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling (Ref. 1). This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has to NRC-15-0061 Page 257 Fermi 2 Emergency Action Level Technical Bases been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.Consistentinventory. Consistent with the EOP definition of cannot be restored and maintained, the determination that RPV water level cannot be restored and maintained above the top of active fuel may be made at, before, or after RPV water level actually decreases to this point. (Ref. 1) When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP RPV Flooding and EOP RPV Flooding - ATWS specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events). (Ref. 2, 3). If RPV water level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists. Note that EOP RPV Control - ATWS may require intentional uncovery of the core and control of RPV water level between 0 in. and -25 in., the Minimum Steam Cooling RPV Water Level (MSCRWL) (Ref. 4). Under these conditions, a ATWS greater then design decay heat level event exists and requires at least an Alert classification in accordance with the RPS Failure EALs. Generic This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold 1.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. to NRC-15-0061 Page 258 Fermi 2 Emergency Action Level Technical Bases This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term cannot be restored and maintained above means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. to NRC-15-0061 Page 259 Fermi 2 Emergency Action Level Technical Bases Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress. & Steam Cooling

3. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress. & Steam Cooling - AWS

4. EOP 29.100.01A Sheet 1 RPV Control - ATWS

5. NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A to NRC-15-0061 Page 260 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 2. RCS Leak Rate Degradation Threat: Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 261 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 2. RCS Leak Rate Degradation Threat: Potential Loss Threshold: None Definition(s): None Basis: N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 262 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 3. PC Conditions Degradation Threat: Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 263 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 3. PC Conditions Degradation Threat: Potential Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 264 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 4. PC Radiation / RCS Activity Degradation Threat: Loss Threshold: A. CHRRM reading > 2.25E+3 R/hr Definition(s): None Basis: Plant-Specific For Fermi 2, the Containment High Range Radiation Monitor (CHRRM) is used to measure drywell radiation levels. A valid CHRRM reading of 2.25E+3 R/hr corresponds to 2.5% gap release (300 µCi/gm DEI I-131) discharged instantaneously into containment atmosphere (Ref. 1). Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Radiation. to NRC-15-0061 Page 265 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1402 Fermi EAL Technical Bases Calculations - CHRRM Series Rev. 0

2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A to NRC-15-0061 Page 266 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 4. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

B. Primary coolant activity > 300 µCi/gm DEI-131 Definition(s): None Basis: Plant-Specific 300 Ci/gm DEI-131 is equivalent to 2.5% fuel clad (gap) damage (Ref. 1). Generic This threshold indicates that RCS radioactivity concentration is greater than 300 Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. There is no Potential Loss threshold associated with RCS Activity. There is no Potential Loss threshold associated with Primary Containment Radiation. Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1402 Fermi EAL Technical Bases Calculations - CHRRM Series Rev. 0

2. NEI 99-01 RCS Activity Fuel Clad Loss 1.A to NRC-15-0061 Page 267 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 4. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 268 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 5. PC Integrity or Bypass Degradation Threat: Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 269 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 5. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 270 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 6. ED Judgment Degradation Threat: Loss Threshold: A. ANY condition in the opinion of the Emergency Director that indicates loss of the fuel clad barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A to NRC-15-0061 Page 271 Fermi 2 Emergency Action Level Technical Bases Barrier: Fuel Clad Category: 6. ED Judgment Degradation Threat: Potential Loss Threshold:

A. ANY condition in the opinion of the Emergency Director that indicates potential loss of the fuel clad barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A to NRC-15-0061 Page 272 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 1. RPV Water Level Degradation Threat: Loss Threshold:

A. RPV water level cannot be restored and maintained above 0 in. (TAF) or cannot be determined. Definition(s): None Basis: Plant-Specific An RPV water level instrument reading of 0 in. indicates RPV water level is at the top of active fuel. When RPV water level is at or above the top of active fuel, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV water level is below the top of active fuel following depressurization of the RPV (automatically, manually or by failure of the RCS barrier), the uncovered portion of the core must be cooled by less reliable means (i.e., spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling (Ref. 1). Consistent with the EOP definition of cannot be restored and maintained, the determination that RPV water level cannot be restored and maintained above the top of active fuel may be made at, before, or after RPV water level actually decreases to this point. (Ref. 1) When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPV water level indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV water level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in EOP RPV Flooding and EOP RPV Flooding - ATWS specify these means, which include emergency depressurization of the RPV and to NRC-15-0061 Page 273 Fermi 2 Emergency Action Level Technical Bases injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in ATWS events). (Ref. 2, 3). If RPV water level cannot be restored and maintained above the top of active fuel or RPV water level cannot be determined with respect to the top of active fuel, a loss of the RCS barrier exists. Note that EOP RPV Control - ATWS may require intentional uncovery of the core and control of RPV water level between 0 in. and -25 in., the Minimum Steam Cooling RPV Water Level (MSCRWL) (Ref. 4). Under these conditions, an ATWS above design decay heat level event exists and requires at least an Alert classification in accordance with the RPS Failure EALs. Generic This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 1.A. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure to NRC-15-0061 Page 274 Fermi 2 Emergency Action Level Technical Bases injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress. & Steam Cooling

3. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress, & Steam Cooling - AWS

4. EOP 29.100.01A Sheet 1 RPV Control - ATWS

5. NEI 99-01 RPV Water Level RCS Loss 2.A to NRC-15-0061 Page 275 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 1. RPV Water Level Degradation Threat: Potential Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 276 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 2. RCS Leak Rate Degradation Threat: Loss Threshold: A. UNISOLABLE break in ANY of the following: Main Steam Line HPCI Steam Line RCIC Steam Line RWCU Feedwater Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Plant-Specific The list of systems included in this threshold are the high energy lines which, if ruptured and remain unisolated, can rapidly depressurize the RPV. These lines are typically isolated by actuation of the Leak Detection system. Large high-energy line breaks such as Main Steam Line (MSL), High Pressure Coolant Injection (HPCI), Feedwater (failure of non-return valves), Reactor Water Cleanup (RWCU) or Reactor Core Isolation Cooling (RCIC) that are UNISOLABLE represent a significant loss of the RCS barrier. Determination of whether the leak is isolated to preclude EAL declaration must occur within the 15-minute assessment period. (Ref.1) The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside Primary Containment exists when flow is not prevented by downstream isolations. Emergency declaration under this threshold would not be required in the case of a failure of both isolation valves to close but no downstream flowpath exists. Similarly, if the emergency response requires the normal process flow of a system outside Primary Containment (e.g., EOP requirement to bypass to NRC-15-0061 Page 277 Fermi 2 Emergency Action Level Technical Bases MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Primary Containment (see PC Loss 5.A) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). Generic Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met. Fermi Basis Reference(s):

1. UFSAR Section 6.2.4.2.2.2.2 Effluent Lines

2. NEI 99-01 RCS Leak Rate RCS Loss 3.A to NRC-15-0061 Page 278 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 2. RCS Leak Rate Degradation Threat: Loss Threshold:

B. Emergency RPV Depressurization is required Definition(s): None Basis: Plant-Specific Emergency RPV Depressurization is specified in the EOP flowcharts (EOP Emergency Depressurization) when symbols containing the phrase EMERGENCY RPV DEPRESS IS REQ'D are reached. (Ref. 1, 2). Generic Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress, & Steam Cooling

3. NEI 99-01 RCS Leak Rate RCS Loss 3.B to NRC-15-0061 Page 279 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 2. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

A. UNISOLABLE primary system leakage into Secondary Containment that results in exceeding EITHER of the following:

1. One or more Secondary Containment Control Max Normal Operating Temperatures (EOP Table 12)

OR 1.2. One or more Secondary Containment Control Max Normal Operating Area Radiation Levels (EOP Table 14) Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Plant-Specific The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of UNISOLABLE primary system leakage outside the Primary Containment. When parameters reach the threshold level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. Determination of whether the leak is isolated to preclude EAL declaration must occur within the 15-minute assessment period. (Ref. 1, 2) The only Secondary Containment Maximum Safe Operating Radiation Level that can be determined remotely in the Control Room is the RBSB Torus Room on ARM Channel 14. No other Secondary Containment Maximum Safe Operating Radiation Levels (> 5 R/hr) can be determine by installed area radiation monitors due to instrument range limitations. Therefore the area radiation threshold (other than for the RBSB Torus Room) for the Primary Containment Potental Loss based on RCS leak rate must be determined via local survey. to NRC-15-0061 Page 280 Fermi 2 Emergency Action Level Technical Bases In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. Generic Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment. A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 5 Secondary Containment and Rad Release

3. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A to NRC-15-0061 Page 281 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 3. PC Conditions Degradation Threat: Loss Threshold:

A. Drywell pressure > 1.68 psig due to RCS leakage Definition(s): None Basis: Plant-Specific The drywell high pressure scram setpoint is an entry condition to the EOP flowcharts: EOP RPV Control, and EOP Primary Containment Control (Ref. 1, 2, 3). Normal Primary Containment (PC) pressure control functions such as operation of drywell cooling and venting are specified in EOP Primary Containment Control in advance of less desirable but more effective functions such as operation of drywell or torus sprays. Primary Containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control Primary Containment vent/purge (Ref. 1). The threshold phrase due to RCS leakage focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect Primary Containment pressure. Primary Containment pressure greater than 1.68 psig with corollary indications (e.g., elevated drywell temperature, indications of loss of RCS inventory) should, therefore, be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS barrier loss. to NRC-15-0061 Page 282 Fermi 2 Emergency Action Level Technical Bases Generic 1.68 psig is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system. There is no Potential Loss threshold associated with Primary Containment Pressure. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 1 RPV Control

3. EOP 29.100.01 Sheet 2 Primary Containment Control

4. NEI 99-01 Primary Containment Pressure RCS Loss 1.A to NRC-15-0061 Page 283 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 3. PC Conditions Degradation Threat: Potential Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 284 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 4. PC Radiation / RCS Activity Degradation Threat: Loss Threshold: A. CHRRM reading > 8.72E+1 R/hr Definition(s): None Basis: Plant-Specific For Fermi 2, the Containment High Range Radiation Monitor (CHRRM) is used to measure drywell radiation levels. A valid CHRRM reading of 8.72E+1 R/hr corresponds to normal coolant activity discharged instantaneously into containment atmosphere (Ref. 1). Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation. Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1402 Fermi EAL Technical Bases Calculations - CHRRM Series Rev. 0

2. NEI 99-01 Primary Containment Radiation RCS Loss 4.A to NRC-15-0061 Page 285 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 4. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 286 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 5. PC Integrity or Bypass Degradation Threat: Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 287 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 5. PC Integrity or Bypass Degradation Threat: Potential Loss Threshold: None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 288 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 6. ED Judgment Degradation Threat: Loss Threshold: A. Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A to NRC-15-0061 Page 289 Fermi 2 Emergency Action Level Technical Bases Barrier: Reactor Coolant System Category: 6. ED Judgment Degradation Threat: Potential Loss Threshold:

A. Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A to NRC-15-0061 Page 290 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 1. RPV Water Level Degradation Threat: Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 291 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 1. RPV Water Level Degradation Threat: Potential Loss Threshold: A. SAG entry is requiredInadequate core cooling as indicated by ANY of the following:

1. RPV water level cannot be restored and maintained > -48 in. with 5725 gpm Core Spray loop flow OR

2. RPV water level cannot be restored and maintained > -25 in. with

      < 5725 gpm Core Spray loop flow OR 1.3. RPV water level cannot be determined and core damage is occurring Definition(s):

None Basis: Plant-Specific Requirements for Primary Containment Flooding correspond to entry into the Severe Accident Guidelines (SAGs) and are established in EOP RPV Control, EOP RPV Control - ATWS and EOP RPV Flooding & EOP RPV Flooding - ATWS (Ref. 1, 2, 3, 4, 5). These EOPs provide instructions to ensure adequate core cooling by maintaining RPV water level above prescribed limits or operating sufficient RPV injection sources when level cannot be determined. Direction is specified that SAG entry is required represents the requirement for Primary Containment Flooding when: EOP RPV Control - RPV water level cannot be restored and maintained above

      -48 in. with required Core Spray flow.

EOP RPV Control - RPV water level cannot be restored and maintained above -25 in. (MSCRWL) with insufficient Core Spray flow. to NRC-15-0061 Page 292 Fermi 2 Emergency Action Level Technical Bases EOP RPV Control - ATWS - RPV water level cannot be restored and maintained above -25 in. (MSCRWL) EOP Flooding & EOP Flooding - ATWS - RPV water level cannot be determined and it is determined that core damage is occurring. This threshold is also a Loss of the Fuel Clad barrier (FC Loss 1.A). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RCS Loss 1.A). Primary Containment Flooding (SAG entry), therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification. Generic The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement of SAG entry is required (Primary Containment Flooding) indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require primary containment flooding. When primary containment flooding SAG entry is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling. PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency. Fermi Basis Reference(s):

1. EOP 29.100.01 Sheet 1 RPV Control

2. EOP 29.100.01A Sheet 1 RPV Control - ATWS

3. EOP 29.100.01 Sheet 3 RPV Flooding, Emerg. Depress. & Steam Cooling

4. EOP 29.100.01 Sheet 3A RPV Flooding, Emerg. Depress. & Steam Cooling

5. EOP Support Documentation Section 1 Plant Specific Technical Guideline

6. NEI 99-01 RPV Water Level PC Potential Loss 2.A to NRC-15-0061 Page 293 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 2. RCS Leak Rate Degradation Threat: Loss Threshold:

A. UNISOLABLE primary system leakage into Secondary Containment that results in exceeding ANY of the following:

1. One or more Secondary Containment Control Max Safe Operating Temperatures (EOP Table 12)

OR 1.2. One or more Secondary Containment Control Max Safe Operating Area Radiation Levels (EOP Table 14)area radiation levels (Table F-2) Table F-2 Secondary Containment Area Radiation Levels Area Rad Levels ARM Channel Rad Level RBSB SE Corner 7 950 mR/hr RBSB SW Corner 8 950 mR/hr RBSB NW Corner 9 950 mR/hr RBSB NE Corner 10 950 mR/hr SB HPCI Room 11 950 mR/hr RBSB Torus Room 14 5,000 mR/hr 1st Floor RB DW Airlock 31 950 mR/hr Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Plant-Specific to NRC-15-0061 Page 294 Fermi 2 Emergency Action Level Technical Bases The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of UNISOLABLE primary system leakage outside the Primary Containment. When parameters reach the threshold level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. Determination of whether the leak is isolated to preclude EAL declaration must occur within the 15-minute assessment period. (Ref. 1, 2) The only Secondary Containment Maximum Safe Operating Radiation Level that can be determined remotely in the Control Room is the RBSB Torus Room on ARM Channel 14. No other Secondary Containment Maximum Safe Operating Radiation Levels (> 5 R/hr) can be determine by installed area radiation monitors due to instrument range limitations. Therefore the area radiation thresholds (other than for the RBSB Torus Room) for the Primary Containment Potential Loss based on RCS leak rate must be determined via local survey. have been limited to those area monitors that have an upper range of 1,000 mR/hr and the thresholds reduced to a value that can be determined from the control room (95% of scale). These values are provided in Table F-2. In general, multiple indications should be used to determine if a primary system is discharging outside Primary Containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment. Generic The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe to NRC-15-0061 Page 295 Fermi 2 Emergency Action Level Technical Bases shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS Potential Loss 2.A this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation Failure. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 5 Secondary Containment and Rad Release

3. NEI 99-01 RCS Leak Rate PC Loss 3.C to NRC-15-0061 Page 296 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 2. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 297 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 3. PC Conditions Degradation Threat: Loss Threshold: A. UNPLANNED rapid drop in Primary Containment pressure following Primary Containment pressure rise Definition(s): UNPLANNED - A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: Plant-Specific None Generic Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity. This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Fermi Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A to NRC-15-0061 Page 298 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 3. PC Conditions Degradation Threat: Loss Threshold:

B. Primary Containment pressure response not consistent with LOCA conditions Definition(s): None Basis: Plant-Specific Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity. USAR Sections 6.2.1.3.2 and 6.2.1.3.3 provide a summary of Primary Containment pressure response for the design basis loss of coolant accident and the conditions resulting in the release of RCS inventory to the containment (Ref. 1, 3). The maximum calculated drywell pressure is approximately 50 psig and then stablizesstabilizes at approximately 30 psig with torus pressure at approximately 25 psig 30 seconds after the break (Ref. 2). These pressures are well below the design allowable drywell pressure of 62 psig. (Ref. 1) Due to conservatisms in LOCA analyses, actual pressure response is expected to be less than the analyzed response. LOCA conditions are manifested on Control Room instrumentation by drywell pressure rising with torus pressure following and eventually equalizing (around 18 psig for the DBA LOCA) (Ref. 3, 4). Generic Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity. to NRC-15-0061 Page 299 Fermi 2 Emergency Action Level Technical Bases This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Fermi Basis Reference(s):

1. USAR Section 6.2.1.3.2 Recirculation Line Break - Short Term Response

2. USAR Figure 6.2-9 Recirculation Line Break Primary Containment Initial Pressure Transient (3499 MWT)

3. USAR Section 6.2.1.3.3 Recirculation Line Break - Long Term Response

4. USAR Figure 6.2-11 Primary Containment Pressure Long Term Response (3499 MWT)

5. NEI 99-01 Primary Containment Conditions PC Loss 1.B to NRC-15-0061 Page 300 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 3. PC Conditions Degradation Threat: Potential Loss Threshold:

A. Primary Containment Pressure > 62 psig Definition(s): None Basis: Plant-Specific The Primary Containment pressure of 62 psig is based on the primary containment design pressure (Ref. 1). Generic The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier. Fermi Basis Reference(s):

1. UFSAR Table 6.2-4 Drywell to Suppression Chamber Vacuum Breaker Valve Data

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.A to NRC-15-0061 Page 301 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 3. PC Conditions Degradation Threat: Potential Loss Threshold:

B. > 6% H2 AND > 5% O2 in EITHER the drywell or suppression chamber Definition(s): None Basis: Plant-Specific Explosive (deflagration) mixtures in the Primary Containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to Primary Containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (Ref. 1). Except for brief periods during plant startup and shutdown, oxygen concentration in the Primary Containment is maintained at insignificant levels by nitrogen inertion. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, Ref. 1) and readily recognizable because 6% hydrogen is well above the EOP Primary Containment Control entry condition of 1% (Ref. 1, 2). If the hydrogen or oxygen monitor is unavailable, sampling and analysis may determine gas concentrations. The validity of sample results must be judged based upon plant conditions, since drawing and analyzing samples may take some time. If sample results to NRC-15-0061 Page 302 Fermi 2 Emergency Action Level Technical Bases cannot be relied upon and hydrogen concentrations cannot be determined by any other means, the concentrations must be considered "unknown." The monitors should not be considered "unavailable" until an attempt has been made to place them in service. Generic If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 2 Primary Containment Control

3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.B to NRC-15-0061 Page 303 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 3. PC Conditions Degradation Threat: Potential Loss Threshold:

C. EOP Heat Capacity Limit (HCL) exceeded Definition(s): None Basis: Plant-Specific The Heat Capacity Limit (HCL) is given in EOP Curves, Cautions and Tables (Ref. 1). This threshold should be considered when EOP Primary Containment Control Step TWT-5 is reached and emergency RPV depressurization is required (Ref. 2). Generic The Heat Capacity Temperature Limit (HCL) is the highest torus water temperature from which Emergency RPV Depressurization will not raise: Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent. The HCL is a function of RPV pressure, torus water temperature and torus water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment. to NRC-15-0061 Page 304 Fermi 2 Emergency Action Level Technical Bases Fermi Basis Reference(s):

1. EOP 29.100.01 Sheet 6 Curves, Cautions and Tables

2. EOP 29.100.01 Sheet 2 Primary Containment Control

3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C to NRC-15-0061 Page 305 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 4. PC Radiation / RCS Activity Degradation Threat: Loss Threshold:

None Definition(s): None Basis: Plant-Specific N/A Generic N/A Fermi Basis Reference(s): N/A to NRC-15-0061 Page 306 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 4. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold: A. CHRRM reading > 1.79E+4 R/hr Definition(s): None Basis: Plant-Specific The "potential loss" EAL value corresponds to at least 20% clad damage with release into the primary containment. This EAL corresponds to loss of both the Fuel Clad and RCS barriers with Potential Loss of the Primary Containment barrier, and would result in declaration of a General Emergency. For Fermi 2, the Containment High Range Radiation Monitor (CHRRM) is used to measure drywell radiation levels. A valid CHRRM reading of 1.79E+4 R/hr corresponds to 20% gap release discharged instantaneously into containment atmosphere (Ref. 1). Generic The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of to NRC-15-0061 Page 307 Fermi 2 Emergency Action Level Technical Bases containment which would then escalate the emergency classification level to a General Emergency. Fermi Basis Reference(s):

1. EP-EALCALC-FERMI-1402 Fermi EAL Technical Bases Calculations - CHRRM Series Rev. 0

2. NEI 99-01 Primary Containment Radiation PC Potential Loss 4.A to NRC-15-0061 Page 308 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 5. PC Integrity or Bypass Degradation Threat: Loss Threshold:

A. UNISOLABLE direct downstream pathway to the environment exists after Primary Containment isolation signal Definition(s): UNISOLABLE - An open or breached system line that cannot be isolated, remotely or locally. Basis: Plant-Specific This Primary Containment isolation failure threshold is based on failure to successfully isolate the Primary Containment following a valid isolation signal (RPV water level 1, 2 or 3 or high drywell pressure) resulting in a direct downstream pathway to the environment for any of the following containment isolation signals (Ref. 1): Group 1 - Main Steam System Group 2 - Reactor Water Sample System Group 4 - RHR Shutdown Cooling and Head Vent Group 6 - HPCI Steam Supply Line Group 8 - RCIC Steam Supply Line Group 10/11 - Reactor Water Cleanup System Inboard/Outboard Group 13 - Drywell Sumps Group 14 - Drywell and Suppression Pool Ventilation These systems, protected by the Primary Containment Isolation System, provide potential direct (non-liquid interfacing system) release pathways from the RCS or Primary Containment atmosphere to the environment should the isolation be unsuccessful. to NRC-15-0061 Page 309 Fermi 2 Emergency Action Level Technical Bases Determination of whether the leak is isolated to preclude EAL declaration must occur within the 15-minute assessment period. (Ref. 1) Generic The use of the modifier direct in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category RICs. Fermi Basis Reference(s):

1. AOP 20.000.21 Reactor Scram Attachment 1 Isolations and Actuations

2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A to NRC-15-0061 Page 310 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 5. PC Integrity or Bypass Degradation Threat: Loss Threshold:

B. Intentional Primary Containment venting, irrespective of offsite radioactivity release rates, per EOPs Definition(s): None Basis: Plant-Specific EOP Primary Containment Control may specify Primary Containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (Ref. 1, 2). The threshold is met when the operator begins venting the Primary Containment in accordance with 29.ESP.07 , not when actions are taken to bypass interlocks prior to opening the vent valves (Ref. 3). Purge and vent actions specified in EOP Primary Containment Control step PCP-1 to control Primary Containment pressure below the drywell high pressure scram setpoint do not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM limits (Ref. 1, 2, 3). Generic EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed. Intentional venting of primary containment for primary containment pressure control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control to NRC-15-0061 Page 311 Fermi 2 Emergency Action Level Technical Bases pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. Fermi Basis Reference(s):

1. EOP Support Documentation Section 1 Plant Specific Technical Guideline

2. EOP 29.100.01 Sheet 2 Primary Containment Control

3. 29.ESP.07 Primary Containment Venting

4. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.B to NRC-15-0061 Page 312 Fermi 2 Emergency Action Level Technical Bases Barrier: Primary Containment Category: 6. ED Judgment Degradation Threat: Loss Threshold:

A. Any condition in the opinion of the Emergency Director that indicates loss of the Primary Containment barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Primary Containment Barrier is lost. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A to NRC-15-0061 Page 313 Fermi 2 Emergency Action Level Technical Bases Barrier: Containment Category: 5. ED Judgment Degradation Threat: Potential Loss Threshold:

A. Any condition in the opinion of the Emergency Director that indicates potential loss of the Primary Containment barrier Definition(s): None Basis: Plant-Specific None Generic This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Primary Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. Fermi Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.A

Enclosure 3 to NRC-15-0061 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Revised (Clean) Emergency Action Level Technical Bases to be incorporated into Implementing Procedure EP-101, Classification of Emergencies to NRC-15-0061 Page 1 Fermi 2 Emergency Action Level Technical Bases TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ................................................................................................... 3 2.0 DISCUSSION .............................................................................................. 3 2.1 Background ....................................................................................... 3 2.2 Fission Product Barrier Thresholds ................................................... 4 2.3 Fission Product Barrier Classification Criteria ................................... 5 2.4 EAL Organization .............................................................................. 5 2.5 Technical Bases Information ............................................................. 8 2.6 Operating Mode Applicability ............................................................ 9 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ................... 11 3.1 General Considerations .................................................................... 11 3.1.1 Classification Timeliness ..................................................... 11 3.1.2 Valid Indications .................................................................. 11 3.1.3 Imminent Conditions ............................................................ 12 3.1.4 Planned vs. Unplanned Events............................................ 12 3.1.5 Classification Based on Analysis ......................................... 12 3.1.6 Emergency Director Judgment ............................................ 13 3.2 Classification Methodology ............................................................... 13 3.2.1 Classification of Multiple Events and Conditions ................. 13 3.2.2 Consideration of Mode Changes During Classification........ 14 3.2.3 Classification of Imminent Conditions .................................. 14 3.2.4 Emergency Classification Level Upgrading and Downgrading ............................................................... 15 3.2.5 Classification of Short-Lived Events .................................... 15 3.2.6 Classification of Transient Conditions.................................. 15 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition ............................................................................. 16 3.2.8 Retraction of an Emergency Declaration ............................. 17 to NRC-15-0061 Page 2 Fermi 2 Emergency Action Level Technical Bases TABLE OF CONTENTS SECTION TITLE PAGE

4.0 REFERENCES

............................................................................................ 17 4.1 Developmental .................................................................................. 17 4.2 Implementing .................................................................................... 18 4.3 Commitments .................................................................................... 18 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS ...................................... 19 5.1 Definitions ......................................................................................... 19 5.2 Acronyms & Abbreviations ................................................................ 25 6.0 FERMI-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE .......................... 28 7.0 ATTACHMENTS .......................................................................................... 32 7.1 Attachment 1 - Emergency Action Level Technical Bases ............... 33 Category R Abnormal Rad Levels / Rad Effluent ............................. 34 Category C Cold Shutdown / Refueling System Malfunction ........... 83 Category H Hazards and Other Conditions Affecting Plant Safety .. 134 Category S System Malfunction....................................................... 181 Category F Fission Product Barrier Degradation ............................. 237 Category E ISFSI ............................................................................. 244 7.2 Attachment 2 - Fission Product Barrier Matrix and Bases ................. 247 to NRC-15-0061 Page 3 Fermi 2 Emergency Action Level Technical Bases 1.0 PURPOSE This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for the Enrico Fermi Unit 2 Power Plant (Fermi 2). It should be used to facilitate review of the Fermi 2 EALs and provide historical documentation for future reference. Decision-makers responsible for implementation of EP-101, Classification of Emergencies, may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Director in making classifications, particularly those involving judgment or multiple events. The basis information may also be useful in training and for explaining event classifications to off-site officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification. Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). 2.0 DISCUSSION 2.1 Background EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the Fermi 2 Emergency Plan. In 1992, the NRC endorsed NUMARC/NESP-007 Methodology for Development of Emergency Action Levels as an alternative to NUREG-0654 EAL guidance. NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included: to NRC-15-0061 Page 4 Fermi 2 Emergency Action Level Technical Bases Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions. Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSIs). Simplifying the fission product barrier EAL threshold for a Site Area Emergency. Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, November 2012 (ADAMS Accession Number ML13091A209) (Ref. 4.1.1), Fermi 2 conducted an EAL implementation upgrade project that produced the EALs discussed herein. 2.2 Fission Product Barrier Thresholds Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. Loss and Potential Loss signify the relative damage and threat of damage to the barrier. A Loss threshold means the barrier no longer assures containment of radioactive materials. A Potential Loss threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier. The primary fission product barriers are: A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets. to NRC-15-0061 Page 5 Fermi 2 Emergency Action Level Technical Bases B. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves. C. Primary Containment (PC): The Primary Containment Barrier includes the drywell, the suppression pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the Emergency Classification Level (ECL) from Alert to a Site Area Emergency or a General Emergency. 2.3 Fission Product Barrier Classification Criteria The following criteria are the bases for event classification related to fission product barrier loss or potential loss: Alert: Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency: Loss or potential loss of any two barriers General Emergency: Loss of any two barriers and loss or potential loss of the third barrier 2.4 EAL Organization The Fermi 2 EAL scheme includes the following features: Division of the EAL set into three broad groups: o EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered. o EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operation mode. to NRC-15-0061 Page 6 Fermi 2 Emergency Action Level Technical Bases o EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refueling or Defueled mode. The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency. Within each group, assignment of EALs to categories and subcategories: o Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The Fermi 2 EAL categories are aligned to and represent the NEI 99-01Recognition Categories. o Subcategories are used in the Fermi 2 scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The Fermi 2 EAL categories and subcategories are listed below. to NRC-15-0061 Page 7 Fermi 2 Emergency Action Level Technical Bases EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory Any Operating Mode: R - Abnormal Rad Levels / Rad Effluent 1 - Radiological Effluent 2 - Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 - Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4 - Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - ED Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI) Hot Conditions: S - System Malfunction 1 - Loss of Essential AC Power 2 - Loss of Vital DC Power 3 - Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier Degradation None Cold Conditions: C - Cold Shutdown / Refueling System 1 - RPV Level Malfunction 2 - Loss of Essential AC Power 3 - RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems to NRC-15-0061 Page 8 Fermi 2 Emergency Action Level Technical Bases The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL Technical Bases Document in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 3.0 and Attachments 1 & 2 of this document for such information. 2.5 Technical Bases Information EAL technical bases are provided in Attachment 1 for each EAL according to EAL group (All, Hot, Cold), EAL category (R, C, H, S, F and E) and EAL subcategory. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided: Category Letter & Title Subcategory Number & Title Initiating Condition (IC) Site-specific description of the generic IC given in NEI 99-01 Rev. 6. EAL Identifier (enclosed in rectangle) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

1. First character (letter): Corresponds to the EAL category as described above (R, C, H, S, F or E)

2. Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A = Alert U = Unusual Event

3. Third character (number): Subcategory number within the given category.

Subcategories are sequentially numbered beginning with the number one (1). If to NRC-15-0061 Page 9 Fermi 2 Emergency Action Level Technical Bases a category does not have a subcategory, this character is assigned the number one (1).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Classification (enclosed in rectangle): Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G) EAL (enclosed in rectangle) Exact wording of the EAL as it appears in the EAL Classification Matrix Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operation, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, D - Defueled, or All. (See Section 2.6 for operating mode definitions) Definitions: If the EAL wording contains a defined term, the definition of the term is included in this section. These definitions can also be found in Section 5.1. Basis: A Plant-Specific basis section that provides Fermi-relevant information concerning the EAL. This is followed by a Generic basis section that provides a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6. Fermi Basis Reference(s): Site-specific source documentation from which the EAL is derived 2.6 Operating Mode Applicability (Ref. 4.1.7) 1 Power Operation Mode Switch in Run to NRC-15-0061 Page 10 Fermi 2 Emergency Action Level Technical Bases 2 Startup Mode Switch in Startup/Hot Standby or Mode Switch Refuel with all reactor vessel head closure bolts fully tensioned. 3 Hot Shutdown Mode Switch in Shutdown and RCS temperature > 200°F 4 Cold Shutdown Mode Switch in Shutdown and RCS temperature 200°F. 5 Refueling Mode Switch in Shutdown or Refuel with one or more reactor vessel head closure bolts less then fully tensioned. D Defueled All nuclear fuel removed from reactor vessel (i.e., full core off load during refueling or extended outage). The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred. to NRC-15-0061 Page 11 Fermi 2 Emergency Action Level Technical Bases 3.0 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 3.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds. 3.1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (Ref. 4.1.12). 3.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicators operability, the conditions existence, or the reports accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment. to NRC-15-0061 Page 12 Fermi 2 Emergency Action Level Technical Bases 3.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. 3.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (Ref. 4.1.4). 3.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). to NRC-15-0061 Page 13 Fermi 2 Emergency Action Level Technical Bases 3.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. 3.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process clock starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process clock started. When assessing an EAL that specifies a time duration for the off-normal condition, the clock for the EAL time duration runs concurrently with the emergency classification process clock. For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (Ref. 4.1.14). 3.2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example: If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared. to NRC-15-0061 Page 14 Fermi 2 Emergency Action Level Technical Bases There is no additive effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (Ref. 4.1.2). 3.2.2 Consideration of Mode Changes During Classification The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 3.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. to NRC-15-0061 Page 15 Fermi 2 Emergency Action Level Technical Bases 3.2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (Ref. 4.1.2). 3.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram. 3.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the to NRC-15-0061 Page 16 Fermi 2 Emergency Action Level Technical Bases applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example: An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a grace period during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 3.2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (Ref. 4.1.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (Ref. 4.1.4) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. to NRC-15-0061 Page 17 Fermi 2 Emergency Action Level Technical Bases 3.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (Ref. 4.1.3).

4.0 REFERENCES

4.1 Developmental 4.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML13091A209 4.1.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007. 4.1.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 4.1.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 4.1.5 10 § CFR 50.73 License Event Report System 4.1.6 Fermi 2 Offsite Dose Calculation Manual (ODCM) TRM Volume II Figure 3.0-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents 4.1.7 Technical Specifications Table 1.1-1 Modes 4.1.8 EP-101 Classification of Emergencies, Rev. 39 4.1.9 Technical Specifications Section 3.6 Containment Systems 4.1.10 UFSAR Figure 1.2-5 Site Plot Plan 4.1.11 WG-001 Fermi Writers Guide 4.1.12 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 4.1.13 Fermi Certificate of Compliance No. 114 Appendix A Technical Specifications for the HI-STORM 100 Cask System to NRC-15-0061 Page 18 Fermi 2 Emergency Action Level Technical Bases 4.2 Implementing 4.2.1 EP-101Classification of Emergencies 4.2.2 NEI 99-01 Rev. 6 to Fermi EAL Comparison Matrix 4.2.3 Fermi EAL Matrix 4.3 Commitments 4.3.1 None to NRC-15-0061 Page 19 Fermi 2 Emergency Action Level Technical Bases 5.0 DEFINITIONS, ACRONYMS & ABBREVIATIONS 5.1 Definitions (Ref. 4.1.1 except as noted) Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below. Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Confinement Boundary The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. The Multi-Purpose Canister (MPC) serves as the Confinement Boundary for contained radioactive materials (Ref. 4.1.13) Containment Closure The conditions or actions taken to secure Primary or Secondary Containment and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. For Fermi 2, this condition is met if either Primary Containment or Secondary Containment are functional (i.e. intact). Emergency Action Level A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. Emergency Classification Level One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual to NRC-15-0061 Page 20 Fermi 2 Emergency Action Level Technical Bases effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Unusual Event (UE), Alert, Site Area Emergency (SAE) and General Emergency (GE). EPA PAGs Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires Fermi 2 to recommend protective actions for the general public to offsite planning agencies. Explosion A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present. Fire Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed. Fission Product Barrier Threshold A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Flooding A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. General Emergency Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE to NRC-15-0061 Page 21 Fermi 2 Emergency Action Level Technical Bases ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Hostage A person(s) held as leverage against the station to ensure that demands will be met by the station. Hostile Action An act toward Fermi 2 or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Fermi 2. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area) (Ref. 4.1.8). Hostile Force One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction. Imminent The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. Independent Spent Fuel Storage Installation (ISFSI) A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. to NRC-15-0061 Page 22 Fermi 2 Emergency Action Level Technical Bases Initiating Condition An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences. Intrusion The act of entering without authorization. Discovery of a bomb in a specified area is indication of intrusion into that area by a hostile force. Normal Levels As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value. Owner Controlled Area The company property immediately surrounding the PROTECTED AREA security fence. Access is normally limited to people on official business (Ref. 4.1.8). Projectile An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety. Protected Area An area encompassed by physical barriers and to which access is controlled. The Protected Area refers to the designated security area around the process buildings and is depicted in UFSAR Figure 1.2-5 Site Plot Plan (Ref. 4.1.10). RCS Intact The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). Refueling Pathway The reactor refueling cavity, spent fuel pool and fuel transfer canal (cattle chute) comprise the refueling pathway. Safety System to NRC-15-0061 Page 23 Fermi 2 Emergency Action Level Technical Bases A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures. Security Condition Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a hostile action. Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary. Site Boundary That line beyond which the land is neither owned, nor leased, nor otherwise controlled by DTE Electric. (Ref. 4.1.6). Unisolable An open or breached system line that cannot be isolated, remotely or locally. Unplanned to NRC-15-0061 Page 24 Fermi 2 Emergency Action Level Technical Bases A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Valid An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicators operability, the conditions existence, or the reports accuracy is removed. Implicit in this definition is the need for timely assessment. Visible Damage Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. to NRC-15-0061 Page 25 Fermi 2 Emergency Action Level Technical Bases 5.2 Acronyms & Abbreviations °F ........................................................................................................Degrees Fahrenheit ° ............................................................................................................................Degrees

................................................................................................................ Feet or Minutes
........................................................................................................... Inches or Seconds
........................................................................................................................... Gamma n ........................................................................................................................... Neutron AC ....................................................................................................... Alternating Current AOP .................................................................................. Abnormal Operating Procedure APRM ............................................................................... Average Power Range Monitor ATWS ...................................................................... Anticipated Transient Without Scram BWR ............................................................................................... Boiling Water Reactor BWROG ................................................................. Boiling Water Reactor Owners Group CHRRM ............................................... Containment Area High Range Radiation Monitor CDE ....................................................................................... Committed Dose Equivalent CFR ...................................................................................... Code of Federal Regulations CS ................................................................................................................... Core Spray CW ......................................................................................................... Circulating Water DBA ................................................................................................Design Basis Accident DC ............................................................................................................... Direct Current EAL .............................................................................................Emergency Action Level ECCS ........................................................................... Emergency Core Cooling System ECL ................................................................................. Emergency Classification Level ED ......................................................................................................Emergency Director EOF .................................................................................. Emergency Operations Facility EOP ............................................................................... Emergency Operating Procedure EPA .............................................................................. Environmental Protection Agency EPIP ................................................................ Emergency Plan Implementing Procedure EPRI ............................................................................. Electric Power Research Institute ERG ............................................................................... Emergency Response Guideline ESF ................................................................................ Engineered Safeguards Feature FAQ ........................................................................................ Frequently Asked Question FEMA .............................................................. Federal Emergency Management Agency GE ..................................................................................................... General Emergency HCL .................................................................................................... Heat Capacity Limit HPCI ................................................................................ High Pressure Coolant Injection HSI ............................................................................................. Human System Interface to NRC-15-0061 Page 26 Fermi 2 Emergency Action Level Technical Bases IC .......................................................................................................... Initiating Condition ID .............................................................................................................. Inside Diameter IPEEE ................. Individual Plant Examination of External Events (Generic Letter 88-20)

ISFSI ........................................................... Independent Spent Fuel Storage Installation Keff ......................................................................... Effective Neutron Multiplication Factor LCO .................................................................................. Limiting Condition of Operation LOCA .........................................................................................Loss of Coolant Accident LPCI ................................................................................. Low Pressure Coolant Injection MCR ................................................................................................... Main Control Room MPC ................................... Maximum Permissible Concentration/Multi-Purpose Canister MPH .......................................................................................................... Miles Per Hour MSIV ...................................................................................... Main Steam Isolation Valve MSL ........................................................................................................ Main Steam Line mR, mRem, mrem, mREM ............................................... milli-Roentgen Equivalent Man MW .....................................................................................................................Megawatt NEI .............................................................................................. Nuclear Energy Institute NPP ...................................................................................................Nuclear Power Plant NRC ............................................................................... Nuclear Regulatory Commission NSSS ................................................................................ Nuclear Steam Supply System NORAD .................................................. North American Aerospace Defense Command (NO)UE............................................................................. (Notification Of) Unusual Event NUMARC1 ................................................. Nuclear Management and Resources Council OBE ...................................................................................... Operating Basis Earthquake OCA .............................................................................................. Owner Controlled Area ODCM/ODAM.......................................... Offsite Dose Calculation (Assessment) Manual ORO ................................................................................ Off-site Response Organization PA ............................................................................................................. Protected Area PAG ........................................................................................ Protective Action Guideline PRA/PSA ...................... Probabilistic Risk Assessment / Probabilistic Safety Assessment PWR ....................................................................................... Pressurized Water Reactor PS ........................................................................................................ Protection System PSIG ................................................................................ Pounds per Square Inch Gauge R .........................................................................................................................Roentgen RB ........................................................................................................... Reactor Building RCC ........................................................................................... Reactor Control Console RCIC ................................................................................ Reactor Core Isolation Cooling 1 NUMARC was a predecessor organization of the Nuclear Energy Institute (NEI). to NRC-15-0061 Page 27 Fermi 2 Emergency Action Level Technical Bases RCS ............................................................................................ Reactor Coolant System Rem, rem, REM ....................................................................... Roentgen Equivalent Man RETS ......................................................... Radiological Effluent Technical Specifications RPS ......................................................................................... Reactor Protection System RPV ............................................................................................Reactor Pressure Vessel RSVR ................................................................................................................. Reservoir RW .................................................................................................................... Radwaste RWCU .......................................................................................... Reactor Water Cleanup SAR ............................................................................................... Safety Analysis Report SAS ......................................................................................... Safety Automation System SBO ......................................................................................................... Station Blackout SCBA ...................................................................... Self-Contained Breathing Apparatus SGTS ............................................................................ Standby Gas Treatment System SPDS ........................................................................... Safety Parameter Display System SRO ........................................................................................... Senior Reactor Operator TAF ...................................................................................................... Top of Active Fuel TB ............................................................................................................ Turbine Building TEDE ............................................................................... Total Effective Dose Equivalent TSC ...........................................................................................Technical Support Center UFSAR ................................................................... Updated Final Safety Analysis Report to NRC-15-0061 Page 28 Fermi 2 Emergency Action Level Technical Bases 6.0 FERMI-TO-NEI 99-01 Rev. 6 EAL CROSS-REFERENCE This cross-reference is provided to facilitate association and location of a Fermi 2 EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the Fermi 2 EALs based on the NEI guidance can be found in the EAL Comparison Matrix. Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL RU1.1 AU1 1, 2 RU1.2 AU1 3 RU2.1 AU2 1 RA1.1 AA1 1 RA1.2 AA1 2 RA1.3 AA1 3 RA1.4 AA1 4 RA2.1 AA2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1 RA3.2 AA3 2 RS1.1 AS1 1 RS1.2 AS1 2 RS1.3 AS1 3 RS2.1 AS2 1 RG1.1 AG1 1 RG1.2 AG1 2 to NRC-15-0061 Page 29 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL RG1.3 AG1 3 RG2.1 AG2 1 CU1.1 CU1 1 CU1.2 CU1 2 CU2.1 CU2 1 CU3.1 CU3 1 CU3.2 CU3 2 CU4.1 CU4 1 CU5.1 CU5 1, 2, 3 CA1.1 CA1 1 CA1.2 CA1 2 CA2.1 CA2 1 CA3.1 CA3 1 CA3.2 CA3 2 CA6.1 CA6 1 CS1.1 CS1 1 CS1.2 CS1 2 CS1.3 CS1 3 CG1.1 CG1 1 CG1.2 CG1 2 FA1.1 FA1 1 FS1.1 FS1 1 FG1.1 FG1 1 HU1.1 HU1 1, 2 3 to NRC-15-0061 Page 30 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL HU2.1 HU2 1 HU3.1 HU3 1 HU3.2 HU3 2 HU3.3 HU3 3 HU3.4 HU3 4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3 HU4.4 HU4 4 HU7.1 HU7 1 HA1.1 HA1 1, 2 HA5.1 HA5 1 HA6.1 HA6 1 HA7.1 HA7 1 HS1.1 HS1 1 HS6.1 HS6 1 HS7.1 HS7 1 HG1.1 HG1 1 HG7.1 HG7 1 SU1.1 SU1 1 SU3.1 SU2 1 SU4.1 SU3 1 SU4.2 SU3 2 SU5.1 SU4 1, 2, 3 to NRC-15-0061 Page 31 Fermi 2 Emergency Action Level Technical Bases Fermi 2 NEI 99-01 Rev. 6 Example EAL IC EAL SU6.1 SU5 1 SU6.2 SU5 2 SU7.1 SU6 1, 2, 3 SA1.1 SA1 1 SA3.1 SA2 1 SA6.1 SA5 1 SA8.1 SA9 1 SS1.1 SS1 1 SS2.1 SS8 1 SS6.1 SS5 1 SG1.1 SG1 1 SG1.2 SG8 1 EU1.1 E-HU1 1 to NRC-15-0061 Page 32 Fermi 2 Emergency Action Level Technical Bases 7.0 ATTACHMENTS 7.1 Attachment 1, Emergency Action Level Technical Bases 7.2 Attachment 2, Fission Product Barrier Matrix and Basis to NRC-15-0061 Page 33 Fermi 2 Emergency Action Level Technical Bases ATTACHMENT 1 EMERGENCY ACTION LEVEL TECHNICAL BASES to NRC-15-0061 Page 34 Fermi 2 Emergency Action Level Technical Bases Category R - Abnormal Rad Levels / Rad Effluent EAL Group: ALL (EALs in this category are applicable to any plant condition, hot or cold.) Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification. At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety. Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification.

to NRC-15-0061 Page 35 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer. EAL: RU1.1 Unusual Event Reading on any Table R-1 effluent radiation monitor > column "UE" for 60 min. (Notes 1, 2, 3) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 36 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific Liquid Releases Fermi does not perform continuous radioactive liquid releases and no longer performs periodic batch radioactive liquid releases, per administrative controls. However, to provide EALs consistent with the template scheme, a liquid effluent EAL threshold has been developed. (Ref. 2) Per ODCM Figure 6.0-1, all sources of liquid effluent converge at a common discharge point prior to reaching the environment (Ref. 1). The D11-K604 Radiation Monitor on the liquid radwaste effluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 MPC at the discharge to Lake Erie. The monitor is located upstream of the Isolation Valve (G11-F733) on the liquid radwaste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir decant flow (Ref. 2). The Circulating Water Reservoir (CWR) Decant Line Radiation Monitor (D11-N402) and recorder (D11-R806) provides indication of the concentration of radioactive material in diluted radioactive liquid releases just before discharge to Lake Erie; and thus being the final monitor in the liquid discharge line is the liquid monitor used to address this EAL threshold. The value shown in Table R-1 column UE represents 2 times the ODCM limit of 1 MPC (Ref. 2). Gaseous Releases The column UE gaseous release values in Table R-1 represent 2 times the calculated release values associated with the ODCM limit for total body (the skin limit requires a higher release rate value) plus background. (Ref. 1, 2). For this initiating condition, the applicable effluent monitors are RB SPING, SGTS I SPING, SGTS II SPING, RW SPING and TB SPING (Ref. 2). to NRC-15-0061 Page 37 Fermi 2 Emergency Action Level Technical Bases Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. Escalation of the emergency classification level would be via IC RA1. Fermi Basis Reference(s):

1. Fermi Offsite Dose Calculation Manual

2. EP-EALCALC-FERMI-1401 Radiological Effluent EAL Values Rev. 0

3. NEI 99-01 AU1 to NRC-15-0061 Page 38 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Levels / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.

EAL: RU1.2 Unusual Event Sample analysis for a gaseous or liquid release indicates a concentration or release rate 2 x ODCM limits for 60 min. (Notes 1, 2) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Mode Applicability: All Definition(s): None Basis: Site Specific None Generic This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional to NRC-15-0061 Page 39 Fermi 2 Emergency Action Level Technical Bases releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RA1. Fermi Basis Reference(s):

1. Fermi Offsite Dose Calculation Manual

2. NEI 99-01 AU1 to NRC-15-0061 Page 40 Fermi 2 Emergency Action Level Technical Bases Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE EAL:

RA1.1 Alert In the absence of real-time dose assessment, reading on any Table R-1 effluent radiation monitor > column "ALERT" for 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded. Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit. Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes. Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. Table R-1 Effluent Monitor Classification Thresholds Release Point Monitor GE SAE Alert UE RB Ventilation SPING (Ch. 5) N/A N/A N/A 3.3E-3 µCi/cc SPING (Ch. 7) N/A N/A N/A 4.1E-2 µCi/cc SGTS Div. I AXM (Ch. 3) 8.0E+2 µCi/cc 8.0E+1 µCi/cc 8.0E+0 µCi/cc N/A Gaseous SPING (Ch. 7) N/A N/A N/A 4.0E-2 µCi/cc SGTS Div. II AXM (Ch. 3) 7.6E+2 µCi/cc 7.6E+1 µCi/cc 7.6E+0 µCi/cc N/A RW Ventilation SPING (Ch. 5) N/A N/A N/A 1.5E-2 µCi/cc TB Ventilation SPING (Ch. 5) N/A N/A N/A 2.0E-4 µCi/cc Liquid CW RSVR D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line Mode Applicability: All Definition(s): None to NRC-15-0061 Page 41 Fermi 2 Emergency Action Level Technical Bases Basis: Plant-Specific Liquid Releases The RA1 IC addresses a release of radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Per ODCM Figure 6.0-1, all sources of liquid effluent converge at a common discharge point prior to reaching the environment (Ref. 1). The D11-K604 Radiation Monitor on the liquid radwaste effluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 MPC at the discharge to Lake Erie. The monitor is located upstream of the Isolation Valve (G11-F733) on the liquid radwaste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir decant flow (Ref. 2). The Circulating Water Reservoir (CWR) Decant Line Radiation Monitor (D11-N402) and recorder (D11-R806) provides indication of the concentration of radioactive material in diluted radioactive liquid releases just before discharge to Lake Erie; and thus being the final monitor in the liquid discharge line is the liquid monitor used to address this EAL threshold (Ref. 1, 2). The value shown in Table R-1 column Alert for liquid releases represents 10 mRem for one hour of exposure (Ref. 2). Gaseous Releases For gaseous releases, the preferred method for classification is by means of the computerized dose assessment program incorporating actual meteorology. This method is preferred since it eliminates uncertainty associated with assumed meteorology and source term data. For this initiating condition, the applicable gaseous effluent monitors are the Division I and II AXMs. to NRC-15-0061 Page 42 Fermi 2 Emergency Action Level Technical Bases The column ALERT gaseous release values in Table R-1 represent offsite dose resulting from an actual or imminent release of gaseous radioactivity that exceeds 10 mRem TEDE or 50 mRem thyroid CDE (1% of the EPA PAGs) (Ref. 2). Generic This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1. Fermi Basis Reference(s):