NL-16-1830, Southern Nuclear Operating Company Response to RAI for License Amendment Regarding Emergency Action Level Schemes to Adopt NEI 99-01 Rev 6 and to Modify Radiation Monitors at Farley Nuclear Plant, Part 1C of 1C: Difference between revisions

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{{#Wiki_filter:Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant Units 1 and 2; Edwin I. Hatch Nuclear Plant Units 1 and 2; Vogtle Electric Generating Plant Units 1 and 2; License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Responses to Requests for Additional Information ENCLOSURE 3 EAL SCHEMES MARKED-UP PAGES
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Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Responses to Requests for Additional Information JOSEPH M. FARLEY NUCLEAR PLANT EAL SCHEME MARKED-UP PAGES
 
FARLEY NUCLEAR PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND .................................................................................. 1 1.1  OPERATING REACTORS .................................................................................................. l 1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ..**.******.***.***********.********. l 1.3  NRC ORDER EA-12-051 ................................................................................................2 1.4  ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION ..***************.**...****.***3 1.5  IC AND EAL MODE APPLICABILITY ..............................................................................3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ....................................... 5 2.1  GENERAL CONSIDERATIONS ********.************************.***..******.******..****...**.**********..************.* 5 2.2  CLASSIFICATION METHODOLOGY ................................................................................. 6 2.3  CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .****..*****.**..*.***..*.*****.********** 6 2.4  CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION *************.****.*****.******. 6 2.5  CLASSIFICATION OF IMMINENT CONDITIONS ............................................................... 7 2.6  EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING *****.************* 7
: 2. 7 CLASSIFICATION OF SHORT-LIVED EVENTS ................................................................. 7 2.8  CLASSIFICATION OF TRANSIENT CONDITIONS ***..***.....***************************.*********************7 2.9  AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION **************** 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS .*..*..*.......*....***.. 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS ***.***.**...**...* 26 5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .*.*****..**. 46 6  FISSION PRODUCT BARRIER ICS/EALS ............................................................... 49 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ........ 65 8  SYSTEM MALFUNCTION ICS/EALS ........................................................................ 88 APPENDIX A - ACRONYMS AND ABBREVIATIONS...................................................... A*1 APPENDIX B - DEFINITIONS ........................................................................................ B*1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title IO, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections for this document are:
* IO CFR § 50.47(a)(l)(i)
* IO CFR § 50.47(b)(4)
* 10 CFR § 50.54(q)
* IO CFR § 50.72(a)
* 10 CFR § 50, Appendix E, IV.B, Assessment Actions
* IO CFR § 50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by regulatory guidance documents. Documents of particular relevance to NEI 99-01 include:
NUREG-0654/FEMA-REP-l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines JO CFR § 50.72 and§ 50. 73 Regulatory Guide l.IOl, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions also may be directed to the NEI Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NE! 99-01 is applicable to licensees electing to use their IO CFR 50 emergency plan to fulfill the requirements of I 0 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of I 0 CFR § 50 and the guidance in NUREG 0654/FEMA-REP- l. The initiating conditions germane to a I 0 CFR § 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a IO CFR § 50.47 emergency plan.
The generic !Cs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HUI covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. . Additionally, appropriate aspects ofIC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI.
 
The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulato1y Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to ModifY Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all U.S. nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (I) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
* A display in an area accessible following a severe event; and
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To ModifY Licenses with Regard to Reliable Spent Fuel Pool Instrumentation'', provides guidance for complying with NRC Order EA-12-051.
NE! 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-05 I. These 2
 
EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the effectiveness of the plan. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with I 0 CFR 50.90.
1.4 0RGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The
    !Cs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode                R        c      E        F        H        s Power Operations          x                x        x        x        x Startup              x                x        x        x        x Hot Standby              x                x        x        x        x Hot Shutdown              x                x      x        x        x Cold Shutdown            x        x      x                x Refueling              x        x      x                x Defueled              x        x      x                x 3
 
Farley Units 1 and 2 Technical Specifications Table 1.1-1 provides the following operating mode definitions:
Reactivity      % Rated Average RCS Mode            Title          Condition      Thermal PowerCa)    Temperature (°F)
(Kerr) 1    Power Operation          :::: 0.99        >5                NA 2      Startup                  2':0.99          :S 5              NA 3    Hot Standby              < 0.99          NA            :::: 350 4      Hot ShutdownCb)          <0.99            NA        350 > Tavg > 200 5    Cold ShutdownCb)          < 0.99          NA            :'.:: 200 6    RefuelingCc)                  NA          NA                NA (a)  Excluding decay heat.
(b)  All reactor vessel head closure bolts fully tensioned.
(c)  One or more reactor vessel head closure bolts less than fully tensioned In addition to these defined modes, "Defueled" is also applicable to the Farley EAL scheme, consistent with NE! 99-01. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Farley EALs with no modifications from NE! 99-01.
When a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
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2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes and the informing basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.
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 provides guidance on implementing this requirement in NSIRIDPR-ISG-01, Interim Staff Guidance, Emergency Planning/or Nuclear Power Plants.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
Indication will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 will be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting in an expected event or condition that 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with 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 &sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
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Although the EALs have been developed to address a full spectrum of possible events and conditions that may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-0 I scheme provides the emergency director with the ability to classify events and conditions based upon judgment using EALs consistent with the emergency class ification level (ECL) definitions (refer to Category H). The emergency director will determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A simi lar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fi ss ion product barrier.
2.2 CLASS IFICATION 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 ifthe EAL has been met or exceeded. An EAL(s) evaluation must be consistent with the related operating mode applicability and notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is dec lared in accordance with plant procedures.
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-0 I.
2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS In the event of mult iple emergencies or conditions, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this rev iew is declared.
For example:
If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared.
There is no " additive" effect from multiple EALs meeting the same ECL. For example:
If two Alert EA Ls are met, an A lert will be declared.
Related guidance for classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency No tifications During Quickly Changing Events.
2.4 CONSIDERATION OF MODE CHANGES D URING CLASS IFICATION The mode in effect at the time an event or cond ition occurred, and prior to any plant or operator response, determines whether 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 diflercnt mode is reached, ai1y ne\\ event or condition, not rdated to the original event or condition. re4uiring emergency classification should be evaluated against the !Cs and I ALs applicable lo lhc operating mode at the time    or the new event or condition.                                ( Commented [JRBl] : RAJ I revision 6
 
2.5  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 IMMIN ENT, the emergency classification will be made as though the EAL has been met. While applicab le to all emergency classification levels, this approach is particularly important at the higher emergency classification levels si nce it provides additional time for implementation of protective measures.
2.6    EMERGENCY CLASSIFICATION LEVEL UPGRAD ING AND DOWNGRADING SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification. Termination criteria contained in proced ure NMP-EP- 110, Emergency Classification and Initial Actions shall be completed for an event to be terminated. At termination, on an event specific basis, the site will enter either normal operating conditions or a recovery condition with a recovery organization established for turnover from the ERO.
2.7  CLASSIFICATION OF SHORT-L IVED 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 end before the emergency classification assessment can be completed,-. If an event occurs that meets or .:xcccds an I Al
* the assodatcd I Cl must l-x: tkclared rcgru*dlt:ss of its continu.:d prcsenc.: at the time of declaration. For exam ple, an earthquake, or failure  ( Commented [JRB2] : RAJ 2 rev1SOon of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
2.8  CLASS IFICATION OF TRANSIENT CONDITIONS Many of the ICs and EALs in this document employ time-based criteria that require IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. ln cases where no time-based criterion is specified, some transient conditions may cause an EAL to be met for a brief period of time. The following guidance will be applied to the classification of these conditions.
EAL momentarily met during expected plant response - When 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 applicable EAL is not considered met and the associated emergency declaration is not required. This example presents an illustration:
An ATWS occurs and the auxil iary feedwater system fails to automatically start.
Steam generator levels rapidly decrease and the plant enters an inadequate RCS 7
 
heat remova l condition (a potential loss of both the fuel clad and RCS barrier ). If an operator manually starts the auxi li ary feedwater system in accordance w ith an EO P step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification will be based on the ATWS onl y.
It is important to note that the 15-minute emergency classification assessment period is not a "grace period" to delay a classification in order to perform a corrective action that wo uld obviate the need to class ify the event. Emergency class ification assessments must be deliberate and timely, w ith no undue delays. The provision discussed above addresses onl y rapi dl y evolving situations in which an operator is able to take corrective action before the emergency director completes the rev iew and necessary steps to make the eme rgency declaration. This provision ensures any public protective actio ns resulting from the emergency classification are truly warranted by the plant conditions.
2.9 AITER-THE-FACT DISCOVE RY OF Ai EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency class ification was not made at the time of the event or condition. Personnel could discover that an event or co ndition ex isted that met an EAL, but no emergency was declared, and the event or condition no longer ex ists at the tim e of discovery. It may be the event or condition was not recognized at the time, or there was an error in the emergency classification process.
In these cases, no emergency declaration is warranted; but, the guidance in NUREG-1022 is applicable. Specifically, the event shou ld be reported to the NRC in accordance w ith I 0 CFR &sect; 50. 72 w ithin o ne hour of the undeclared event or condition is discovered. T he licensee will also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
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3    ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/ EALS GENERAL                SITE AREA ALERT              NUSUAL EVENT EMERGENCY                EMERGENCY RG I Release of          RSI Release of          RA J Release of        RUI Release of gaseous radioactivity    gaseous radioactivity  gaseous or liquid      gaseous or liquid resulting in offsite    resulting in offsite    radioactivity resulting radioactivity greater dose greater than 1,000 dose greater than I 00  in offsite dose greater than 2 times the mrem TEDE or 5,000      mrem TEDE or 500        than 10 mrem TEDE      ODCM lim its for 60 mrem thyroid CDE.        mrem thyroid CDE .      or 50 mrem thyroid      minutes or longer.
Op. Modes: All          Op. Modes: All          CDE.                    Op. Modes: All Op. Modes: All RG2 Spent fuel pool      RS2 Spent fuel pool    RA2 Significant        RU2 UNPLANNED level cannot be          level al HG- 111 feet  lowering of water level loss of water level restored to at least HG  (Le,*el 3).            above, or damage to,    above irradiated fuel. Commented [JRB4] : RAJ J.d rev1S1on
                                                                                                                    ~~~~~~~~~--
111 feet (Le,*el 3) for Op. Modes: All          irradiated fuel.        Op. Modes: All        Commented [JRBJ] : RAJ 3 d rev1S1on 60 minutes or longer.                            Op. Modes: All Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
RG1 ECL: General Emergency Initiating Condition: Release of gaseo us radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director wi ll declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will like ly be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* lfthe 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.
* The pre-calculated effluent monitor value presented in EAL# I will be used for emergency c lassification assessments until the results from a dose assessment using actual meteorology are avai lable.
( 1)      Readings on ANY of the following radiation monitors greate r than the reading shown below for 15 minutes or longer:
Steam Jet Air Ejector RE-I SC              130 &#xb5;Ci/cc ( 130 R/hr)
Plant Vent Stack RE-298 (NG )              0.8 &#xb5;C i/cc (2)      Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE at or beyond the site boundary.
(3)      Field survey results indicate EITHER of the following at o r beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid COE greater than 5,000 mrem for one hour of inhalation.
Basis:
Th is IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protecti ve Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.
Radiolog ical effluent EALs are included to provide a bas is for classifying events and conditions that cannot be readi ly or appropriate ly 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified 10
 
in the IC. The meteorology and source term (noble gases, particulates, and haloge ns) used are the same as those used to determine the monitor reading threshold values in !Cs RS I and RA I. This protocol will maintain intervals between the threshold values for the three classifications. Since doses are genera lly not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of IOOO mRJhour whole body or 5000 mR/hour thyro id, whi chever is more limiting.
The TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in considerati on of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.
C lass ification based on effl uent monitor readings ass um es that a release path to the environment is established. If the effl uent flo w past an effluent monitor is known to have stopped due to actions to iso late the release path, then the effl uent monitor reading is no longer valid for classification purposes.
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RG2 ECL: General Emerge ncy Initiating Condition: Spent fuel pool level cannot be resto red to at least +JQ- 1 <I feet (Le,*el 3) [ Commented [lRBS]: RAl 3 d rev1s1on for 60 minutes or longe r.
Operating Mode App licability: A ll Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
( I)    Spent fuel pool level cannot be restored to at least +JQ- 131 feet (Le\*el 3) for 60 minutes ( Commented [JRB6] : RAJ 3.d rev1s1on or longer.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. The spent fuel level instrument is located outside the Control Room but in close proximity. This condition will lead to fuel damage and a radiological release to the environment.
It is recogni zed that thi s IC wou ld li kely not be met until well after another General Emergency IC was met; however, it is included to provide classification divers ity.
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RS1 ECL: Site Area Emerge ncy Initiating Co ndition: Re lease of gaseous radioactivity resul ti ng in offsite dose greater than I 00 mrem TE DE or 500 mrem thyro id C OE.
Operating Mode Applicabili ty: A ll Emergency Action Levels : ( I o r 2 o r 3)
Notes :
* The emergency director will declare the ite Area Emergency promptly upo n dete rmining that the applicable time has been exceeded, or w ill like ly be exceeded.
* If an ongo ing release is detected and the release start ti me is unknown, ass um e that the release duratio n has exceeded 15 minutes.
* If the effl uent flow past an efflue nt mo nitor is known to have stopped due to actions to isolate the release path, then the e ffluent monitor reading is no lo nger valid fo r class ification purposes.
* The pre-calculated efflue nt mo nitor values presented in EAL# I w ill be used fo r emergency class ification assessments until the res ults from a dose assessment using actual meteorology are available.
( I)    Reading o n ANY of the foll ow ing radi ation monitors g reater than the reading shown fo r 15 minutes or lo nger:
Steam Jet Air Ejector RE- 1SC                13 &#xb5;Ci/cc ( 13 R/hr)
Plant Vent Stack RE-298 (NG )                0.08 &#xb5;C i/cc (2)      Dose assessment using actual meteoro logy indicates doses greater than 100 mre m TE DE or 500 mrem thyroid COE at or beyo nd the site boundary.
(3)      Field survey results indicate EITHER of the fo ll ow ing at or beyond the site boundary:
* Closed w indow dose rates greater than 100 mR/hr expected to contin ue fo r 60 minutes or longer.
* Analyses o f fi eld survey sampl es ind icate thyroid COE greater than 500 mrem fo r o ne hour of inhalation.
Basis:
This IC addresses a release o f gaseo us radi oactivity that res ult in projected or actual offsite doses greater than or equal to I 0 perce nt of the EPA Protective Action Guides (PA Gs). It includes both monitored and un-m onitored releases. Releases o f thi s magnitude are associated with the fa ilure o f plant systems needed fo r the protection of the public.
Rad iological effl uent EALs are included to prov ide a bas is fo r class ifying events and condi tions that cannot be readi ly or appropriate ly class ified on the bas is of plant conditions alone. T he inclusion of both plant condi tion and radiological effluent EALs more fully addresses the spectrum of possible accident eve nts and co nd itions. The mo ni tor reading thres ho ld val ues are 13
 
determined usi ng a dose assessment method that back calculates fro m the dose va lue pecified in the IC. The meteorol ogy and source term (noble gases, particul ates, and haloge ns) used is the same as those used to determine the monitor reading threshold values in ICs RG I and RA I. This protocol mai ntain intervals between the threshold values fo r the th ree class ifications. Since doses are generally not monitored in real-time, a release duration of one hour is assum ed, an d the threshold values are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour thyroid, whichever is more lim iting.
The TEOE dose is set at IO percent of the EPA PAG of 1,000 mrem while the 500 mrem thyro id COE was established in consideration of the I :5 ratio of the EPA PAG for TEOE and thyro id COE.
Class ification based on e ffluent monitor readings assum es that a release path to the enviro nment is established. If the effluent fl ow past an effluent monitor is known to have stopped due to act ions to iso late the re lease path, then the effl uent monitor reading is no longer valid fo r class ificat ion purposes.
Escalat ion of the emerge ncy classi fi cati on level uses IC RG I.
14
 
RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level    at ~ 111  feet (Level 3).                          ( Commented [JRB7]: RAJ 3 d revision Operating Mode Applicability: All Emergency Action Levels:
( I)    Lowering of spent fuel pool level  to ~ I 'I  feet (Le\*el 3).                            [ Commented [JRB8]: RAJ 3 d revmon Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capabi li ty leadi ng to IMMINENT fuel damage. This conditi on stems from major fai lures of plant functions needed to protect the public that warrant a Site Area Emergency declaration. The spent fuel pool level instrument is located outside the Control Room but in close proximity.
It is recognized that this IC would likely not be met until we ll after another Site Area Emergency IC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level uses via IC RGI or RG2.
15
 
RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3 or 4)
Notes:
* The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL# I will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(I)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Steam Jet Air Ejector RE-15C                1.3 &#xb5;Ci/cc (1.3 R/hr)
Plant Vent Stack RE-29B (NG)                0.008 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary.
(3)    Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary for one hour of exposure.
(4)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to I percent 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 plant safety as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
16
 
Radiological effluent EALs are 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in !Cs RG 1 and RS 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at 1 percent 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 uses IC RS I.
17
 
RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
(1)    Uncovery of irradiated fuel in the REFUELING PATHWAY.
(2)    Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on ANY of the following radiation monitors:
Spent Fuel Pool Ventilation Monitor RE-25A ORB Spent Fuel Pool Area Radiation Monitor RE-5 Containment Purge Ventilation Monitor RE-24A ORB (3)    Lowering of spent fuel pool level to 140 feet (Level 2).
Basis:
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly. 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 plant safety.
This IC 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 ICE-HUI.
Escalation of the emergency is based on either Recognition Category R or C ICs.
This EAL escalates from RU2. 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), 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 will 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 18
 
indication of whether the fuel is actually uncovered. To the degree possible, readings will be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.
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 will be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Containment Purge Ventilation Monitors are not available during all modes.
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. The spent fuel pool level instrument is located outside the Control Room but in close proximity. This condition reflects a significant loss of spent fuel pool water inventory and is 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 uses !Cs RS 1 or RS2.
19
 
RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
Note: 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.
(1)    Dose rate greater than I5 mR/hr on RE-IA, Control Room Radiation Monitor.
(2)      An UNPLANNED event results in radiation levels that prohibit or impede access to any Tabl_e HI plant rooms or areas:
Table HI Mode      Room Name                          Room Number Electrical Penetration Room        334, 333, 347 I 2334,2333,2347 Hallway Outside Filter Room        312, 332/
3 IA I 2A MCC areas                  2312,2332 Sample Room and Primary CHM labs    323, 324 I 2323,2324 Sample Room and Primary CHM labs    323, 324 I 2323,2324 4
RHRHxRoom.                        128/
2128 Basis:
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses elevated radiation levels in certain plant rooms or 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 plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL#!, the area requiring continuous occupancy is the control room and the central alarm station. The central alarm station is in the control room envelope. The value of 15mR/hr is derived from the GDC I 9 value of 5 Rem in 30 days with adjustment for expected occupancy times.
For EAL #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 20
 
necessary at the time of the increased radiation levels. Access will 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.
* 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 to 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.
Escalation of the emergency classification level uses Recognition Category R, C or F ICs.
21
 
RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
(I)    Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
Liquid Effluents Steam Generator Slowdown Effluent Line RE-238        2.80 x 10 3 cpm Gaseous Effluents Steam Jet Air Ejector RE-15                          3.5 x 10 2 cpm Plant Vent Gas R-14                                        3.2 x IO~cpm RE-22                                      4.0 x 102 cpm RE-298 (NG)                                8.9 x Io-~ &#xb5;Ci/cc (2)    Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
Li uid Radwaste Effluent Line RE-18        2 x release ermit set oint ( lanned release)
Plant Vent Gas R-14                        2 x release permit setpoint (planned release)
(3)    Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
This IC addresses a potential decrease in the level of plant safety 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.
22
 
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment indicates degradation in these features and/or controls.
Radiological effluent EALs are 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 will not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
EAL# 1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EAL #3 - 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 uses IC RA!.
23
 
RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels:
(I)      a.      UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Personnel report of low water level Annunciator EH2 "SFP LVL HI/LO" AND
: b.      UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
RE-5 in the spent fuel pool building RE-2 in containment RE-27A ORB in containment Basis:
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition can be a precursor to a more serious event and indicates a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of plant safety.
A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications 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 will be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor 24
 
vessel head or movement ofa 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 with Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level uses IC RA2.
25
 
4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL                SITE AREA ALERT          UNUSUAL EVENT EMERGENCY              EMERGENCY CGI Loss ofRPV          CSI Loss ofRPV        CAI Loss ofRPV        CUI UNPLANNED inventory affecting      inventory affecting    inventory.            loss ofRPV inventory fuel clad integrity with core decay heat        Op. Modes: Cold        for 15 minutes or containment              removal capability. Shutdown, Refueling    longer.
challenged.              Op. Modes: Cold                              Op. Modes: Cold Op. Modes: Cold          Shutdown, Refaeling                          Shutdown, Refueling Shutdown, Refueling CA2 Loss of all        CU2 Loss of all but offsite and all onsite one AC power source AC power to            to emergency buses for emergency buses for    15 minutes or longer.
15 minutes or longer. Op. Modes: Cold Op. Modes: Cold        Shutdown, Refueling, Shutdown, Refueling,  Defiteled Defiteled CA3 Inability to      CU3 UNPLANNED maintain the plant in  rise in RCS cold shutdown.        temperature.
Op. Modes: Cold        Op. Modes: Cold Shutdown, Refi1eling  Shutdown, Refiteling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling cus    Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refueling, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Cold Shutdown, Refueling 26
 
CG1 ECL: Gene ral Eme rgency Initia ting Co ndition : Loss of RPV inventory affecting fuel clad integrity with co ntainment challenged.
Oper a ting Mo de A pplica bi lity: Cold Shutdown, Refueling E me rge ncy Ac tio n Levels:      I                                                          [ Commented [JRB9] : RAJ I 0 revision Note: The emergency directo r will declare the Gene ral Emergency promptly upon determ ining that 30 minutes has bee n exceeded, or will likely be exceeded.
(I)      a.        RPV level  lc~s than Al\Y ofthc folio\\ ing for 30 minutes or longer:
* 0% RVI I'-> (\'1odl' 5l
* 119* ll'mpormy Lc1l'l Indicator (Mode 6)
                  \"I>
h.
Reactor vesse l level cannot be monitored for 30 minutes or longer.          [ Commented [JRBlO]: RAJ I0 rev1S1on AN D
: b.      Co re uncovery is indicated by ANY of the following:
* Containment High Range Radi ation Monitor RE27A or 278 read ing greater than or equal to 100 R/hr.
* Erratic so urce range monito r indi cation
* UNPLANNED rise in Containment Sump, or Reactor Coolant Drain Tank (RCDT), or Waste Ho ldup Tank (WHT) leve ls of sufficient magnitude to indicate core uncovery AN D
: c.      ANY indication from the Containment Challenge Table C I.
Co ntainm ent C hallenge Table C l
* CONTAINMENT CLOSURE not established*
* Greater than or equal to 6 % Hz ex ists inside containment
            ** lfCONTAlNMENT CLOSURE is re-estab lis hed prior to exceeding the 30-minute UNPLANNED increase in containment pressure time limit, then declaration of a General Emergency is not required.
Bas is:
CO TAINMENT CLOSURE : Per FNP- 1(2)-STP- 18.4, " Containment Integrity Verification and C losure''.
27
 
UN PLA        ED: A parameter change or an event that is not I) the result of an intended evo lution or 2) an expected pl ant respo nse to a trans ient. T he cause of the parameter change or eve nt may be known o r unknown.
T hi s IC addresses the inabili ty to resto re and ma intain reacto r vesse l leve l above the top of acti ve fu e l with co nta inm ent challenged. This condition re prese nts actua l or IMMIN ENT substantia l core degrad ation or me lting with potential for loss of containment integrity. Re leases can be reasonabl y expected to ex ceed EPA PA G exposure levels offsite for more than the immediate s ite area.
Fo ll ow ing an extended loss of core decay heat removal and invento ry makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vesse l leve l. If RPV leve l cannot be restored, fue l dam age is probable.
With CONTAINM ENT C LOS URE not established, the re is a hi gh potential fo r a direct and unm onitored re lease o f radioactiv ity to the environment. If CONTAINM ENT CLOSUR E is re-established prior to exceeding the 30-minute time limit, the n declaration of a General Emergency is not required.
The exi stence of an ex plosive mixture means, at a minimum, that the containm ent atm os pheri c hydrogen concentrati on is suffi c ient to suppo rt a hydrogen burn (i.e., at the lower de fl agration limit). A hydrogen burn will raise containment pressure and could res ul t in co ll ateral equ ipment dam age leading to a loss o f conta inm ent integri ty. It the refore represents a cha lle nge to Conta inment integri ty.
In the early stages of a co re uncove ry event, it is unlike ly that hydrogen buildup due to a core uncovery could result in a n explos ive gas mixture in containment. If all installed hydrogen gas mo nitors are o ut-of-service during an event leading to fue l cladding damage, it may not be poss ible to obta in a containm ent hydrogen gas concentratio n reading as ambient co nditio ns within the co ntainme nt w ill prec lude perso nnel access. During periods when install ed conta inm ent hydroge n gas monitors are out-o f- service, operators may use the other listed indications to assess whether co nta inment is chall enged.
In EAL +2.b, the 30-minute crite rio n is tied to a readil y recognizable event start time (i.e., the        I Commented [lRBll] : RAJ 10 rev151on tota l loss o f ability to mo nitor leve l), and all ows suffic ient time to monitor, assess and correlate reactor and plant cond iti ons to dete rmine if core uncovery has actually occurred (i.e., to account for various accide nt prog ress ion and instrum entation un certainties). It also allows suffi c ie nt tim e fo r actio ns to terminate leakage, recover in ventory co ntrol or makeup equipm ent, and/or restore leve l monito ring.
The inability to moni tor RPV leve l may be caused by instrume ntation and/or power failures, or wate r leve l dropping be low the range of available instrumentation. If water leve l cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump an d/or tank leve ls. Sump and/or tank leve l changes must be eva luated against other pote nti a l sources o f wate r flo w to ensure they indicate leakage from the RPV .
These EA Ls address co ncern s raised by Generi c Letter 88-1 7, Loss of Decay Heat Removal; SECY 9 1-283, Evaluation ofShutdown and Low Power Risk Issues; NU REG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United tales; and U MARC 9 1-06, Guidelines for Industry Actions to Assess Shutdown Management.
28
 
CS1 ECL: Site A rea Emergency Initiating C ondition: Loss of RPY inventory affecting core decay heat removal capabi lity.
Operating Mode Applicability: Cold Shutdown, Refueling E mergency Action Levels: ( I or 2    *1 ' )                                                    ( Commented [JRB12] : RAJ JOrevision Note: The emergency director wi ll declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will li kely be exceeded.
( I)    a. CO T AINME T CLOSURE not estab lished.
ANO
: b. RVLI S (Mode 5) level less than 16&deg;0 ( 12 1'0" (ti" aelew aettem ID efRCS Jeep). Commented [JRB13]: Ed1tonaJ change* RVLJS reads in %
('.!)    <I. CO    l/d 'MU\ l ( l OSl RI cstablish.:d.
: b. RPV level less than \ \' ol'th..: folio\\ ing*
* 0&deg;o RV I J<; (Mode 5)
* 119* l cmporary I cvcl Indicator (\foul.! 6)                                  [ Commented [JRB14]: RAJ JOrevision f;!1< 3) a. RPY level cannot be monitored for 30 minutes or longer.
ANO
: b. Core uncovery is indi cated by ANY of the fo llowing:
* Containment High Range Rad iation Monitor RE27A or 278 reading greater than or equal to I 00 R/hr
* Erratic source range monitor indication
* UNPLANNED ri se in Containment Sump, or Reactor Coolant Drain Tank (RCDT), or Waste Ho ld up Tank (WHT) levels of sufficient magnitude to indicate core uncovery Basis:
CONTAINMENT CLOSURE: Per FNP-1 (2)-STP-1 8.4, "Contai nment Integrity Verification and Closure" .
UNPLANNE D: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected pla nt response to a transient. The cause of the parameter change or event may be known or unknown.
29
 
This IC addresses a significant and prolonged loss ofRPV inventory contro l and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RC component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entai l major fai lures of plant functions needed to protect the public and warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory make up, decay heat wi ll cause reactor coolant boiling and a further reduction in reactor vesse l leve l. If RPV level cannot be restored, fuel damage is probable.
The level specified in EAL l .b repre ents a level in the RPV that is 6 inches below the bottom ID of the reactor vesse l penetration. This level is lower than the RPV monitoring capability of RCS level instrumentation and therefore must be monitored using RVLIS. This level wi ll only be observable in Mode 5 with RVLI S operable. In Mode 6, when RVLIS is not operable, thi s IC shou ld be evaluated using EAL #2.b or \                                                              { Commented [lRB15]: RAJ 10 rev ision Outage/shutdown contingency plans typicall y provide fo r re-establishing or verifying CO TAINME T CLOSURE following a loss of heat removal or RCS inventory control functions. The specified RCS/reactor vesse l levels of EAL I .b anJ 2.b reflect th at without          ( Commented [JRB16]: RAJ IO revisi on CO T AINMENT CLO URE established, there is a higher probability of a fission product release to the environment.
In EAL ;n a, the 30-minute criterion i tied to a readily recognizable event start time (i .e., the    { Commented [lRB17]: RAJ 10 revisi on total loss of ability to monitor level), and all ows sufficient time to monitor, assess and co rrelate reactor and plant conditions to determine if core uncovery has actually occu rred (i.e., to account fo r various accident progression and instrumentation uncertainties). It also allows sufficient time for actions to terminate leakage, recover inventory control o r makeup equipm ent, 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 occurri ng 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 indicate leakage from the RPV.
These EA Ls address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown and Low Power Risk Issues; NUREG -1 449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMA RC 91 -06, Guidelines for Industry Actions to Assess hutdown Management.
Escalation of the emergency classificat ion level uses IC CG I or RGI.
30
 
CA1 EC L: Alert Initiating Condition: Loss of RPY inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: ( I or 2)
Note: The emerge ncy director will declare the Alert promptly upon determining that 15 minutes has bee n exceeded, or will likely be exceeded.
( I)    Loss of RPV inventory as indicated by level less than 122' 11 ".
(2)      a.      RPV level cannot be monitored fo r 15 minutes or longer AND
: b.      UN PLA        ED increase in Containment sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tank (WH T) levels due to a loss of RPY inventory.
Basis:
UNPLANNED: A parameter change or an event that is not I) the res ult 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.
This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradi ated fuel (i.e., a precursor to a challenge to the fuel clad barri er). This condi tion represents a potential substantial reduction in the level of plant safety.
For EAL # 1, a lowering of water leve l below 122' 11 " indicates that operator actions have not been successful in restoring and maintai ning RPV water level. The 122' 11 " level specified in EAL # I is the minimum RCS level fo r RHR operation pro vided in procedure for mid loop operations. Below this level, loss of RHR pump net positive suction head (N PSH) may occur resulting in a loss of decay heat removal capability. The heat-up rate of the coolant wi ll increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
Although related, EAL # I is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed fo r decay heat removal (e.g., loss of a residual heat removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3 .
For EAL #2, the inability to monitor RPV level may be caused by instrum entation and/o r power fa ilures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determ ine that an inve ntory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated again t other potential sources of water flow to ensure they indicate leakage from the RPY.
31
 
The 15-minute duration for the loss of level indication was chose n because it is hal fof the EAL duration speci fied in IC CS I.
If the RPV inventory level co ntinues to lower, then escalation to Site Area Emergency uses IC cs I.
32
 
CA2 ECL: A lert Initiating Condition: Loss of all offs ite and all onsite AC power to emergency buses fo r 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Re fu eling, De fue led Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( I)    Loss of ALL offsite and ALL onsite AC Power (I able 'i Ii to BOTH 4 I 60Y ESF busses        ( Commented [JRB18]: RAI 12 revision 1(2)F AND 1(2)G for 15 minutes or longer.
Table SI Unit I                                          Unit 2 Start-up Aux XFMR IA                            Start-up Aux XFMR 2A Start-up Aux XFMR I B                          Start-up Aux XFMR 2B Diesel Generator l-2A                          Diesel Generator l-2A Diesel Generator I B                          Diese l Generator 2B Diesel Generator IC                            Diesel Generator IC Diesel Generator 2C                            Diesel Generator 2C Basis:
T hi s IC addresses a total loss of AC power (see Table I above) that compromises the performance of all SAFETY SYSTEMS requi ring electric power including those necessary for emergency core cooling, containment heat removal/press ure control, spent fuel heat removal and the ultimate heat si nk.
When in the co ld shutdown, refueling, or de fueled mode, thi s condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service.
Additional time is available due to the reduced core decay heat load, and the lower temperatures and press ures in various plant systems. When in these modes, thi s condition represents an actual o r potential substantial degradation of the level of plant safety.
Fifteen minutes is the threshold to exclude transient or momentary power losse .
Escalation of the emergency classification level uses IC CS I or RS I.
33
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(I)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F for greater than the duration specified in Table C2.
Table C2: RCS Heat-up Duration Thresholds RCS Status            Containment Closure Status            Heat-up Duration Not Intact                  Not Established                    0 minutes (or at reduced inventory)              Established                    20 minutes*
Intact Not applicable                  60 minutes*
(but not at reduced inventory)
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)    UNPLANNED RCS pressure increase greater than 10 psig. (This EAL does not apply during water-solid plant conditions).
Basis:
CONTAINMENT CLOSURE: Per FNP-1 (2)-STP-18.4, "Containment Integrity Verification and Closure".
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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because I) the evaporated reactor coolant may be released directly into the Containment 34
 
atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, 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 will allow sufficient time to address the temperature increase
. without a substantial degradation in plant safety.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS I or RS I.
35
 
CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
(I)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 36
 
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 plant safety.
The first threshold for EAL 1.b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL 1.b 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 all 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 uses IC CS I or RS I.
37
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss ofRPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1or2)
Note:    The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED rise in Containment sump, Reactor Coolant Drain Tank, or Waste Holdup Tank levels.
Basis:
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL# 1 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 ifthe 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.
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 brieflowering of water level.
EAL #2 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 38
 
sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they indicate leakage from the RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CAI or CA3.
39
 
CU2 ECL: Notifi cation of Unusual Event Init iatin g Co ndition : Loss of all but o ne AC power so urce to eme rge ncy buses fo r 15 min utes or longer.
Oper a ting Mo de Appli cab ili ty: Cold Shutdown, Refueling, Defu eled E me rge ncy Actio n Levels:
Note: The emergency director w ill declare the Unusual Event promptly upon determining that 15 m inutes has bee n exceeded, or will li ke ly be exceeded.
( l)    a.      AC power capabili ty to BOT H 41 60V ESF busses 1(2) F AN D 1(2)G is reduced to a single power so urce nablc ':.. l l fo r 15 minutes o r longer.                  I Commen t ed [JRB19]: RAI 13 b rev1s1on AN D
: b.      A ny addi tional sing le power source fa ilure will result in loss of all AC power to SAFETY SYST EMS.
Tab le S I Unit I                                        Unit 2 Start-up A ux XFMR IA                        Start-up Aux XFMR 2A Start-up Aux XFMR l B                        Start-up Aux XFMR 2B Diesel Gene rator l -2A                      D iesel Generator l-2A Diesel Generator l B                        Diesel Generato r 2B Diesel Generator IC                          Diese l Generato r IC Diesel Generator 2C                          Diese l Generator 2C Bas is:
SA FETY SYSTEM: A system required fo r safe plant operation, cooling down the pla nt and/or plac ing it in the co ld shutdown co ndition, including the ECC . These are typically systems class ified as safety-related.
This IC describes a significant degradation of offsite and onsite AC power sources where any additional single fa ilu re would result in a loss of all AC power to SAFETY SYSTEMS . In thi s condition, the sole AC powe r so urce may be powering one, or more than one, train o f safety-related equipment.
When in the cold shutdown, refue ling, o r defueled mode, this condition is not class ifi ed as an A lert because of the increased tim e ava ilab le to restore another power so urce to service.
Add itional time is avai lable due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. When in these modes, this co ndi tion is co nsidered to be a potential degradation o f the leve l o f plant safety.
An "AC power source" is a source recogni zed in AO Ps and EOPs, and capable of suppl ying required power to an eme rge ncy bus (see Table S l above). Examples o f thi s condi tion include:
40
 
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from an offsite power source.
Fifteen minutes is the 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.
41
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1or2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F.
(2)      Loss of ALL RCS temperature AND RPV level indication for 15 minutes or longer.
Basis:
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.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level. It represents a potential degradation of the level of plant safety. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the emergency director will 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.
EAL# 1 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, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators are 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 is the threshold to exclude transient or momentary losses of indication.
42
 
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
43
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(l)    Indicated voltage is less than 105 VDC on Technical Specification required 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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. This condition is considered to be a potential degradation of the level of plant safety.
As used in this EAL, "required" means the vital DC busses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, ifTrain A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
44
 
CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (I or 2 or 3)
(I)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the followingNRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 theNRC.
This IC will be assessed only when extraordinary means are used 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).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Alabama, Georgia, and Florida; Houston and Henry Counties, Alabama; and Early County, Georgia.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
45
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HUI Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 46
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY of the values listed in Table El.
Table El Location of Dose Rate                  Total Dose Rate (Neutron+ Gamma mR/hr)
HI-TRAC 125 Side-Mid-height                        1360 Top                                260 HI-STORM 100 Side - 60 inches below mid-height                340 Side - Mid- height                        350 Side - 60 inches above mid-height                170 Center of lid                          50 Middle of top lid                        60 Top (outlet) duct                      160 Bottom (inlet) duct                      460 Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY ofa 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 that could cause challenges in removing the cask or fuel from storage.
The existence of"damage" is determined by radiological survey. The radiation reading values listed in the table represent 2 times the site-specific cask specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical specification multiple of"2 times" 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 47
 
recognized that in the case of extreme damage to a loaded cask, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under ICs HUI and HAI.
48
 
6 FISSION PRODUCT BARRIER ICS/EALS                          LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS FUEL CLAD                                      CONTAINMENT Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.                                              Lossofatleast2 Barriers?
                                                                                                                    -  YES    Il.1J. *Luss of A..~Y T1t11 Barri.:n .Mfill.oss or J'ul~'Jl!ial Loss ofThird BJrri.:r FGl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown SITE AREA EMERGENCY                                                                        LOSS    POTENTIAL      - NO--
Loss or Potential Loss of any two barriers. LOSS    PO~~~~IAL                                          LOSS 1--~----1 FUEL CLAD                                      CONTAINMENT FSl    Op. Modes: Power Operation, Hot Standby,              ~---------+------~
Startup, Hot Shutdown ALERT                                                                                              Eil *Loss or Poknrial 1.oss of ANY Two Bllfri=
Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
FAl                                                          POTENTIAL        POTENTIAL LOSS                LOSS Op. Modes: Power Operation, Hot Standby,                LOSS              LOSS FUELClAD              RCS Startup, Hot Shutdown ill* ANY Loss or ANY Pot.:nti.il b*~~ of.filil.!1ill.
FuelC!ad!IB.RCS 49
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY                            FSI SITE AREA EMERGENCY                              FAlALERT Loss of any two barriers and Loss or          Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either Potential Loss of the third barrier.                                                      the Fuel Clad or RCS barrier.
Fuel Clad Barrier                                    RCS Barrier                                Containment Barrier LOSS              POTENTIAL LOSS                  LOSS              POTENTIAL LOSS              LOSS              POTENTIAL LOSS I. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage Not Applicable          A CORE COOLING          A An automatic or          A Operation of a    A A leaking or          Not Applicable CSF-ORANGE            manual ECCS                standby charging    RUPTURED SG is entry conditions        actuation is required      pump is required    FAULTED outside met.                    by EITHER of the          by EITHER of the    of containment.
following:                following:
* UNISOLABLE
* UNI SOLAB LE RCS leakage                  RCS leakage
* SG tube
* SG tube RUPTURE.                    leakage.
OR B. RCS INTEGRITY CSF - RED entry conditions met 50
 
Fuel Clad Barrier                                  RCS Barrier                                  Containment Barrier LOSS              POTENTIAL LOSS                  LOSS            POTENTIAL LOSS                LOSS              POTENTIAL LOSS
: 2. Inadequate Heat Removal                      2. Inadequate Heat Removal                      2. Inadequate Heat Removal A. CORE COOLING A. CORE COOLING                  Not Applicable          A. HEAT SINK CSF -      Not Applicable          A. CORE COOLING CSF - RED entry          CSF-ORANGE                                      RED entry                                      CSF - RED entry conditions met            entry conditions                                conditions met.                                conditions met for 15 met                                                                                              minutes or longer OR                                          NOTE: Heat Sink CSF B. HEAT SINK CSF -                              should not be RED entry                                    considered RED if total conditions met                                AFW flow is less than 395 gpm due to NOTE: Heat Sink CSF                              operator action.
should not be considered RED if total AFW flow is less than 395 gpm due to ooerator action.
: 3. RCS Activity I Containment Radiation          3. RCS Activity I Containment Radiation          3. RCS Activity I Containment Radiation A. Containment            Not Applicable        A. Containment          Not Applicable          Not Applicable          A. Containment radiation monitor                                radiation monitor                                                        radiation monitor RE-27 A or B                                      RE-2 greater than I                                                      RE-27 AorB greater than 600                                  R/Hr                                                                      greater than 8000 R/Hr.                                            OR                                                                        R/Hr.
OR                                                Containment B. Indications that                                  radiation monitor reactor coolant                                  RE-7 greater than activity is greater                              500 mR/Hr.
than 300 &#xb5;Ci/gm dose equivalent 1-131.
51
 
Fuel Clad Barrier                          RCS Barrier                                  Containment Barrier LOSS              POTENT IAL LOSS          LOSS            POT ENT IAL LOSS              LOSS                  POT ENT IAL L O SS
: 4. Co ntainment Integ rity or By pass    4. C ontainment Integrity or By pass        4. Containment Integ rity or Bypass Not Applicable          Not Appl icable  Not Applicable          Not Applicable    A. Containment isolation    A. CONTAINM ENT CSF is required                  RED entry conditions AN D                          met.
EITHE R of the                OR
                                                                                          . following:
Containment integrity has been B. Containment Hydrogen concentration greater than ~ %                [ Commented [JRB21]: RA! 18 revision lost based on            OR Emergency Director
: c. I. GGN+,O.ll'I~ 46N+
G!>i;:QR,,<\l'IG6
                                                                                          .      judgment.
UNISOLABLE pathway from the 60AQll10AS meK prcssL..      I
                                                                                                                                              -r I containment to              ~7  r                [ Commented [JRB22]: RA! 19 revision the environment exists.                    AND OR                            2. Less than one CT MT fan coolers B. Indications of RC                  and one full train of leakage outside of              CTMT Spray is contai nment as                    operating per design indicated by alarms on          for 15 minutes or any of the following                longer.
                                                                                          ..instruments:
RE- 10
                                                                                          .. RE-14 RE-2 1 RE-22 Nott Incrcasc> m ;ump le\ds temperatures.
prc>>urcs, Jlm, rate> und or raJ1auon le\d reading outside of the conttunnh..::llt mav 111d1cate that the KCS ma~s 1s h~1 ng  lost outstd~
ol contai~mcnt                                              ( Commented [JRB20]: RA! t 7 revision
: s. Other Indications                    s. Other Indications                      s. Other Indications Not appli cable          Not applicable  Not appli cable          Not applicable    Not applicable                  Not aoo licable 52
 
Fuel Clad Barrier                                    RCS Barrier                                      Containment Barrier LOSS            POTENTIAL LOSS                  LOSS                POTENTIAL LOSS                    LOSS              POTENTIAL LOSS
: 6. Emergency Director Judgment                  6. Emergency Director Judgment                        6. Emergency Director Judgment A. ANY condition in    A. ANY condition in      A. ANY condition in the    A. ANY condition in        A. ANY condition in      A. ANY condition in the the opinion of the      the opinion of the      opinion of the              the opinion of the        the opinion of the        opinion of the emergency              emergency director      emergency director          emergency director        emergency director        emergency director director that          that indicates          that indicates loss of      that indicates            that indicates loss of    that indicates indicates loss of      potential loss of the    the RCS barrier.            potential loss 'of the    the containment          potential loss of the the fuel clad          fuel clad barrier.                                  RCS barrier.              barrier.                  containment barrier.
barrier.
53
 
Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The fuel clad barrier consists of the cladding material that contains the fuel pellets.
: 1. RCS or SG Tube Leakage There is no Loss threshold associated with RCS or SG Tube Leakage.
Potential Loss I.A This condition indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
: 2. Inadequate Heat Removal Loss 2.A This condition indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Potential Loss 2.A This condition indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
Potential Loss 2.B NOTE: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the fuel clad barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS barrier potential loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity/Containment Radiation 54
 
Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300&#xb5;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 of2 percent to 5 percent 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 3.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.
Loss 3.8 This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;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 of2 percent to 5 percent 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.
It is recognized that sample collection and analysis ofreactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no potential loss threshold associated with RCS activity/containment radiation.
: 4. Containment Integrity or Bypass Not applicable (included for numbering consistency)
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
55
 
RCS BARRIER THRESHOLDS:
The RCS barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the containment barrier loss threshold I.A will also be met.
Potential Loss I .A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the containment barrier loss threshold I .A will also be met.
Potential Loss 1.B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
: 2.      Inadequate Heat Removal There is no loss threshold associated with inadequate heat removal.
56
 
Potential Loss 2.A NOTE: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 g m due too erator action.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to fuel clad barrier potential loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity/Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold 3.A since it indicates a loss of the RCS barrier only.
There is no potential loss threshold associated with RCS activity/containment radiation.
: 4. Containment Integrity or Bypass Not applicable (included for numbering consistency)
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
57
 
CONTAINMENT BARRIER THRESHOLDS:
The containment barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss I.A and Loss I.A, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-0 I methodology. This determination is not necessarily dependent on entry into, or diagnostic steps within, an EOP. For example, ifthe pressure in a steam generator is decreasing uncontrollably [part ofthe FAULTED definition] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent ofa loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation ofa valve (e.g., a stuck-open safety valve) do meet this threshold.
58
 
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R !Cs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate                      Yes                        No Less than or equal to 25 gpm          No classification          No classification Greater than 25 gpm                Unusual Event per SU4      Unusual Event per SU4 Requires operation of a Site Area Emergency standby charging (makeup)                                          Alert per FA I        /
per FSl pump (RCS barrier potential loss)
Requires an automatic or              Site Area Emergency Alert per FA 1 manual ECCS (SI) actuation                  per FSl (RCS barrier loss)
There is no potential loss threshold associated with RCS or SG Tube Leakage.
: 2. Inadequate Heat Removal There is no loss threshold associated with inadequate heat removal.
Potential Loss 2.A This condition represents an IMMINENT core melt sequence that, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS barrier and the fuel clad barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the containment barrier.
The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing or if reactor vessel level is increasing. Whether the procedure(s) will be effective should be apparent within 15 minutes. The emergency director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1I50) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it 59
 
is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
: 3. RCS Activity/Containment Radiation There is no loss threshold associated with RCS activity/containment radiation.
Potential Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent 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 percent 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 containment that would then escalate the emergency classification level to a General Emergency.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. There may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2.
4.A. l - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the emergency director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment).
Two simplified examples are provided in the middle piping run ofFigure 6-F-1. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example is a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of 60
 
containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R !Cs.
4.A.2 - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in tum, communicate with the outside-the-plant atmosphere (e.g.,
through discharge ofa ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
See a simplified example in the top piping run ofFigure 6-F-1. The inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
A simplified example is shown in the bottom piping run of Figure 6-F-1. Leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. Ifthere is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. 1 to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R !Cs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using loss threshold I.A.
61
 
Loss 4.B Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment will be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly.
However, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
In the simplified example in the middle piping run of Figure 6-F-l, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. l to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS loss and/or potential loss threshold I .A to be met.
Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the containment barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and fuel clad barriers would already be lost. This threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Potential Loss 4.B 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 potential loss of the containment barrier.
Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment because containment heat removal/depressurization 62
 
systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the containment barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
63
 
Figure 6-F-1: PWR Containment Integrity or Bypass Examples
                                                      ,----------1      ......... a *0.        *:
* 4.i\.2 *_ Airborne *: *:*: *
: Effluent :  ** * * * * * * * *      *: *: *: release from : *: *: *. *
                                                                                                  '. . yat.h~~Y. . . '.
* Inside Containment l-~~~~t~~-::
Area Monitor ,:
:----------i
: Airborne '
                                                          ~~~~---------
RCP Seal Cooling 64
 
7    HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT              UNUSUAL EVENT EMERGENCY                EMERGENCY HGI HOSTILE              HSI HOSTILE                HAI HOSTILE              HUI Confirmed ACTION resulting in      ACTION within the          ACTION within the        SECURITY loss of physical control PROTECTED AREA.            OWNER                    CONDITION or threat.
of the facility.        Op. Modes: All              CONTROLLED AREA          Op. Modes: All Op. Modes: All                                      or airborne attack threat within 30 minutes.
Op. Modes: All HU2 Seismic event greater than OBE levels.
Op. Modes: All HU3 Hazardous event.
Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HAS Gaseous release impeding access to equipment necessary for normal plant operations, cooldown, or shutdown.
Op. Modes: All HS6 Inability to            HA6 Control Room control a key safety        evacuation resulting in function from outside      transfer of plant control the Control Room.          to alternate locations.
Op. Modes: All              Op. Modes: All HG7 Other conditions    HS7 Other conditions        HA7 Other conditions      HU7 Other conditions exist which in the      exist which in the          exist which in the        exist which in the judgment of the          judgment of the            judgment of the          judgment of the emergency director      emergency director          emergency director        emergency director warrant declaration of a warrant declaration of a    warrant declaration of    warrant declaration of a General Emergency.      Site Area Emergency.        an Alert.                NOUE.
Op. Modes: All          Op. Modes: All              Op. Modes: All            Op. Modes: All 65
 
HG1 ECL: General Emergency In itiatin g Condition: HOSTILE ACTION resulting in loss of phys ical control of the facility.
Operati ng Mode Appl icabili ty : All Em e rgency Actio n Levels:
(I)      a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the site seettrity feree      1 c..,    c                  ( Commented [JRB23]: RA! 21 b rev1S1on AND
: b.      EITHER of the following has occurred:
I. ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMfN ENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., thi s may include violent acts between individuals in the owner controlled area (OCA)).
IMMfNENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of m itigation or corrective actions.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facil ity 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 IMMfNENT damage to spent fuel due to I) 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 hift supervision and the control room is essential for proper classification of a security-related event.
66
 
Security plans and terminology are based on the guidance provided by NE! 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
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 Security Plan.
67
 
HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a General Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a General Emergency.
68
 
HS1 ECL: Site Area Emergency Initiating Condition: HO T ILE A TION within the PROTECTED AREA.
Operating Mode Applicabili ty: All Emergency Action Levels:
( I)    A HOSTILE ACTIO is occurring or has occurred within the PROTECTED AREA as reported by the site seelifity feree        \        l                                      ( Commented [JRB24] : RAJ 2J b revisio n Basis:
HOSTIL E ACTION: An act toward a nuclear power plant (NPP) or its personnel th at includes the use of vio lent force to destroy equipment, take HOST AGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land or water using gu ns, exp losives, PROJ ECTILES, veh icles, 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 di sobedience or felo nious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i .e., this may include violent acts between individuals in the owner controlled area (OCA)).
PROTECTED AREA (PA): The area that encompasses all controlled areas within the securi ty protected area fence.
This IC addresses the occurrence ofa HOST ILE ACTIO within the PROTECTED AREA (PA). This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security shift supervision 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 NE I 03- 12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and conditions allow, these eve nts 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 ORO reso urces and have them ava ilab le 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 app ly to a HOH!Le ACTIOJ>I eireetee et BH 1si;:s1 PROTeCTeD AREA Ieeatee etttsiee the pl!lflt PROTE.CTED A.-.:tEA (PA); stteh en ettaek sl!ettle Ile assesses ttsiRg IC HA I. It else eees Ret apply te incidents that are acc idental events, acts of civil disobed ience, or [ Commented [JRB25] : RAJ 22 re v1S1on otherw ise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small ai rcraft, shots from hunters, physical disputes between employees, etc.
Reporting of these types of events is adequate ly addressed by other EA Ls, or the requirements of IO CFR &sect; 73 .7 1 or10 CFR &sect; 50.72.
69
 
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 Security Plan.
Escalation of the emergency classification level uses IC HG 1.
70
 
HSS ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(!)    a.      An event has resulted in plant control being transferred from the control room to the remote shutdown panel.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* Core cooling
* RCS heat removal Basis:
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 ofa 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 "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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level uses IC FG I or CG 1.
71
 
HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration ofa Site Area Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(I)    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, I) toward site personnel or equipment that could lead to the likely failure of or, 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a Site Area Emergency.
72
 
HA1 ECL: Alert Initiating Condition: HOSTILE ACT IO              within the OWNER        0  TROLLED AREA o r airborne attack threat w ithin 30 minutes.
Operating Mode Applicability: All Emergency Action Levels: ( I or 2)
(I)      A HOSTILE ACTIO is occurring or has occurred within the OWNER CO TROLLED AREA as reported by the 15ite seeurity foree .. n ,11111 '>1.:u.ri. t pt in l rt 1.; !fin ...,.    ( Commented [JRB26]: RA! 21 b rev is ion (2)      A validated notificat ion from N RC of an aircraft attack threat within 30 minutes of the site.
Basis :
HOSTILE ACTION : An act toward a nuclear power plant (NPP) or its personnel that includes the use o f violent fo rce to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achi eve an end. This includes attack by air, land, or water using guns, explosives, PROJ ECTILES, vehicles, o r other devices used to deli ver destructi ve force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civi l di so bedi ence or felonious acts that are not part of a concerted attack o n the PP. No n-terrori sm-based EALs should be used to address such acti vi ties (i .e., thi s may include vio lent acts between indi viduals in the owner controlled area (OCA)).
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the contr?l of FNP securi ty.
This IC add resses the occurrence ofa HOSTILE ACTION within the OWNER CONTROLLED AREA (OCA) or notificat ion of an aircraft attack threat. This event w ill require rapid respo nse and assistance due to the possibility of the attack progress ing to the PROTECTED AREA (PA),
or the need to prepare the plant and staff fo r a potenti al aircraft impact.
Timely and accurate communications between Security shi ft supervision and the co ntrol room is esse ntial for proper class ification of a security-related event.
Securi ty plans and terminology are based o n the guidan ce prov ided by El 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
As time and co nditions allow, these events require a heightened state of readiness by the plant staff and implementat ion ofonsite protective measures (e.g., evacuatio n, dispersal, or sheltering).
The Alert declaration will also he ighten the awareness of offsite response organizations, allo wing them to be better prepared should it be necessary to consider further act ions.
T hi s IC does not appl y to incidents that are acc idental eve nts, acts of civil di sobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTIL E FORCE. Examples include the crash of a small ai rcraft shots from hunters, phys ical di sputes between employees, etc.
73
 
Reporting of these types of events is adequately addressed by other EA Ls, or the requirements of I0 CFR &sect; 73.71 or I0 CFR &sect; 50.72.
EAL# I is app licable for any HOST! LE ACTION occurring, or that has occurred, in the OWNER CO TROLLED AREA (OCA). This iRel1:1Eles !lR)' aetieA EliFeelea egaiAst aA 18F81 tllat is lee!llee 01:1tsiee tl~e pl!lflt PROTeCTeD AReA (PA).                                      { Commented [JRB27]: RAJ 22 revision EAL #2 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 station procedures.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe threat involves an aircraft. The status and size of the plane may be provided by ORAD through the RC.
In some cases, it may not be readil y apparent if an aircraft impact within the OWNER CO TROLLED AREA (OCA) was intentional (i.e., a HOSTILE ACTIO ). 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, will not be undul y delayed whi le 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. Securi ty-sensitive in formation sho uld be contained in non-public documents such as the Security Plan .
Escalation of the emergency classification level uses IC HS I.
74
 
HAS ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(1)    a.      Release of a toxic, corrosive, asphyxiant or flammable gas into any Table HI plant rooms or areas:
Table Hl Mode      Room Name                              Room Number Electrical Penetration Room            334, 333, 347 I 2334,2333,2347 Hallway Outside Filter Room            312, 332/
3                                              2312,2332 IA I 2A MCC areas Sample Room and Primary CHM labs      323, 324 I 2323,2324 Sample Room and Primary CHM labs      323, 324 I 2323,2324 4
RHRHxRoom.                            128/
2128 AND
: b.      Entry into the room or area is prohibited or impeded.
Basis:
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 plant safety.
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 director's 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.
75
 
Access should be considered as impeded if extraordinary 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 I 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 4.
* The gas release is a planned activity that includes compensatory measures to 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.
* An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19 percent, 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.
Escalation of the emergency classification level uses Recognition Category R, C or F !Cs.
76
 
HAG ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(I)    An event has resulted in plant control being transferred from the control room to the remote shutdown panel.
Basis:
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 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 uses IC HS6.
77
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
(!)      Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for an Alert.
78
 
HU1                                        1 ECL: Notification of Un usual Eve nt I
Initiating Condition: Confi rmed SECURJTY CO DITIO                  or threat.
Operating Mode Applicability: All Emergency Action Levels: ( I or 2 or 3)
( I)      A SECURJTY CONDITION that does not involve a HOSTILE ACTION as reported by the site seet1rit)' feree<*r ''    u 1~ (                                                  ( Commented [JRB28]: RAJ 21.b rev1S1on (2)      Notifi cation of a credible security threat directed at FNP.
(3)      A validated notification from the N RC providing information of an aircraft threat.
Basis:
SECURJTY CONDIT ION: Any Security Event as listed in the approved security conti ngency plan that constitutes a threat/compromise to site security, threat/ri sk to site personnel, or a potential degradatio n to the level of safety of the plant. A SECURJTY CONDITION does not invo lve a HOSTILE ACTION.
HOSTILE ACT ION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of vio lent force to destroy equipment, take HOST AGES, and/or intimidate the licensee to achieve an end. T his includes attack by air, land, or water using gu ns, explosives, PROJECTILEs, veh icles, 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 fe lonious acts that are not part of a concerted attack o n the N PP. Non-terrorism-based EALs should be used to address such acti vities (i.e., thi s may include vio lent acts between indi viduals in the owner controlled area (OCA)).
Thi s IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipm ent, and represe nt a potential degradation in the level of plant safety. Securi ty events wh ich do not meet one of these EA Ls are adequately addressed by the requirements of I 0 CF R &sect; 73. 71 or I 0 CF R &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under !Cs HA I, HS I and HG I.
Timely and accurate communications between Security Shift Supe rvisio n and the Contro l Room is essenti al for proper classification ofa security-related event. C lassi fication of these events w ill initiate appropriate threat-related notifications to plant personnel and OROs.
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 [and Independent Spent Fuel Storage Installation Security Program].
EAL# I references site security force because these are the indi viduals trained to confirm that a security event is occurring or has occurred. Training on security event confirm ation and classi fi catio n is controlled due to the nature of safeguards and I 0 CFR &sect; 2.39 information.
EAL #2 addresses the receipt ofa credible security threat. The credibili ty of the threat is 79
 
assessed in accordance with station procedures.
EAL #3 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HA I.
80
 
HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels.
Operating Mode Applicability: All Emergency Action Levels:
(1)    Seismic event greater than Operating Basis Earthquake (OBE) as indicated by seismic switch activation with the seismic system computer indicating EITHER of the following:
Cumulative Absolute Velocity (CAV) greater than 0.160 g-sec AND Spectral Acceleration greater than 0.200 g.
Cumulative Absolute Velocity (CA V) greater than 0.160 g-sec AND Spectral Velocity greater than 15.240 cm/sec.
Basis:
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 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 plant safety.
Event verification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should readily be felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are in excess of0.08g). The shift manager or emergency director may seek external verification if deemed appropriate (e.g., a call to the USGS or check of 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 uses IC CA6 or SA9.
81
 
HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3 or 4 or 5)
Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
(1)    A tornado strike within the PROTECTED AREA.
(2)    Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
(3)    Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
(4)    A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
(5)    Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours.
Basis:
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of plant safety.
EAL# 1 addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 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.
EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA (PA).
82
 
EAL #4 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, or dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in I 992, the flooding around the Cooper Station during the Midwest floods of I 993, or the flooding around Ft. Calhoun Station in 20 I I.
EAL #5 addresses phenomena of the hurricane based on the severe weather mitigation procedure.
Escalation of the emergency classification level is based on ICs in Recognition Categories A, F, Sor C.
83
 
HU4 ECL : Notification of Unusual Event I nitiating C o nditi on: FIRE potent ially degrad ing the level of afety of the plant.
Ope r ating Mode A ppli cability: A ll E m e rge ncy Ac tion Leve ls: ( I or 2 or 3 or 4)
Note : The e merge ncy director will declare the Unusual Event promptly upon determ in ing that the applicable tim e has been exceeded, o r w ill likely be exceeded.
( I)    a.      A FIRE is NOT extinguished within 15-minutes of A NY of the fo llowing FIRE detecti on indi cations:
* Report from the fi eld (i.e., visual observati o n)
* Receipt of multiple (more than I) fire alarms o r indi cations
* Field ve rifi cation of a single fire alarm A ND
: b.      The FIRE is located within ANY Table H2 rooms or areas.
(2)      a.      Receipt ofa single fire alarm (i.e., no other indi cations ofa FIRE).
AND
: b.      The FIRE is located within ANY Table H2 rooms or areas AN D
: c.      The ex iste nce of a FIRE is not ve rifi ed within 30-minutes of alarm receipt.
(3)      A FIRE w ithin the plant PROTECTED AREA er 18f81 PROTeCTt;O AREA not                        ( Commented [JRB29] : RA! 22 revision exting ui shed within 60-min utes of the initial repo rt, alarm or indi cation.
(4)      A FIRE w ithin the plant PROTECTED AREA er l8f81 PROTt;CTt;O ARM that                      ( Commented [JRB30] : RA! 22 revision requires fire fi ghting support by an offs ite fire response agency to extingui sh.
Table H2 Auxil iary Building Diesel Generator Building Servi ce Water Intake Structure (SW IS)
Containment R\.\S r                                                            Commented [JRB31] : RAJ 24 revision csr1                                                                Commented [JRB32]: RAJ 24 revision 84
 
Basis :
FIRE: Combustion characterized by heat and light. Sources of smoke such a slipping drive belts or overheated e lectrical equ ipment do not constitute FIRES. Observation of fl ame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTE D AREA (PA): The area that encompasses all controlled areas within the securi ty protected area fe nce.
This IC addresses the magnitude and extent of FIRES that may be indicative ofa pote ntial degradation of the leve l of plant safety.
The intent of the 15-minute duration i to size the FIRE and to di scriminate against small FIRES that are readil y extinguished (e.g., smoldering waste paper basket) . In addition to a larms, other indications of a FIRE include a drop in fire main pressure, automatic activation of a suppress ion system.
Upon rece ipt, operators wi ll take prompt actions to confirm the validity of an initia l tire a larm, indi cation, or report. Fo r EA L assessment purposes, the emergency decl aration clock starts at the tim e that the initial alarm , indication, or report was rece ived, and not the time that a subsequent verificatio n action was performed. Similarly, the tire duration clock a lso starts at the time of rece ipt of the initial a larm , indication or report.
This EAL addresses rece ipt of a s ingle tire a larm , and the existence of a FIRE is not verified (i.e., proved or di sproved) w ithin 30-minutes of the alarm . Upon rece ipt, operators will take prompt actions to confirm the validity of a sing le tire alarm. For EA L assess ment purposes, the 30-mi nute clock starts at the time that the initial alarm was rece ived, and not the time that a subsequent verification action was pe rfo rmed.
A s ingle tire alarm , absent other indi cation(s) of a FIRE, may be indicative of equipment fa ilure or a spurious activation, and not an actual FIRE. For thi s reason, additional tim e is allowed to verify the va lidi ty of the alarm. The 30-minute period is a reasonable amount of tim e to determine if an actual FIRE exists; however, a fter that tim e, and absent information to the contrary, it is assumed that an actua l FIRE is in progress.
I fan actual FIRE is verified by a report from the field, then EAL # I is immediate ly app licable, and the emergency must be declared if the Fl RE is not extingui shed within 15-minutes of the repo rt. lfthe a larm is ve rifi ed to be due to an equipment failure or a spurious acti vation, and this verification occurs w ithin 30-minutes of the rece ipt of the alarm, then thi s EA L is not applicable and no emergency dec laration is warranted.
In addition to a FIRE addressed by EAL# I o r EA L #2, a FIRE within the plant PROTECTE D AREA (PA) not extinguished within 60-minutes may also potentially degrade the leve l of plant safety. This basis e1i!eAds ta a flRE 0eet1rriRg withiA the PROTeCTED AREA (PA) efaA ISfSI leeated 0t1tsiae the plaRt PROTHCTHD AREA (PA).                                                    ( Commented [JRB33] : RA! 22 rev1S1on - - - - - -
85
 
lfa FIRE within the plant er ISfSl ,PROTECTED AREA is of sufficient size to require a                [ Commented [JRB34]: RAJ 22 rev isi on J response by an offsite firefightin g age ncy (e.g. , a local town Fire Department), then the level of plant safety is potentially degraded. The di spatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to active ly support fi refighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Basis-Rel ated Requirements from Appendix R Appendix R to I 0 CFR 50, states in part:
Criterion 3 of Appendix A to thi s part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requ irements, the probability and e ffect of fires and explosions."
When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assum e major importance to safety because dam age 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 ma intain safe shutdown conditions is greater than the need to limit fire damage to those systems required to miti gate the consequences of des ign bas is accidents.
Appendix R to I 0 CFR 50, requires, among other considerations, the use of I-hour fire barriers for the enclosure of cable and equipment and assoc iated non-safety circuits of one redundant train (G .2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this wo rst-case I-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level uses IC CA6 or SA9.
86
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a NOUE.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a NOUE.
87
 
8    SYSTEM MALFUNCTION ICS/EALS GENERAL                  SITE AREA EMERGENCY                  EMERGENCY                        ALERT                UNUSUAL EVENT SGI      Prolonged loss of SSI    Loss of all offsite  SAI    Loss of all but one SUI    Loss of all offsite all offsite and all onsite and all onsite AC power      AC power source to          AC power capability to AC power to emergency      to emergency buses for      emergency buses for 15      emergency buses for 15 buses.                    15 minutes or longer.        minutes or longer.          minutes or longer.
Op. Modes: Power          Op. Modes: Power            Op. Modes: Power            Op. Modes: Power Operation, Startup, Hot    Operation, Startup, Hot      Operation, Startup, Hot    Operation, Startup, Hot Standby, Hot Shutdown      Standby, Hot Shutdown        Standby, Hot Shutdown      Standby, Hot Shutdown SA2 UNPLANNED              SU2 UNPLANNED loss of Control Room        loss of Control Room indications for 15          indications for 15 minutes or longer with a    minutes or longer.
significant transient in    Op. Modes: Power progress.                  Operation, Startup, Hot Op. Modes: Power            Standby, Hot Shutdown Operation, Startup, Hot Standby, Hot Shutdown SU3    Reactor coolant activity greater than Technical Specification allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS    Inability to        SAS    Automatic or        SUS Automatic or shutdown the reactor        manual trip fails to        manual trip fails to causing a challenge to      shutdown the reactor, and  shutdown the reactor.
core cooling or RCS heat    subsequent manual          Op. Modes: Power removal.                    actions taken at the        Operation Op. Modes: Power            reactor control consoles Operation                    are not successful in shutting down the reactor.
Op. Modes: Power Operation 88
 
GENERAL                  SITE AREA                      ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY SU6    Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Failure to iso 1ate containment or loss of containment pressure control.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS    Loss of all AC    SSS    Loss of all vital and vital DC power        DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 89
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
(I)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 4160V ESF busses 1(2)F AND 1(2)G.
AND
: b.      EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* CORE COOLING CSF - RED conditions met.
Basis:
This IC addresses a prolonged loss of all power sources to AC emergency busses. 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG I. This will allow additional time for implementation of offsite protective actions.
Escalation of the emergency classification from Site Area Emergency will occur ifit is projected that power cannot be restored to at least one AC emergency 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 emergency bus will be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success will 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 ifthe loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
90
 
SGS ECL: General Emergency Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 4 l 60V ESF busses 1(2)F AND 1(2)G for 15 minutes or longer.
AND
: b.      Indicated voltage is less than 105 VDC on ALL 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. 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 loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
91
 
551 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    Loss of ALL offsite and ALL onsite AC power to BOTH 4160V ESF busses 1(2)F AND 1(2)G for 15 minutes or longer.
Basis:
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 is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses !Cs RG 1, FG 1 or SG 1.
92
 
555 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.
Operating Mode Applicability: Power Operation Emergency Action Levels:
Note: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 gpm due to operator action.
(!)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      All manual actions to shutdown the reactor have been unsuccessful.
AND
: c.      EITHER of the following conditions exist:
* Core Cooling CSF - RED conditions met
* Heat Sink CSF - Red conditions met Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 warrants the declaration ofa 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 because the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level uses IC RG 1 or FG I.
93
 
558 ECL: Site Area Emergency Initiating Condition: Loss of all vital DC power for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    Indicated voltage is less than 105 VDC on ALL 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RG 1, FGI or SGS.
94
 
SA1 ECL: Alert Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode App licability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director wi ll declare the Alert promptly upon determining that 15 minutes has been exceeded, or will li kely be exceeded.
( I)    a.      AC power capability to BOTH 4160V ESF busses I (2)F AND I (2)G is reduced to a ingle power source *1able', I) fo r 15 minutes or longer.                        ( Commented [JRB35] : RAJ 13 b rev1s1on AND
: b.      Any additio nal si ngle power source fai lure will result in a loss of all AC power to SAFETY SYSTEMS .
Table S I Unit I                                      Unit2 Start-up Aux XFMR IA                      Start-up Aux XFMR 2A Start-up Aux XFMR I 8                      Start-up Aux XFMR 28 Diesel Generator l-2A                      Diesel Generator 1-2A Diesel Generator I 8                      Diese l Generator 28 Diesel Generator IC                        Diesel Generator IC Diesel Generator 2C                        Diese l Generator 2C Basis:
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 typicall y systems classified as safety-related.
This IC describes a significant degradation of otfsite and onsite AC power source where any additional si ngle fa ilure 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 SU I.
An "AC power source" is a source recogni zed in AOPs and EO Ps, and capable of suppl ying requi red power to an eme rge ncy bus (see Table SI above). ome exam ples of this condition are presented below.
* A loss of all offsite power with a concurrent fai lure of all but one emergency power source (e.g., an onsite diesel generator).
95
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generator ) with a single train of emergency bus es being back-fe d from the unit main generato r.
* A loss of emergency power sources (e.g., onsite diesel generato rs) with a single train of emergency busses being back- fed from an offsite power so urce.
Fiftee n minutes is the threshold to exclude transient or momentary losses of power.
Escalation of the emergency class ification level uses IC SS 1.
96
 
                                                                                                                                            - 1 I
I SA2 ECL: Alert Initiating Co ndition : UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Act ion Levels :
Note: The emerge ncy director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or w ill likely be exceeded.
( I)    a.      An UNPLANNED event results in the inab ili ty to monitor one or more of the following parameters from w ithin the Control Room for 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Ex it Temperature Wide Range Levels in at least one steam ge nerator Steam Generator Auxiliary _      -            feed Water Flow                    ( Commented [JRB36]: RAI 27 rev1S1on AND
: b.      ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% fu ll electrical load
* Reactor trip
* ECCS actuation Basis:
UNPLANNED: A parameter change or an event that is not I) the result of an intended evo lution or 2) an expected plant respo nse to a transient. The cause of the parameter change o r event may be known or unknown.
This IC add resses the difficulty associated w ith monitoring rapidly changing plant conditions during a transient without the abili ty to obtain SAFETY SYSTEM parameters from within the Control Room . During this condition, the margi n to a potential fission product barrier challenge is reduced. It represe nts a pote nti al substantial degradation in the leve l of plant safety.
As used in this EAL, an " inability to monitor" means that values for one or more of the listed paramete rs cannot be determined from within the Control Room . This situatio n 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 . Various instrumcntat1on is also uscJ tn d..:tcrminc RCS LL*vcl RV! !"I, pr.:ssuri/..:r kvel, digital or r..:cordcr,;. A los~ of all Control Room ,;ourccs for this partunctcr woulJ also appl).                                                                                      ( Commented [JRB37]: RAJ 26 rev1S1on 97
 
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with IO CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event is 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, core cooling 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses ICs FS I or IC RS I.
98
 
SAS ECL: Alert Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Emergency Action Level:
(I)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 plant safety. An emergency declaration is required even ifthe 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 consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
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 consoles (e.g., locally opening breakers). 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".
The plant response to the failure of an automatic or manual reactor trip 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 shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS 1, 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
99
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train ofa SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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 iflarge quantities of smoke and heat are observed.
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 may require a post-event inspection to determine ifthe attributes of an explosion are present.
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.
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.
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 100
 
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 plant safety.
The first threshold for EAL 1.b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL 1.b 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 all 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 uses IC FS I or RS I.
101
 
SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode A pplicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels :
Note: The emergency directo r w ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( I)    Loss of ALL offsite AC power capability { rabl~ '-;2) to BOTH 4160V E F busses 1(2)F    { Commented [JRB38]: RAJ 13 b revision AN D 1(2)0 for 15 minutes or longer.
Table S2 Unit I                                      Unit 2 Start-up Aux XFMR IA                          Start-up Aux XFMR 2A Start-up Aux XFMR I B                        Start-up Aux XFMR 28 Basis :
Thi s re 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 emergency busses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offs ite AC power source(s) is available to the emergency busses (see Table S2 above), whether or not the busses are powered from it.
Fifteen minutes is the thres hold to exclude transient or momentary losses ofoffsite power.
Escalation of the emergency classification level uses IC SA I.
102
 
SU2 ECL: Notificatio n of Unusual Event Initiating Conditio n: UN PLANNE D loss of Co ntro l Roo m indications for 15 minutes or longer.
O peratin g Mode Applicab ili ty:          Power Operation, Startup, Hot Standby, Hot Shutdown Emerge ncy Actio n Levels:
Note: The Emergency Director should declare the Unusual Event promptly upo n determining that 15 minutes has been exceeded, o r will likely be exceeded.
( I)      An UNPLANNED event results in the inability to monitor one or more of the following parameters from w ithin the Contro l Room fo r 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Leve l in at least one steam ge nerator Steam Generator A uxiliary :;* : ... _        - Feed Water Flow                      ( Commented [JRB39]: RAJ 27 rev1S1on Bas is:
UNPLANNED: A parameter change or an event that is not I) the result of an intended evo lution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
This IC addresses the difficulty associated with monitoring norm al plant conditi ons w ithout the abil ity to obtain SAFETY SYSTEM parameters from within the Co ntrol Roo m. This condition is a precurso r to a more significant event and represents a potential degradation in the level of plant safety.
As used in this EAL, an " inabili ty to monitor" means that va lues for one or more of the listed paramete rs cannot be determined from within the Control Room. Th is 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 Co ntrol Room . lvm*ious instrumentation is abo us..:d to Jet..:rmin.: RCS I.eve! RVLl'i prcssuri1cr level. digital or r.:corJ..:rs. /\ loss of all C. onlrol R<xim sources for this param..:tl'r
\\could also appl}                                                                                      [ Commented [JRB40] : RAJ 26 rev1S1on An event invo lving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with IO CFR 50.72 (and associated guidance in UREG- 1022) to determine if an NRC event report is required. The event is reported if it significantly impai red the capability to perform emerge ncy assessments. In particular, emergency assessments necessary to imp lement abnormal operati ng proced ures; emergency operating procedures; and emergency plan implementing procedures addressi ng emergency classification, acc ident assessment, o r protective action decision-making.
103
 
This EAL is focused on a selected subset of plant parameters associated with the key safety functions ofreactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the control room is considered 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
104
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)    RCS coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits as indicated by ANY of the following:
Dose Equivalent I-131 greater than 0.5 &#xb5;Ci/gm for greater than 48 hours Dose Equivalent I-131 greater than Technical Specification figure 3.4.16-1.
IF less than 20% power, THEN use the Dose Equivalent I-131 20% power limit on Technical Specification figure 3.4.16-1 RCS gross specific activity greater than I 00/E &#xb5;Ci/gm.
Basis:
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 plant safety.
Escalation of the emergency classification level uses Ics FA I or the Recognition Category R I Cs.
105
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)      RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)      Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
This IC addresses RCS leakage that could 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 plant safety.
EAL# 1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL# 1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification is required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level uses ICs of Recognition Category R or F.
106
 
SUS ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (1 or 2)
(1)      a.      An automatic trip did not shutdown the reactor.
AND
: b.      A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a.      A manual trip did not shutdown the reactor.
AND
: b.      EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic trip is successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and represents a potential degradation of the level of plant safety.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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 plant's decay heat removal systems.
If an initial manual reactor trip 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 trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
This action does not include manually driving in control rods or implementation of boron 107
 
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".
The plant response to the failure of an automatic or manual reactor trip 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 SA5. Depending upon the plant response, escalation is also possible via IC FA I. Absent the plant conditions needed to meet either IC SA5 or FA!, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance will be applied.
* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and will be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.
108
 
SUS ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1or2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 theNRC.
This IC will be assessed only when extraordinary means are used 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).
EAL #I addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Alabama, Georgia, and Florida; Houston and Henry Counties, Alabama; and Early County, Georgia.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
109
 
SU7 ECL: Notification of Unusual Event Initiating Condition: Failure to isolate containment or loss of containment pressure control.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2)
(!)      a.      Failure of containment to isolate when required by an actuation signal.
AND
: b.      ALL required penetrations are not closed within 15 minutes of the actuation signal.
(2)      a.      Containment pressure greater than 27 psig.
AND
: b.      Less than one CTMT fan cooler AND one full train ofCTMT spray is operating per design for 15 minutes or longer.
Basis:
This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety.
For EAL #1, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - will be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. At Farley, a single CTMT fan cooler along with one train ofCTMT spray is required per design basis. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event will escalate to a Site Area Emergency in accordance with IC FS 1 ifthere were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
I JO
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure ATWS ................................................................................... Anticipated Transient Without Scram CC ......................................................................................................................... Cubic Centimeter CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CHM ................................................................................................................................. Chemistry CPM .................................................................................................................... Counts Per Minute CTMT/CNMT ............................................................................................................... Containment CSF ............................................................................................................. Critical Safety Function DC .............................................................................................................................. Direct Current EAL ........................................................................................................... Emergency Action Level ECCS ............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level ENN ............................................................................................. Emergency Notification Network ENS ................................................................................................ Emergency Notification System EOP ............................................................................................... Emergency Operating Procedure EPA ............................................................................................. Environmental Protection Agency ESF .......................................................................................................... Engineered Safety Feature FAA ............................................................................................... Federal Aviation Administration FBI .................................................................................................. Federal Bureau of Investigation FEMA ............................................................................. Federal Emergency Management Agency FNP .................................................................................................................. Farley Nuclear Plant FTS ......................................................................................... Federal Telecommunications System GE ...................................................................................................................... General Emergency HOO .................................................................................. Headquarters Operations Officer (NRC)
Hx ............................................................................................................................. Heat Exchanger IC ........................................................................................................................ Initiating Condition ID ............................................................................................................................. Inside Diameter ISFSI ........................................................................... Independent Spent Fuel Storage Installation K.rr ...................................................................................... Effective Neutron Multiplication Factor MCB .................................................................................................................. Main Control Board MCC ................................................................................................................ Motor Control Center
&#xb5;Ci ................................................................................................................................. micro-Curie mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. North American Aerospace Defense Command NOUE .............................................................................................. Notification Of Unusual Event OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM ........................................................................................... Offsite Dose Calculation Manual ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG ....................................................................................................... Protective Action Guideline PBX ........................................................................................................... Private Branch Exchange PWR ........................................................................................................ Pressurized Water Reactor A-1
 
PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCDT ................................................................................................... Reactor Coolant Drain Tank RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RHR ............................................................................................................. Residual Heat Removal RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RVLIS ...................................................................... Reactor Vessel Level Instrumentation System SAE ................................................................................................................. Site Area Emergency SCBA ..................................................................................... Self-Contained Breathing Apparatus SFP ........................................................................................................................... Spent Fuel Pool SG ........................................................................................................................... Steam Generator SI .............................................................................................................................. Safety Injection SNC ....................................................................................................... Southem Nuclear Company SPDS ............................................................................................ Safety Parameter Display System TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel TV ........................................................................................................................... Threshold Value VDC .................................................................................................................. Volts Direct Current VOiP ................................................................................................... Voice Over Internet Protocol WHT ................................................................................................................. Waste Holdup Tank A-2 J
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
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.
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.
Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The following are key terms necessary for overall understanding the NE! 99-0 l emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL): 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 (!) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NOUE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT CLOSURE: Per FNP-1 (2)-STP-18.4, "Containment Integrity Verification and Closure".
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 may require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The area that encompasses all controlled areas within the FNP site boundary but outside the security protected area fence.
B-2
 
PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
REFUELING PATHWAY: This includes the reactor refuel cavity, the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
RUPTURE(D): The condition ofa steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
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.
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.
UNISOLABLE: 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.
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.
B-3 J
 
Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 Responses to Requests for Additional Information EDWIN I. HATCH NUCLEAR PLANT EAL SCHEME MARKED-UP PAGES
 
HATCH NUCLEAR PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND .................................................................................. 1 1.1 OPERATING REACTORS .................................................................................................. !
1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ..................................... l 1.3 NRC ORDER EA-12-051 ................................................................................................ 2 1.4 0RGANIZATION AND PRESENTATION OF INFORMATION ............................................... 3 1.5 IC AND EAL MODE APPLICABILITY ..............................................................................3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS **.**......**..*...*...*..**..***..*... 5 2.1 GENERAL CONSIDERATIONS .*********************.*********.*******.**************.**...***************...*********** 5 2.2 CLASSIFICATION !VIETHODOLOGY ..............................................................._**************.*** 6 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ********.***************..***************. 6 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ................................ 6 2.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................... 7 2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ******..*********.* 7 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS ................................................................. 7 2.8 CLASSIFICATION OF TRANSIENT CONDITIONS ..............................................................7 2.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION **************** 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS **.**..***.*..***..*..*.* 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS ***..***.*...*****. 26 -
5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ............. 47 6  FISSION PRODUCT BARRIER ICS/EALS ............................................................... 50 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ........ 64 8  SYSTEM MALFUNCTION ICS/EALS ........................................................................ 88 APPENDIX A - ACRONYMS AND ABBREVIATIONS ...................................................... A*1 APPENDIX B - DEFINITIONS........................................................................................ B*1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections of this document are:
* 10 CFR &sect; 50.47(a)(l)(i)
* 10 CFR &sect; 50.47(b)(4)
* IO CFR &sect; 50.54(q)
* IO CFR &sect; 50.72(a)
* IO CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* IO CFR &sect; 50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by various regulatory guidance documents.
Documents of particular relevance to NEI 99-0I include:
NUREG-0654/FEMA-REP-I, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix I, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG- I 022, Event Reporting Guidelines 10 CFR &sect; 50. 72 and&sect; 50. 73 Regulatory Guide I. l 01, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions may also be directed to the NE!
Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NEI 99-0 I is applicable to licensees electing to use their I 0 CFR 50 emergency plan to fulfill the requirements of IO CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR &sect; 50 and the guidance in NUREG 0654/FEMA-REP- l. The initiating conditions germane to a 10 CFR &sect; 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a 10 CFR &sect; 50.47 emergency plan.
The generic I Cs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HU I covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. Additionally, appropriate aspects ofIC HUI and IC HAI will also be included to address a HOSTILE ACTION directed against an ISFSI.
I
 
The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed I rem Effective Dose Equivalent.
1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision-makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (I) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool.
* A display in an area accessible following a severe event.
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-I 2-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-05 I.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These 2
 
EALs are included within existing IC RA2, and new !Cs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the plan's effectiveness. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with I 0 CFR 50.90.
1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The
    !Cs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode              R          c        E          F        H          s Power Operations          x                    x          x          x          x Startup            x                    x          x          x          x Hot Shutdown          x                    x          x          x          x Cold Shutdown          x          x        x                    x Refueling          x          x        x                    x Defueled            x          x        x                    x 3
 
Hatch Units I and 2 Technical Specifications Table I.I-I provides the following mode definitions:
Average Reactor Reactor Mode Mode            Title                                                  Coolant Switch Positon Temocrature ('F)
I      Power Ooeration    Run                                          NA 2      Startup            Refue1<*J or Startup/Hot Standby            NA 3      Hot Shutdown(*)    Shutdown                                  >212 4      Cold ShutdownCbJ    Shutdown                                  :S 212 5      RefuelingCbJ        Shutdown or Refuel                          NA (a)  All reactor vessel head closure bolts fully tensioned.
(b)  One or more reactor vessel head closure bolts less than fully tensioned.
In addition to these defined modes, "Defueled" is also applicable to the Hatch EAL scheme, consistent with NEI 99-0 I. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Hatch EALs with no modifications from NEI 99-0 I.
When a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
4
 
2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes, and the informing basis information. In the recognition category F matrices, EALs are referred to as fission product barrier thresholds; the thresholds serve the same function as an EAL.
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.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The indications will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 will be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with the operating license is maintained, provided the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of IO&sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
5
 
Although the EA Ls have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision fo r classification based on operator/management experience and judgment is sti ll necessary. The NEI 99-0 I scheme provides the emergency director with the ability to cla sify events and conditions based upon judgme nt using EALs that are consistent with the emergency classification leve l (ECL) definitions (refe r to Category H). The emergency director will need to determine ifthe effects or consequences of the event or condition rea onably meet or exceed a particular ECL definition. A similar provision is incorporated into the fi ssion product barrier tables; judgment may be used to determine the status of a fi ssion product barrier.
2.2  CLASS IFICATION METHODOLOGY To make an emerge ncy classification, the user wi ll compare an event or condition (i.e. ,
the relevant plant indications and reports) to an EAL(s) and determine if the EAL has bee n met or exceeded. An EAL(s) evaluation must be consistent with the re lated operating mode applicability and notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is declared in accordance with plant procedures.
When assess ing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrent ly with the e merge ncy classification process "clock." For a full discussion of this timing requirement, refer to SIR/DPR-ISG-01 .
2.3  CLASSIFICATIO ' OF M ULT IPLE EVE TS AND CONDITIONS In the event of multiple emergencies or conditions, the user wi ll identify all met or exceeded EALs. The highest applicab le ECL identified during this review is declared.
For example:
If an Alert EAL and a Site Area Emerge ncy EAL are met, a Site Area Emergency will be declared.
There is no "add iti ve" effect from multiple EA Ls meeting the same ECL. For example:
If two Alert EALs are met, an Alert will be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Iss ue Summary (RJS) 2007-02, Clarification ofN RC Guidance for Emergency Notifications During Quickly Changing Events.
2 .4 CONSIDERATION OF MODE CHANGES D URI NG CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, determines whether an IC is applicable. If an event or condition occurs and results in a mode change before the emergency is declared, the emergency class ification leve l 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 dilkrcnt mode 1s n:ached, an~ nC\\ event or condition not rclatcJ to the original event or condition. requiring enicrgcnc) classilicaiion shoulJ be evaluateJ against the !Cs mid EALs apr.Iicablc to the operating mode at the time of the nc\\ event or condition.                                  ( Commented [JRBl] : RAI I revJSion 6
 
2.5 CLASSIFICATION OF IMMI NENT 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 IMMIN ENT). If, in the judgment of the emergency director, meeting an EAL is IMMIN ENT, the emergency classification will be made as ifthe EAL has been met. While applicable to all emergency classification leve ls, this approach is particularly important at the higher emergency classification leve ls since it provides additional time for implementation of protective measures.
2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRAD ING SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower class ification. Term ination criteria contained in procedure NMP-EP-110, Emergency Classification and Initial Actions shall be completed for an event to be terminated. At termination, on an event specific basis, the site can either enter normal operati ng conditions or enter a recovery condition with a recovery organization established for turnover from the ERO.
2.7 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 end before the emergency classification assessment can be completed,-. rran evt'nt occurs that meet-; or C'<CCt'ds an EA! , the associated !-.Al must be declared regardless of its continued pre~ence at the time of declaration. For example, an earthquake, or failure    [ Commented (JRB2] : RAJ 2 revision of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
2.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the !Cs and/or EALs in this document employ time-based criteria that require 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, some transient conditions may cause an EAL to be met for a brief period of time. The following guidance wi ll be applied to the classification of these conditions.
EAL momentarily met during expected plant response - When 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 emerge ncy 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 applicable EAL is not considered met and the associated emergency declaration is not required. This example presents an illustration :
An ATWS occurs and RCIC fai ls to automatically start. RPY level rapidly decreases and the plant enters an inadequate RPV Water Level condition (a 7
 
potential loss of the fuel clad barrier and a loss of the RCS barrier). If an operator manually starts RCIC in accordance with an EOP step and clears the inadequate RPV Water Level condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to note that the 15-minute emergency classification assessment period is not a "grace period" to delay a classification in order to perform 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 where an operator is able to take corrective action before the emergency director completes the review and necessary steps to make the emergency declaration. This provision ensures any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
2.9 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. Personnel could discover 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. It may be the event or condition was not recognized at the time, or there was an error in the emergency classification process.
In these cases, no emergency declaration is warranted, but the guidance contained in NUREG-1022 is applicable. Specifically, the event will be reported to the NRC in accordance with 10 CFR &sect; 50.72 within one hour of the discovery of the undeclared event or condition. The licensee will also notify appropriate State and local agencies in accordance with the agreed-on arrangements.
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3    ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL                  SITE AREA EMERGENCY                                              ALERT            UNUSUAL EVENT EMERGENCY RGI Release of            RSI Release of          RAl Release of        Rut Release of gaseous radioactivity    gaseous radioactivity  gaseous or liquid      gaseous or liquid resulting in offsite      resulting in offsite    radioactivity resulting radioactivity greater dose greater than 1,000  dose greater than 100  in offsite dose greater than 2 times the mrem TEDE or 5,000        mrem TEDE or 500        than IO mrem TEDE      ODCM limits for 60 mrem thyroid CDE.        mrem thyroid CDE.      or 50 mrem thyroid      minutes or longer.
Op. Modes: All            Op. Modes: All          CDE.                    Op. Modes: All Op. Modes: All RG2    Spent fi.1el pool RS2 Spent fuel po(il    RA2 Significant        RU2 UNPLANNED k1 el cn111wt be          le\ el al Le1*ei 3. lowering of water level loss of water level restored lo at least      Ofl. .\lodes: All      above, or damage to,    above irradiated fuel.
Level 3 for 60 minutes                            irradiated fuel.        Op. Modes: All ur longer*.                                      Op. Modes: All Op ..\lodes.* .Ill RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL # 1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reactor Building Vent Accident Range Monitor:
IDil-P601(feeding1Dll-R631, Rx Bldg Vent Wide Range)        2.6 &#xb5;Ci/cc 2Dl l-P601 (feeding 2Dl 1-R631, Rx Bldg Vent Wide Range)      2.6 &#xb5;Ci/cc Main Stack Accident Range Monitor:
ID! 1-P007 (feeding IDl 1-R631, Main Stack Wide Range)      8.1 x I 0 3 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
Basis:
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 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 IO
 
inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used to determine the monitor reading threshold values in ICs RS 1 and RA I. This protocol will maintain intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at the EPA PAO of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAO 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.
II
 
RG2 CCL: Gcncrnl F.mcrgcncy Initiating Condition: Spent fuel [)l)Ol lc\*d cannot k restored lo at ka:'t Leh: I 3 tl\r 60 minute:'
or longer.
Operating :\lode .\pplitabilit): 1\ll Emergency Action        Lcn~Js:
'.'iote: The .orncrgency director 1\ill d<Cclarc till* Gcncrnl l*:mcrgcncy promptly upun dctcr111ini11g tbu1 6lJ rninllkS llclS bci.:n i.:\Cccdcd. or 1Yill likcly be c\cceckd.
(I)      Spent Ji1cl pool levc'l cannot be rcston::cl to at k:1st l.c\*cl :l for 60 minute, or longer.
Basis:
This IC addrcosc,; a signilicam lo,;s or spent li1el pool inwnt\lr) control and maki:np capahilit:
lc~1ding to a prolu11gccl uncO\'Cl')' of spent fi1d. This condition 11*ill lcacl to l'ucl damage ancl cl rudiological rc:lca:'e to thc cnl'ironmrnt.
It i;; recognized that this IC would likely not be met until w~ll ulkr another Cienernl Emergency IC wa;; m-:t:  huw~vc:r. it is included to prol'ide cia:'silk:ition clivcr*sity.
12
 
R51 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL# I should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
( 1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reactor Building Vent Accident Range Monitor:
                      !DI l-P601 (feeding !DI l-R631, Rx Bldg Vent Wide Range)    2.6 x 10- 1 &#xb5;Ci/cc 2Dl l-P601 (feeding 2Dl l-R63 l, Rx Bldg Vent Wide Range)    2.6 x 10- 1 &#xb5;Ci/cc Main Stack Accident Range Monitor:
ID! l-P007 (feeding !DI l-R631, Main Stack Wide Range)      8.1 x 102 &#xb5;Ci/cc (2)      Dose assessment using actual meteorology indicates doses greater than I 00 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.
(3)      Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to IO-percent 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 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 13
 
inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in ICs RG 1 and RA 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 100 mRJhour whole body or 500 mRJhour thyroid, whichever is more limiting.
The TEDE dose is set at IO-percent 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. Ifthe 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 uses IC RGI.
14
 
RS2 ECL: Sile: Arca Enh.:rgenc)
Initiating Condition: Spent fuel pool level at Level 3.
Operating Mode .\pplfrahility: All Emergency . .\ction Levels:
( l)    l .owcring of spent fuel pool level to Level 3.
Basis:
*r*bis iC uddrcsscs a signilicant loss ofspcm li1el pool inventory control and makeup capability kading w !Mtvl!NENT fuel damage. This condition sterns frn111 major failures of plant fi.111ctions needed for protection of the public and warrant a Site t\ren Emergency declarntion.
lt is recognized that this IC \\'Otild like!; not be met until \\*ell afkr another Site Ar.ca Emergency IC was m-:t: however. it is included lo provide classilkation diversity.
Escalation of the c:mcrgenc.:y classifica1ion level uses JC RG l or RG2.
15
 
RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3 or 4)
Notes:
* The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #I should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(I)    Reading on ANY of the following radiation monitors greater than the reading shown for I 5 minutes or longer:
Reactor Building Vent Accident Range Monitor:
IOI l-P601 (feeding IOI l-R63 I, Rx Bldg Vent Wide Range)      2.6 x 10-2 &#xb5;Ci/cc 201 l-P601 (feeding 201 l-R63 l, Rx Bldg Vent Wide Range)      2.6 x I 0-2 &#xb5;Ci/cc Main Stack Accident Range Monitor:
IOI 1-P007 (feeding IOI l-R631, Main Stack Wide Range)          8.1 x 10 1 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid COE at or beyond the site boundary.
(3)    Analysis ofa liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid COE at or beyond (site-specific dose receptor point) for one hour of exposure.
(4)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid COE greater than 50 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to I-percent 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 plant safety as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled 16
 
release).
Radiological effluent EALs are 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in ICs RGI and RS I. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and that the threshold values are based on a site boundary (or beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at I-percent 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 uses IC RS I.
17
 
RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3)
(I)    Uncovery of irradiated fuel in the REFUELING PATHWAY.
(2)    Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on ANY Table Rl radiation monitors:
Table RI Refuel Floor Area Radiation Monitors Unit I                                          Unit2 I D2 l-K60 I A - Rx Head Laydown Area              2D21-K601 A- Rx Head Laydown Area I D2 l-K60 I B - Refueling Floor Stairway          2D21-K601 M - Spent Fuel/Fuel Pool Areas ID21-K601 D -Refuel Floor                          2D21-K60 I E - Dryer/Separator Pool ID21-K60! E- Drywell Shield Plug                  2D21-K61 ! K - RPV Refuel Floor 228' ID2 l-K60 I M - Spent Fuel Pool and New Fuel      2D21-K61 ! L - RPV Refuel Floor 228' Storage area Refuel Floor Ventilation Monitors Unit I                                          Unit2 ID! l-K609 A-D-Rx Bldg. Potential                  201 l-K609 A-D -Rx Bldg. Potential Contaminated Area Vent Exhaust Rad Monitor        Contaminated Area Vent Exhaust Rad Monitor
          !DI l-K61 l A-D -Refuel Floor Vent Exhaust        201 l-K61 l A-D - Refuel Floor Vent Exhaust 2011-K634 A-D - Refuel Floor Rx Well Vent.
Exhaust 201 l-K635 A-D - Refuel Floor OW/Sep. Vent.
Exhaust (3)    Lowering of spent fuel pool kn::! to Level 2.
Basis:
REFUELING PATHWAY: This includes the reactor cavity, the transfer canal, and the spent fuel pool.
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 spentfi1el 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 plant safety.
This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded 18
 
storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with ICE-HUI.
Escalation of the emergency is based on either Recognition Category R or C ICs.
This EAL escalates from RU2. 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), 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 will be considered in combination with other available indications of inventory loss.
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.
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 will be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
Spent J"ud pool watl.':r 11.':vi;:i at this value is \\ithin lhc lower l.':nd oftlw kvcl range necessary to prevent significant dose i.:onscqucncc'S from direct gamma radiation to personnel performing opl.':rntions in the vicinity of the spent li.1ci pool. This condition rcJlccts a signiiic:ant loss of spent fuel pool water im*cntory and is a precursor to a loss of the ability to adequately cool the irradiated !"uel assembles stored in the pool.
Escalation of the emergency classification level uses I Cs RS 1 or RS2 19
 
RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
Note: 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.
(!)    Dose rate greater than 15 mR/hr in ANY of the following areas:
Control Room area radiation monitor ID2 l-K600 8 or C    \
Central Alarm Station (by survey)                        I (2)    An UNPLANNED event results in radiation levels that prohibit or impede access to any Table HI plant rooms or areas:
Table Hl Buildiu                              Rooms                      Aoolicable Modes Diesel enerator buildin      All                                                All Unit 1/2 130'                                      All Reactor building            Unit 1/2 SE Diagonals (RHR)                        All
        ~---------~U_n_it 1/2 NE Diagonals (RHR)                                        All Basis:
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses elevated radiation levels in certain plant rooms or 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. It represents an actual or potential substantial degradation of the level of plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL #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.
20
* 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 I 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.
Escalation of the emergency classification level uses Recognition Category R, C or F ICs.
21
 
RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
(1)      Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
Reactor Building Vent Normal Range Monitor:
ID! l-K619 A(B) 20 l l-K636 A(B)
Main Stack Normal Range Monitor:
ID l !-K600 A(B)
Liquid Radwaste Effluent Line Monitor:
IDl!-K604 201 !-K604 Service Water System Effluent Line Monitor:
IDl!-K605 201 l-K605 (2)      Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
(3)      Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
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.
22
 
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment indicates 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.
EAL# 1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EAL #3 - 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 uses IC RA!.
23
 
RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels:
(1)    a.      UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Personnel report of low water level SFP low level alarm annunciator - Spent Fuel Storage Pool Level Low 654-022-1/2 AND
: b.      UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
ID2l-K601 A - Rx Head Laydown Area ID21-K601 D - Refuel Floor ID21-K601 E-Drywell Shield Plug ID21-K601 M - Spent Fuel Pool and New Fuel Storage area 2D21-K601 A -Rx Head Laydown Area 2D21-K601 M- Spent Fuel/Fuel Pool Areas 2D21-K60! E - Dryer/Separator Pool 2D21-K61 l K-RPV Refuel Floor 228' 2D21-K61 l L-RPV Refuel Floor 228' Basis:
REFUELING PATH WAY: This includes the reactor cavity, the transfer canal, and the spent fuel pool UNPLANNED: A parameter change or an event that is not I) 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.
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 plant safety.
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 24
 
(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 ofa 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 with Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level uses IC RA2.
25
 
4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL                  SIT E A REA ALE RT            UNUSUA L EVENT EMERGENCY                  EME RGENCY CGI Loss ofRPV            CS l  Loss of RPV    CA I Loss ofRPV          CU I UNPLANNED inventory affecting        inventory affecting    inventory.              loss of RPV inve ntory fue l clad integrity w ith core decay heat        Op. Modes: Cold          for 15 minutes or conta inment              removal capability. Shutdown, Refueling      longer.
cha llenged.              Op. Modes: Cold                                Op. Modes: Cold Op. Modes: Cold            Shutdown, Refueling                            Shutdown, Refueling Shutdown, Refueling CA2 Loss of all          CU2 Loss of all but offsite and all onsite  one AC power source AC power to              to essentia h: m~rgcncy esseru,Httc merge ncy    buses for 15 minutes or buses fo r 15 minutes or longer.
longer.                  Op. Modes: Cold Op. Modes: Cold          Shutdown, Refueling, Shutdown, Refueling,    De.fueled De.fueled CA3 Lnability to        c  3 u PLANN ED maintain the pl ant in  increase in RCS cold shutdown .          temperature .
Op. Modes: Cold          Op. Modes: Cold Shutdown, Refueling      Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling cus    Loss of all onsite or offsite communications capab ilities.
Op. Modes: Cold Shutdown, Refu eling, De.fueled CA6 Hazardous event affecting a SAFETY SYSTEM needed fo r the current operating mode.
Op. Modes: Cold Shutdown, Refueling 26
 
I I
I I
I CG1                                    I ECL: General Eme rge ncy Initiating Condition: Loss of RPY inventory affecting fue l clad integrity with containment                                            I challenged.
Operating Mode Ap plicability: Cold Shutdown, Refueling I
I Emergency Ac tion Levels: ( I or 2)
Note: The emergency director wi ll declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or wi ll likely be exceeded.
( I)    a.      RPY level less than -155" (T AF) for 30 minutes or longer.
AND
: b.      ANY indication from the Containment Challenge Table C I.
(2)    a.      RPV level cannot be monitored for 30 minutes or longer.
AN D
: b.      Core uncovery is indicated by ANY of the following:
* monitors:
[ Commented [JRB3] : RAJ 9 revision l nit I                                      l nit 2 1D2 l-K601 A      Rx llcad l anlt)\\n '\rea    2D21-K601 A    R:-. 1lead 1 av down Arca ID 11-K t>O 1 D  Rcfocl r loor                2D2l-K601 \1    Spent l*ud.lud Pool Areas ID2 l-K601 [ -    Drvw..:11 Shield Plug        ~D21-K601 F    Dr\criScparator Pl>Ol 1D21-K601 M      Spent lucl PlX>I and !'kw Fud 2D21-K611 K    RPV Refuel floor 228' Stor:w.e Arca                2D21-l\.611 I  Rl'V Rcfud l loor 228'
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Reactor Buildin Floor Drain Sum s Turbine Buildin Floor Drain Sum s Rad Waste Tanks Torus Room Sum s AND
: c.      ANY indication from the Containment Challenge Table C I.
27
 
Containment Challenge Table Cl l '\Pl \'\\;I D 111 c:hc 11 Primary Containment Pressure ~        ..
Containment 1-1 2 greater than or equal to 6% AND 02 greater than or eq ual to 5%
rr
( Commented [JRB4] : RAJ 11 revis ion Secondary CONTAINMENT INTEG RITY NOT established
* Secondary Containment radiation monitors greater than Max Safe values (SC EOP - Table 6)
* If Secondary CONTA INM ENT INTEGRJTY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Eme rgency is not required.
Basis :
CO TAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6. 1. 1. Secondary Containment OPERABLE per Technical Specification 3.6.4.1 UN PLANN ED: A parameter change or an eve nt that is not I) the res ult 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.
This IC addresses the inability to restore and maintain reactor vesse l level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substanti al core degradation or melting with potential fo r loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
Water level for top of active fuel is calculated at -158.44". Although slightly more conservative, the -155" EO P va lue for top of active fuel is provided for this EAL to ai d in operator recognition of the event.
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 vesse l level. If reactor vessel leve l cannot be restored, fuel damage is probable.
With Secondary CONTAINMENT !NTEGRJTY not established, there is a high potential fo r a direct and unmonitored release of radioactivity to the environment. If Secondary CO TA INMENT INTEGRJTY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emerge ncy is not required.
The existence of an explosive mi xture means, at a minimum, that the containment atmospheric hydrogen co ncentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen bum will raise containment pressure and could res ult in collateral equipment damage leading to a loss of containment integrity. It therefore represe nts a challenge to Containment integri ty.
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 possi ble to obtain a containment hydrogen gas concentration reading as ambient conditions wi thin 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.
In EAL 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total 28
 
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 actua lly occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for actions to terminate leakage, recover inventory control or makeup equipment, and/or restore level monitoring.
The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water leve l 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 agai nst other potential sources of water flow to ensure they are indicative of leakage from the RPV.
These EA Ls address concerns raised by Generic Letter 88- l 7, Loss of Decay Heat Removal; SECY 9 1-283, 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 NU MARC 9 1-06, Guidelines for Industry Actions to Assess Shutdown Management.
29
 
CS1 ECL: Site Area Emergency Initiatin g Condition: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability: Cold Shutdown, Refueling Emerge ncy Action Levels : ( I or 2 or 3)
No te: The emergency director will declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(I)    a.      Secondary CO TAINME T INTEGRITY not established.
AN D
: b.      RPV level less than - 41 " (6" below the Level 2 actuation setpoint).
(2)    a.      Secondary CONTAINMENT INTEGRITY established.
AN D
: b.      RPV level less than -155" (TAF).
(3)    a.      RPV level cannot be monitored for 30 minutes or longer.
AN D
: b.      Core uncovery is indicated by ANY of the fo llowing:
                *                                                                                      ( Commented [JRBS] : RAJ 9 revision l nit I                                        l nit 2 1D2 l -K60 I ,\ - R'< f lead Lavd1)\\ n *\n:a    2D11-f...601 :\  R:-. Head 1 a)'dO\\ n Arca 11DII-K601 D- Reli1dfloor                          2D21-h601    l\1 Spent Fuclll uel Pool Areas IID21-K601 l Drvwdl Shield Plug                    2D21-f...601 L Df\cr/Scparator Pool l021-K601 M - ',pent Fuel P1l0l and New Fud      2D1 l-K.611  K. RPV Rcfud floor 228'
                          '>torag.:: Ar.::a              2D21-f...611 l Rl'V Refud !Joor 228 '
* UNPLANNED level increase in any of the following of sufficie nt magnitude to indicate core uncovery:
Reactor Buildin Floor Drain Sum s Turbine Buildin Floor Drain Sum s Rad Waste Tanks Torus Room Sum s 30
 
Basis:
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1. Secondary Containment OPERABLE per Technical Specification 3.6.4.1.
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.
This IC addresses a significant and prolonged loss of RPV 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 warrant a Site Area Emergency declaration.
Water level for top of active fuel is calculated at-158.44". Although slightly more conservative, the -155" EOP value for top of active fuel is provided for this EAL to aid in operator recognition of the event.
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 reactor vessel level cannot be restored, fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying Secondary CONTAINMENT INTEGRITY following a loss of heat removal or RCS inventory control functions. The difference in the specified reactor vessel levels of EALs l .band 2.b reflects that with Secondary CONTAINMENT INTEGRITY established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, 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 actions to terminate leakage, recover inventory control or 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 ofleakage from the RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss ofDecay Heat Removal; SECY 91-283, Evaluation ofShutdown 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 uses IC CG 1 or RG 1.
31
 
CA1 ECL: Alert Initiating Condition: Loss ofRPV inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss ofRPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
(2)    a.      RPV level cannot be monitored for 15 minutes or longer AND
: b.      UNPLANNED level increase in any of the following due to a loss ofRPV inventory:
Drywell Floor Drain Sumps              Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps          Turbine Building Floor Drain Sumps Torus                                  Rad Waste Tanks Torus Room Sumps Basis:
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.
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 EAL #1, a lowering of water level below-35" (Level 2 actuation setpoint) 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.
Although related, EAL# 1 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 residual heat removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, 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 32
 
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 RPV.
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS I If the RPV inventory level continues to lower, then escalation to Site Area Emergency uses IC CS!.
33
 
CA2 ECL: Alert Initiating Condition: Loss of a ll offsite and all ons ite AC power to essenliak nu: rgcnc} buses fo r 15 minutes or longer.
Operating Mode A pplicability: Cold Shutdown, Refu eling, Defue led Emergency Action Levels :
Note: T he emergency director will declare the Alert promptly upon determ ining that 15 minutes has been exceeded, or will likely be exceeded.
( I)    Loss of ALL offsite and ALL onsite AC Power (Table'\ I) to 4160 VAC                          [ Commented [JRB6]: RAJ 12 revision E;,ssential bnergcnc} Buses l/2E, l /2F, AND 1/20 fo r 15 minutes or longer.
Table SI Unit I                                Unit2 Start-uo Aux XFMR IC                  Start-uo Aux XFMR 2C Start-uo Aux XFMR ID                  Start-up Aux XFM R 2D Diesel Generator IA                  Diesel Generator 2A Diesel Generator I B                  Diesel Ge nerator I B Diesel Generator IC                  Diesel Ge nerator 2C Basis:
T hi s IC addresses a tota l loss of AC power (see Table S I above) that co mpromises the performance of all SA FETY SYSTEMS requiring electric power, including those necessary fo r essential core coo li ng, containment heat removal/pressure control, spent fuel heat removal, and the ult imate heat sink.
When in the cold shutdown, refueling, or defueled mode, thi s condi tion is not class ified as a Site A rea Emergency because of the increased time available to restore an essentialem.::rgency bus to service. Additional time is ava ilable due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. When in these modes, this condition represe nts an actual or pote ntial substantial degradati on of the level of plant safety.
Fiftee n minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classi fi cation level uses IC CS I or RS I.
34
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(1)      UNPLANNED increase in RCS temperature to greater than 212 &deg;F for greater than the duration specified in Table C2.
Table C2: RCS Heat-up Duration Thresholds Secondary CONTAINMENT RCS Status                                                    Heat-up Duration INTEGRITY Status Not Established                  0 minutes*
Not Intact Established                  20 minutes Intact                        Not applicable                  60 minutes*
* If RHR is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)      UNPLANNED RCS pressure increase greater than 10 psig.
Basis:
CONTAINMENT INTEGRJTY: Primary Containment OPERABLE per Technical Specification 3.6. l. l. Secondary Containment OPERABLE per Technical Specification 3.6.4.1.
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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact, and Secondary CONTAINMENT INTEGRITY is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because l) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
35
 
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when Secondary CONTAINMENT INTEGRITY is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, the RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of Secondary CONTAINMENT INTEGRITY 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.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS I or RS I.
36
 
CA6 ECL: Alert Ini tiating Co nd ition : Hazardo us event affecting a SAFETY SYSTEM needed fo r the current operating mode.
O per atin g Mode A pplicability: Cold Shutdown, Refueling E merge ncy Actio n Levels:
( I)    a.      The occurrence of A NY of the fo llowing hazardous events :
* Se ismic event (earthquake)
* Interna l or external fl ooding event
* High winds ( 35 mph sustain~d) or tornado stri ke                              [ Commented [JRB7] : RAJ 13 a rev1Sion
* FIRE
* EXP LOS ION
* Other events with similar hazard characteri stics as determined by the Shi ft Manager AND
: b.      EITHER of the fo llowing :
* Eve nt damage has ca used indications of degraded perform ance in at least one train o f a SAFETY SYSTEM needed fo r the current operating mode .
* The event has caused VI SIB LE DAMAG E to a SA FETY SY TEM compo nent or structure needed for the current operating mode .
Bas is :
EXP LOS ION: A rapid, violent and catastrophic failure of a piece of equipment due to co mbustion, chemi cal reactio n o r overpressurization. A release of steam (from hi gh energy lines or components) or an electrical component fa ilure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present. FIRE : Combustion characterized by heat and lig ht. Sources o f smoke such as slipping dri ve belts or overheated electrical equipment do not constitute FI RES . Observation of fl ame is preferred but is OT required if large q uantities of smoke and heat are observed.
SAFETY SYST EM: A syste m 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 class ified as safety-related.
VISIBL E DAMAG E: Damage to a component or structure that is readily observable without meas urements, testing, o r analys is. The visual impact of the damage is suffic ient to cause concern regarding the ope rab ili ty o r reliability of the affected component or structure.
T hi s IC addresses a hazardous eve nt that causes damage to a SA FETY SYSTEM, or a structure containing SA FETY SYST EM components, needed for the current operating mode. T his 37
 
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 plant safety.
EAL I.a identifies hazardous events that could result in damage to plant systems. A seismic event is indicated by entry into IC HU2. Flooding is indicated by a significant increase in water levels (external or internal). High winds are indicated by sustained winds at the site meteorological tower exceeding 35 mph.
The first threshold for EAL l.b 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 threshold for EAL 1. b 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 all 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 uses IC CS 1 or RS 1.
38
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss ofRPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    UNPLANNED foss of reactor coolant results in RPV level less than the lower limit of the controlling level band for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED level increase in any of the following:
Drywell Floor Drain Sumos                Reactor Building Floor Drain Sumos Drywell Ecmioment Drain Sumos            Turbine Building Floor Drain Sumps Torus                                    Rad Waste Tanks Torus Room Sumos                            ' ,.
Basis:
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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL #1 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 ifthe 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.
39
 
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.
EAL #2 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 RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CAI or CA3.
40
 
CU2 EC L: Notification of Unusua l Event Initiating Co ndition: Loss of a ll but one AC power source to esseRtialc mergc nc) buses for 15 minutes or longer.
Operating Mode Applicabili ty: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The emergency director wi ll declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( 1)    a.      AC power capabi li ty to 4 160 VAC EsseRtial l:-.merge nc~ Buses l/2E, 1/2F, AND l/2G is reduced to a si ngle power source (l able SI) for 15 minutes or longer.      { Commented [JRBS]: RAJ 13 b rev1S1on AND
: b.      A ny additional s ing le power sou rce failure will result in loss of all AC power to SAFETY SYSTEMS.
Table SI Unit 1                            Unit 2 Start-up Aux XFMR IC              Start-up Aux XFMR 2C Start-up Aux XFMR ID              Start-up Aux XFMR 2D Diesel Generator I A              Diesel Generator 2A Diesel Generator I B              Diesel Generator I B Diesel Generator IC                Diesel Generator 2C Basis:
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 typ ically systems class ified as safety-related.
This IC describes a sig nifica nt degradation of offsite and ons ite AC power sources such that any additional single fa ilure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source (see Table S I above) may be powering one, or more than one, train of safety-related eq uipme nt.
When in the cold shutdown, refueling, or defueled mode, this condi tion is not classified as an Alert because of the increased time avai lable to restore another power source to se rvice.
Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. When in these modes, this condition is cons idered to be a potential degradation of the leve l of plant safety.
A n " AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an esseRtialemergcncy bus. Examp les of this conditi on incl ude:
A loss of all offsite power with a concurrent fa ilure of all but one esseRtialemergcncy power 41
 
source (e.g., an onsite diesel generator).
A loss of all offsite power and loss of all esseAtialcmergency power sources (e.g., onsi te di esel generators) with a sing le train of esseAtialcrnergcnc)' buses being back-fed from the uni t main generator.
A loss of esseAtialcmcrgcnC)' power sources (e.g., onsite diesel generators) with a sing le train of esseAtiak mcrgc11l::) buses being back-fed from an offsite power source.
Fifteen minutes is the threshold to exclude transient or momentary losses of power.
The subsequent loss of the remaining si ng le power source would escalate the eve nt to an A lert in accordance with IC CA2 .
42
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      UNPLANNED increase in RCS temperature to greater than 212 &deg;F.
(2)      Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
Basis:
UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event niay be known or unknown.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of plant safety. If the RCS is not intact and secondary CONTAINMENT INTEGRITY is not established during this event, the emergency director will 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.
EAL #I 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, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions, and operators are 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 is a threshold to exclude transient or momentary losses of indication.
43
 
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
44
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105/210 VDC on Technical Specification required 125/250 VDC buses l/2R22-S016 AND l/2R22-SOl 7 for 15 minutes or longer.
Basis:
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. This condition is considered to be a potential degradation of the level of plant safety.
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 Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
45
 
CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
Plant telephones (Includes hardwired and wireless)
Plant page Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 theNRC.
This IC will 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).
EAL # 1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OR Os referred to here are the State of Georgia, Appling County, Jeff Davis County, Tattnall County and Toombs County.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
46
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HUl Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 47
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table EI.
Table El Location of Dose Rate                Total Dose Rate (Neutron+ Gamma mR/hr)
HI-TRAC 125 Side - Mid- height                    450 Top                              110 HI-STAR 100 or HI-STORM 100 Side - 60 inches below 80 mid- height Side - Mid- height                        80 Side - 60 inches above 30 mid- height Center oflid                          10 Middle of top lid                      20 Top (outlet) duct                    40 Bottom (inlet) duct                    140 Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY ofa 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 that could cause challenges in removing the cask or fuel fro~ storage.
The existence of"damage" is determined by radiological survey. The radiation reading values listed in the table represent 2 times the site-specific cask-specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical 48
 
specification multiple of "2 times" 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, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under !Cs HUI and HAI.
49
 
POTENTIAL            POTENTIAL 6 FISSION PRODUCT BARRIER ICS/EALS                        LOSS POTENTIAL LOSS    LOSS LOSS LOSS LOSS FUELCtAO                RCS              CONTAINMENT Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.                                                          - YES-      ID -1.uss of ANY T\10 llarri<:rs dlfil l.o~s or Put.:nliil L11Ss ofThinl llarri~T FGl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown SITE AREA EMERGENCY POTENTIAL            POTENTIAL  LOSS    POTENTL'IL    - NOl Loss or Potential Loss of any two barriers. LOSS LOSS LOSS LOSS                LOSS FUELCtAD                RCS              CONTAINMENT FSl    Op. Modes: Power Operation, Hot Standby,            ~--------+------~
                                                                                                    ~-~--~
Startup, Hot Shutdown
                                                                                ?~'---------L_-Sl-*l<>_"_~_*"_'""_"_ll.o~ssufANYT11ollmiQ'S ALERT Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
FAl                                                          POTENTIAL            POTENTIAL LOSS                LOSS Op. Modes: Power Operation, Hot Standby,                LOSS                LOSS FUEL CLAD              RCS Startup, Hot Shutdown 50
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY                            FSl SITE AREA EMERGENCY                                  FAlALERT Loss of any two barriers and Loss or            Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either the Potential Loss of the third barrier.                                                          Fuel Clad or RCS barrier.
Fuel Clad Barrier                                  RCS Barrier                                  Containment Barrier LOSS              POTENTIAL                LOSS              POTENTIAL LOSS                    LOSS                POTENTIAL LOSS                                                                                                LOSS
: 1. RCS Activity                              1. Primary Containment Pressure                  1. Primary Containment Conditions A. Activity of 300    Not Applicable        A. Primary              Not Applicable            A. UNPLANNED rapid          A. Primary
  &#xb5;Ci/gm DEirn                                containment                                        drop in primary              containment pressure greater                                  containment pressure          pressure than 1.85 psig due                                following primary            greater than 56 to RCS leakage.                                    containment pressure          psig rise                        OR OR                      B. Greater than or B. Primary containment            equal to 6% H1 pressure response not        AND 5%02 consistent with LOCA          exists inside conditions.                  primary containment OR
: c. HCTL exceeded.
: 2. RPV Water Level                          2. RPV Water Level                                2. RPV Water Level A. SAG entry is        A. RPV water level    A. RPV water level      Not Applicable            Not Applicable              A. SAG entry is required.              cannot be            cannot be restored                                                              required.
restored and        and maintained maintained          above -155 inches above -155          or cannot be inches or cannot    determined.
be determined.
51
 
Fuel Clad Barrier                        RCS Barrier                                  Containment Barrier LOSS            POTENTIAL          LOSS            POTENTIAL LOSS                    LOSS                POTENTIAL LOSS                                                                                        LOSS
: 3. Not Applicable                  3. RCS Leak Rate                                3. Primary Containment Isolation Failure Not Applicable      Not Applicable A. UNISOLABLE        A. UN ISO LAB LE        A. UNISOLABLE direct        Not Applicable break in Main          primary system          downstream pathway Steamline, HPCI,        leakage that results    to the environment Feed water,            in exceeding            exists after primary RWCU, or RCIC          EITHER of the          containment isolation OR                      following:              signal B. Emergency RPV        I. Max Normal            OR Depressurization.          Operating        B. Intentional primary Temperature          containment venting OR                  per EOPs
: 2. Max Normal              OR Operating Area    C. UNISOLABLE Radiation Level.      primary system leakage that results in exceeding EITHER of the following:
I. Max Safe Operating Temperature.
OR
: 2. Max Safe Operating Area Radiation Level.
: 4. Primary Containment Radiation    4. Primary Containment Radiation                4. Primary Containment Radiation A. DWRRM            Not Applicable A. DWRRM greater    Not Applicable            Not Applicable              A. DWRRM greater greater than                        than 40 R/hr.                                                              than 26,000 R/hr.
1,400 R/h.
52
 
Fuel Clad Barrier                                  RCS Barrier                                  Containment Barrier LOSS            POTE NTIAL                  LOSS            POTENTIAL LOSS                      LOSS                POTENTIAL LOSS                                                                                                LOSS
: 5. Other Indications                        5. O t her Indications                            5. Other Indications A. Qi:fgas PFe BflEI  Not Applicable.        A. Drywe ll Fission    Not Applicable.            Not Applicable.            Not Applicable.
Pest +FeatmeA&#xa3;                              Product Monitor MeAiteFs                                    readi ng 5.0 x 10 5
  ~                                            cpm.
Higft ot Applicable.                                                                                                                                    [ Commented [JRB9): RAJ J e revision
: 6. Emergency Director Jud gment              6. Emergency Director Judgment                    6. Emergency Director Judgment A. ANY condition A. AN Y condition A. ANY conditi on in            A. ANY condition in the    A. ANY condition in the    A. ANY conditi on in the opinion of    in the opini on        the opinion of the    opinion of the            opinion of the              in the opini on of the emergency        of the                emergency              emerge ncy d irector      emerge ncy director        the emergency di rector that        emergency              director that          that indicates            th at indi cates loss of    director that indi cates loss of    director that          indicates loss of      potenti a l loss of the    the Conta inment            indicates the fue 1 clad        indicates              the RCS Barrier.      RCS Barrier.              Barrier.                    pote nti al loss of barrier.              poten ti al loss of                                                                                  the Containment the fue l clad                                                                                      Barrier.
barrier.
53
 
Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The fuel clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets.
: l.      RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;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 percent 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.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no potential loss threshold associated with RCS Activity.
: 2.      RPV Water Level Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.
Potential Loss 2.A 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 2.A. 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.
This threshold is considered 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 54
 
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 ofRPV 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 A TWS/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 SA5 or SS5 will dictate the need for emergency classification.
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.
: 3. Not Applicable (included for numbering consistency between barrier tables)
: 4. Primary Containment Radiation Loss 4.A 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 &#xb5;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 percent 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.
: 5. Other Indications 55
 
ReadiAgs fFe1T1 Offgas pre EtAd pest treatffieAt !fleRiters that iAElieale Offseale High are H5ee te deteet the effi1:1eAt efthe Offgas syste!fl EtAd tl~erefere iRElieate fissieA pred1:1ets eseapieg the elae. Cale1:1latee reaEliAgs ffir 3()() t:tCi.lgm are 4.82 11 e+8 eps BREI tile iRstr1:1meAts te ef seale is I e+G eps. These iRstr1:1n1eAts geiAg high effseale previee aA iAElieatiee that there is elae damage le aie iA elassifieatieA efBR e>>eAt. S0TAple results are still Aeeeeel to estaelish that the 3QQ t:tCi/gffi threshelel is eeiAg e1*eeeeee.There is Ae peteAtial less threshele asseeiated with Other IAElieatiens.                                    Commented [JRB ~ c revision IL    1)k Commented [JRB11] : RAI l e revm on
: 6. Emergency Director Jud gment Loss 6.A This thres ho ld addresses any other fac tors that used by the e me rgency director in determining whethe r the fuel clad barrier is lost.
Potenti al Loss 6.A This thresho ld addresses any othe r facto rs that may be used by the emerge ncy d irector in determining whether the Fue l Clad Barrier is pote ntia lly lost. T he emergency director will also conside r whether or not to declare the barrie r pote ntia lly lost in the event that barrie r status cannot be monitored.
56
 
RCS BARRIER THRESHOLDS:
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.
: 1.      Primary Containment Pressure Loss I.A The greater than 1.85 psig primary containment pressure 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.
: 2.      RPV Water Level Loss 2.A 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 2.A. 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 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 injection sources to restore RPV water level or 2) no low pressure RPV iajection 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 ofRPV 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 beyond the limit. The threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
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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, !Cs SA5 or SS5 will dictate the need for emergency classification.
There is no RCS potential loss threshold associated with RPV water level.
: 3. RCS Leak Rate Loss Threshold 3.A 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.
Loss Threshold 3.8 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). 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.
Potential Loss Threshold 3.A 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 that 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.
: 4. Primary Containment Radiation Loss 4.A 58
 
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.
: 5. Other Indications A Drywell Fission Products Monitor reading 5.0 x 10 5 cpm indicates a breach of the RCS as an effluent. The monitor value calculated in Calculation SMNH-13-021, Rev 1, was 1.008 x 106 cpm; however, the top of the scale for the monitor is 1 x 106 cpm. Therefore, the EAL threshold value has been established at one half decade below top of scale to aid the operator in distinguishing between a loss of RCS event and an instrument failure resulting in the monitor reading high off scale. No radiation monitors capable of indicating a potential loss of the RCS barrier were identified.
There is no Potential Loss Threshold associated with Other Indications.
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
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CONTAINMENT BARRIER THRESHOLDS:
The primary containment barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      Primary Containment Conditions Loss LA and l.B 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. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
These thresholds rely 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.
Potential Loss I .A 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 represents a potential loss of the containment barrier.
Potential Loss l .B 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.
Potential Loss l.C The heat capacity temperature limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise:
Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
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The HCTL is a function of RPV pressure, suppression pool temperature and suppress ion pool wate r level. It is used to prec lude fa ilure of the containment and equ ipment in the containment necessary fo r the safe shutdown of the plant and therefore, the inabili ty to maintain plant paramete rs be low the limit constitutes a potential loss of co nta inment.
: 2. RPV Water Level There is no loss threshold associated with RPV water leve l.
Potential Loss 2.A The potential loss thres ho ld is identi cal to the Fuel Clad Loss RPV Water Level thres hold 2.A. The potenti al loss requirement fo r entry into the Severe Accident Guide lines indicates adequate core cooling cannot be assured and that core damage is poss ible. BWR EPGs/SAGs specify the conditions when the EPGs are ex ited and SAGs are entered.
Entry into SAGs is a logical escalation in response to the inability to ass ure adequate core cooling .
PRA studies indicate that the cond ition of this pote ntial loss threshold is a core melt sequence that, if not corrected, co uld lead to RPV failure and increased potential for primary containment fa ilure. In conjuncti on with the RPV water level loss thresholds in the fuel clad and RCS barri er columns, this thres hold results in the declaration of a General Emergency.
: 3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation (automatic or manual) that allows an UN ISOLABLE direct release to the environment. A release path is ' direct' if it allows for the migration o f radioacti ve material from the containment to the environment in a generally uninterrupted manner (e.g., litt le or no holdup time); therefore, within the contexi of a containment barri er loss or potential loss threshold, a release path te-through ( Commented [JRB12] : RAJ 16 revision the wetwe ll is a direct release path.
Loss 3.A 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 Contai nment Isolat ion System (PCIS) .
The existence of a filter is not co nsidered in the threshold assessment. Filters do not remove fission product noble gases . In addition, a filter could become ineffecti ve due to iodine and/or parti cul ate loading beyond design limits (i.e., retention abili ty has been exceeded) or water saturation from steam/high humidi ty in the release stream.
Following the leakage of RCS mass into primary containment and a rise in primary containment press ure, there may be minor radiological releases assoc iated with allowable primary containment leakage through various pe netrations or system components. Minor releases may also occur if a primary containment isolation va lve(s) fa ils to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potenti a l loss of primary containment but should be evaluated using the Recogniti on Category R ICs .
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Loss 3.B 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 will also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas 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 pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.
Loss 3.C 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 shutdown of the plant will be precluded. EOPs use these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
The temperatures and radiation levels will 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 that connect directly to the RPV ensuring 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 3.A this threshold would result in a Site Area Emergency.
There is no potential loss threshold associated with Primary Containment Isolation Failure.
: 4. Primary Containment Radiation There is no loss threshold associated with primary containment radiation.
Potential Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20-percent 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-percent 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 containment that would then escalate the emergency classification level to a General Emergency.
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: 5. Other Indications Not Applicable (included for numbering consistency between barrier tables)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the emergency director in determining whether the containment barrier is lost.
Potential Loss 6.A This threshold addresses any other factors that may be used by the emergency director in determining whether the containment barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
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7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                EMERGENCY HGl HOSTILE              HSl HOSTILE              HAl HOSTILE            Hut Confirmed ACTION resulting in      ACTION within the        ACTION within the      SECURITY loss of physical        PROTECTED AREA.          OWNER                  CONDITION or control of the facility. Op. Modes: All          CONTROLLED              threat.
Op. Modes: All                                    AREA or airborne        Op. Modes: All attack threat within 30 minutes.
Op. Modes: All HU2 Seismic event greater than OBE levels.
Op. Modes: All HU3 Hazardous event.
Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HAS Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All HS6 Inability to        HA6 Control Room control a key safety    evacuation resulting in function from outside    transfer of plant the Control Room.        control to alternate Op. Modes: All          locations.
Op. Modes: All 64
 
GENERAL              SITE AREA ALERT          UNUSUAL EVENT EMERGENCY              EMERGENCY HG7 Other              HS7 Other                HA7 Other              HU7 Other conditions exist which conditions exist which    conditions exist which conditions exist in the judgment of the in the judgment of the    in the judgment of the which in the emergency director    emergency director        emergency director    judgment of the warrant declaration of warrant declaration of    warrant declaration of emergency director a General Emergency. a Site Area              an Alert.              warrant declaration of Op. Modes: All        Emergency.                Op. Modes: All        a (NO)UE.
Op. Modes: All                                  Op. Modes: All 65
 
HG1 ECL : General Emergency Initiating Condition : HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability: All Emergency Action Levels:
(I)      a.      A HOSTILE ACT!O is occurri ng or has occurred within the PROTECTED AREA (PA) as reported by the 1            Security Shtft-.Captain or designee.      [ Commented [JRB13]: RAJ 21 .a revision AND
: b.      EITHER of the fol lowing has occurred:
I.      ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* RPV wate r level
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMINENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) 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 civi l disobedience or felon ious acts that are not part of a concerted attack on the NPP. Non-te rrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner contro lled area (OCA)).
IMMIN ENT: The trajectory of events or conditions is such that an EAL wi ll be met within a relatively short period of time regardless of mitigation or corrective actions.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence .
This IC addresses an event in which a HOSTILE FORCE has taken physical control of the faci lity to the extent that the plant staff can no longer operate equipment necessary to maintain key safety funct ions. It also addresses a HOSTILE ACTION leading to a loss of physical contro l that results in actual or IMMINENT damage to spent fuel due to I) damage to a spent fue l pool coo ling system (e .g. , pumps, heat exchangers, controls) or 2) loss of spent fuel pool integrity so that sufficient water level cannot be maintained.
66
 
T imely and accurate communications between Security shi ft supervision and the control room is essenti al for proper class ification of a security-related event.
Security plans and terminology are based on the guidance prov ided by N EI 03-1 2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate security-sensitive in fo rm ation. This includes info rmati on that may be advantageous to a potenti al adversary, such as the particulars co ncerning a specific threat or threat location. Securi ty-se nsitive info rm ation should be conta ined in non-publi c documents such as the Security Plan.
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HG7 ECL: General Emergency In itiating Condition: O ther conditions ex ist whi ch in the j udgment of the emergency director warrant declaration of a General Emergency.
Operating Mode A pplicab ili ty: All Emergency Actio n Levels:
( I)      Other conditions ex ist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMIN ENT substantial core degradation or melting wi th potential for loss of containment integrity or HOSTIL E ACTION that res ults in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protecti ve Action Guideline expos ure leve ls offsite for more than the immedi ate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (N PP) or its personnel that includes the u e of violent fo rce to destroy equipment, take HOSTAGE , and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJ ECTILEs, vehicles, or other dev ices used to de live r destructive force. Other acts that sati sfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the N PP. Non-terrorism-based EA Ls sho uld be used to address such activities (i.e., this may include violent acts between indi viduals in the owner controlled area (OCA)).
IMMIN ENT : The trajectory of events or conditions is such that an EA L will be met within a relative ly short period of time regardless of miti gation or correcti ve actions.
T hi s IC addresses unanticipated conditions not addressed expl ic itly e lsewhere but that warrant declaration of an emerge ncy because conditions exist that are believed by the emerge ncy director to fa ll under the emergency class ification level description fo r a General Emerge ncy.
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HS1 ECL: Site A rea Emergency Initiating Condition: HOSTILE ACTIO                within the PROTECTED AREA.
Operating Mode Applicabi lity: A ll Emergency Ac tio n Leve ls:
( I)    A HOSTIL E ACTI ON is occurring or has occurred within the PROT ECTE D AREA (PA) as reported by the            Security Sfillt.-Captain or des ignee.                    ( Commented [JRB14]: RAJ 2 1 a revision Basis:
HOSTIL E ACTION: An act toward a nu clear powe r plant (N PP) or its personnel that includes the use of violent fo rce to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. T hi s includes attack by air, land, or wate r using guns, explos ives, PROJ ECTI LES, vehi cles, or other devices used to deliver destructive fo rce. Other acts that satisfy the overall intent may be included. HOSTI LE ACTIO should not be construed to include acts of civ il disobedience or felonious acts that are not part of a concerted attack on the N PP. Non-terrorism-based EA Ls should be used to address such acti vities (i.e., thi s may include violent acts betwee n indi viduals in the owner controlled area (OCA)).
PROT ECTED AREA (PA): The area that encompasses all contro lled areas within the security protected area fence.
T hi s IC addresses the occurrence of a HOSTILE ACTION w ithin the PROTECTED AREA.
T hi s event will require rapid response and ass istance due to the possibil ity fo r damage to plant equipment.
T ime ly and accurate commun ications between Security shift supervision and the control roo m is essential fo r proper class ification of a sec urity-re lated eve nt.
Sec urity plans and term inology are based on the guidance prov ided by NE I 03-1 2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independe nt Spent Fuel Storage Installation Security Program}.
As time and condi tions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective meas ures (e.g., evacuation, di spe rsal, or sheltering).
T he Site A rea Emergency declaration will mobilize ORO resources and have them avail able to develop and implement publ ic protective actions in the unli ke ly event that the attack is successful in impairing multipl e safety fun ctions.
T hi s IC does not apply to a HOSTIL E ACT ION directed at an ISFS I PROTECT ED AREA located outside the plant PROTECTED A REA; such an attack should be assessed using IC HA I.
It also does not apply to incidents that are accidental eve nts, acts of civil di sobedi ence, or otherwise are not a HOSTILE ACTI O perpetrated by a HOSTILE FO RCE. Examples include the cras h o f a small aircraft, shots fro m hunters, or phys ical disputes between employees.
Repo rting of these types of events is adequ ately addressed by other EA Ls, or the requirements of IO CF R &sect; 73.7 1 or IO CF R &sect; 50.72.
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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 Security Plan.
Escalation of the emergency classification level uses IC HG 1.
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HSG ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    a.      An event has resulted in plant control being transferred from the control room to remote shutdown panels.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* RPV water level
* RCS heat removal Basis:
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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location.
Escalation of the emergency classification level uses IC FG 1 or CG 1.
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HS7 ECL: Site A rea Emergency Ini tiati ng Condition: Other cond itions ex ist whi ch in the judgment of the emergency director warrant declaration of a Site Area Emergency.
Operating Mode Applicabi lity: A ll Emergency Action Leve ls:
( I)      Other conditions ex ist which in the judgment of the emergency director indicate that events are in progress o r have occurred which invo lve actual or likely major failures of plant functions needed for protection of the public or HOSTIL E ACTION that results in intentional damage or malicious acts, I) toward site personnel or equipment that could lead to the likely fa ilure of or, 2) that prevent effective access to equipment needed for the protection of the public. Any re leases are not expected to re ult in expos ure levels which exceed EPA Protective Action Guideline exposure leve ls beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personne l that includes the use of violent force to destroy equipment, take HOSTAG ES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJ ECTILEs, vehicles, or other dev ices used to de live r destructive force . Other acts that sati sfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil di sobedi ence or fe lonious acts that are not part of a concerted attack on the NPP. on-terrorism-based EALs should be used to address such acti vities (i.e., this may include violent acts between indi viduals in the owner controlled area (OCA)).
This IC add resses unanticipated conditions not addressed exp li citly e lsewhere but that warrant declarat ion of an emergency because conditions exist that are believed by the eme rgency director to fall under the emergency classification level description for a Site Area Emergency.
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HA1 ECL: Alert Ini tiati ng Co ndition: HOSTILE ACT ION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Ope rati ng Mode Ap plicabili ty : A ll Emerge ncy Ac tion Levels: ( I or 2)
( I)      A HOSTIL E ACTIO is occurring or has occurred within the OWNER CONTROLLED AREA (OCA) as reported by the          I I Security Sfli.l+.Captain or designee.            { Commented [lRBl S]: RAJ 21 a rev1S1on (2)      A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
Basis:
HOSTILE ACTION : An act toward a nuclear power plant (N PP) or its personne l that includes the use of vio lent fo rce to destroy equipment, take HOST AGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJ ECTILES, ve hicles, or other device used to deli ver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedi ence or felonious acts that are not part of a concerted attack on the N PP. 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 (OCA)).
OWNER CONTRO LLED AREA (OCA): The site property owned by or otherwise under the control of HNP Security This IC addresses the occurrence ofa HOSTILE ACTION within the OWNER CONTROLL ED AREA or notifi cation of an aircraft attack threat. This event will require rapid res ponse and assistance due to the possibility of the attack progress ing to the PROTECTE D AREA (PA), or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications betwee n Security shift supervision and the control room is esse ntial for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided by El 03- 12. Template for the Sec11rity Plan, Training and Qualification Plan, Safeg11ards Contingency Plan [and Independent Spent F11el Storage Installation Sec11rity Program}.
As time and conditions allow, these events require a he ightened state of readiness by the plant staff and implementation ofonsite protecti ve measures (e.g., evacuation, di spersal, or sheltering).
The Alert declaration will also heighten the awareness ofoffsite response organizations, allowing them to be better prepared should it be necessary to consider further act ions.
This IC does not apply to incidents that are acc idental events, acts of civil disobedience, or otherwise are not a HOSTILE ACT ION perpetrated by a HO TILE FORCE. Examples include the crash ofa small ai rcraft, shots from hunters, or phys ical di sputes between employees.
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Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71or10 CFR &sect; 50.72.
EAL# l is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
EAL #2 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 station procedures.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe 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 0 WNER 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 Security Plan.
Escalation of the emergency classification level uses IC HS 1.
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HA5 ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(I)    a.      Release ofa toxic, corrosive, asphyxiant or flammable gas into any Table HI plant rooms or areas:
Table HI Building                        Rooms                  Applicable Modes Diesel generator building  All                                        All Unit 1/2 130'                              All Reactor building            Unit 1/2 SE Diagonals (RHR)                All Unit 1/2 NE Diagonals (RHR)                All AND
: b.      Entry into the room or area is prohibited or impeded.
Basis:
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 plant safety.
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 director's 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 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.
75
* 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 I 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 4.
e  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.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19-percent, 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 uses Recognition Category R, C or F !Cs.
76
 
HAS ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(!)    An event has resulted in plant control being transferred from the control room to remote shutdown panels.
Basis:
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 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 uses IC HS6.
77
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
( 1)    Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for an Alert.
78
 
HU1 ECL: Notification of Unusual Event Initiatin g Condition: Confirmed SECURlTY CO DITIO              or threat.
Operating Mode Applicability: All Emergency Action Levels: ( I or 2 or 3)
(I)      A SECURlTY CONDITION that does not involve a HOSTILE ACTION as reported by the        Security &htlt-Captain or designee.                                            ( Commented [JRB16]: RAJ 21 a revis ion (2)      Notification of a credible security threat directed at HNP.
(3)      A validated notification from the NRC providing information of an aircraft threat.
Basis:
SECURlTY CONDITION: Any Security Event as listed in the approved security contingency plan that constitutes a threat/co mpromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURlTY COND ITION does not involve a HOSTILE ACTION.
HOSTILE ACTION: An act toward a nuclear power plant (NPP) 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 ACTIO should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the PP. on-terrorism-based EA Ls should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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 that do not meet one of these EA Ls are adequately addressed by the requirements of I 0 CFR &sect; 73.71 or I 0 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIO Sare clas ifiable under ICs HA I, HSlandHGI.
Timely and accurate communications between Security shift supervision 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 OROs.
Security plans and terminology are based on the guidance provided by El 03-1 2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
EAL# I references the Security Shift Captain or designee because these are the indi viduals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is co ntroll ed due to the nature of safeguards and 10 CFR &sect; 2.39 information.
79
 
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with station procedures.
EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HAI.
80
 
HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels Operating Mode Applicability: All Emergency Action Levels:
(1)      Seismic event greater than Operating Basis Earthquake (OBE) as indicated by ANY of the following
* Unit One "Seismic Peak Shock Recorder High G Level" (657-066) alarm
* Unit Two "Seismic Instrumentation Triggered" (657-048) alarm
* A 12.7 Hz amber light illuminated in the N/S OR E/W column on panel IHI l-P701
* A 12.7 Hz red light illuminated in the N/S OR E/W column on panel IHI l-P701 Basis:
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 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 plant safety.
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., typical lateral accelerations are in excess of0.08g). The Shift Manager or emergency director may seek external verification if deemed appropriate (e.g., a call to the USGS or check internet news sources); 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 uses IC CA6 or SA9.
81
 
HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3 or 4 or 5)
Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
(I)    A tornado strike within the PROTECTED AREA (PA).
(2)    Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
(3)    Movement of personnel within the PROTECTED AREA (PA) is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
(4)    A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site in personal vehicles.
(5)      Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours.
Basis:
PROTECTED AREA (PA): The area which encompasses all controlled areas within the security protected area fence.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of plant safety.
EAL #I addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 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 ifthe water level or related wetting causes an automatic isolation ofa 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.
EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA (PA).
82
 
EAL #4 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, or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses the phenomena of the hurricane based on the severe weather mitigation procedure.
Escalation of the emergency classification level is based on !Cs in Recognition Categories A, F, Sor C.
83
 
HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Note: The emergency director will declare the Unusual Event promptly upon determining that          I the applicable time has been exceeded, or will likely be exceeded.
(1)    a.      A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications
* Report from the field (i.e., visual observation)
* Receipt of multiple (more than I) fire alarms or indications
* Field verification of a single fire alarm AND
: b.      The FIRE is located within ANY Table H2 rooms or areas (2)    a.      Receipt of a single fire alarm (i.e., no other indications of a FIRE)
AND
: b.      The FIRE is located within ANY Table H2 rooms or areas AND
: c.      The existence ofa FIRE is not verified within 30-minutes of alarm receipt.
(3)    A FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
(4)    A FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H2 Buildin                                    Rooms CB 147' Cable Spreading Room Control Building U 112 CB 112' Station Battery Rooms A,B 11 Diesel generator build_in~1g"'-lr-AA_1,,..1 _ _ _ _ _ _ _ _ _ _ _ _--1 Primary containment_--1r-U--,ni-t_ _ _ ..,... , . . . , - - - - - - - - - - - 1 112 130 Unit 1/2 SE Diagonals (RHR)
Unit 1/2 NE Diagonals (RHR)
Reactor building Unit I SW Diagonals (RCIC)
Unit 2 NW Diagonals (RCIC)
Unit 1/2 HPCI Rooms Intake structure                    All 84
 
Basis:
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 (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of plant safety.
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 or automatic activation of a suppression system.
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.
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 ofa 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 EAL# I is immediately applicable, and the emergency must be declared ifthe 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.
In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA (PA) not extinguished within 60 minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA (PA).
85
 
!fa FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency, 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 not necessary ifthe agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
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."
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.
Appendix R to 10 CFR 50, requires, among other considerations, the use of I-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 EAL #2, the 30 minutes to verify a single alarm is well within this worst-case I-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level uses IC CA6 or SA9.
86
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Notification of Unusual Event (NOUE).
Operating Mode Applicability: All Emergency Action Levels:
( 1)    Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
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 NOUE.
87
 
8      SYSTEM MALFUNCTION ICS/ EALS G E E RAL                SITE A R EA EM E RG ENCY              EMERGE NC Y AL ERT                UN    s  A L EVE 'T SGI Prolonged loss of      SSI    Loss of all offsite  SAi      Loss of all but one s  I Loss of all offs ite all offsite and all onsite and all onsite AC powe r      AC power source to          AC power capability to AC power to                to esseAtia k m-:rgcnc)      esseffi-iakmcrgcnc) buses    esseAtia k mcrgcnc~ buses essef!t.ia.k me rgc nC)    buses for 15 minutes or        fo r 15 minutes or longer. fo r 15 minutes or longer.
buses.                    longer.                      Op. Modes: Power            Op. Modes: Power Op. Modes: Power          Op. Modes: Power              Operation, Startup, Hot      Operation, Startup, Hot Operation, Startup, Hot    Operation, Startup, Hot      Standby, Hot Shutdown        Standby, Hot Shutdown Standby, Hot Shutdown      Standby, Hot Shutdown SA2 UN PLANNED              SU2 UN PLANNE D loss of Contro l Room        loss of Control Room ind ications for 15          indi cations for 15 minutes or longer with a    minutes or longer.
sign ificant transient in    Op. Modes: Power progress.                    Operation, Startup, Hot Op. Modes: Power            Standby, Hot Shutdown Operation, Startup, Hot Standby, Hot Shutdown SU3 Reactor coo lant activity greater than Technical Specificati on allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS    In abi lity to        SAS Auto matic or            SUS Automati c or shutdown the reactor          manual scram fa ils to      manual scram fai ls to causing a challenge to        shutdown the reactor, and    shutdown the reactor.
RP V water level or RCS      subsequent manu al          Op. Modes: Power heat remova l.                actions taken at the        Operation Op. Modes: Power              reactor control consoles Operation                    are not successful in shutting down the reactor.
Op. Modes: Power Operation 88
 
GENERAL                  SITE AREA                      ALERT EME RGENCY                EMERGENCY                                        UNUSUAL EVENT SU6    Loss of all onsite or offsite communications capabi lities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS    Loss of all AC    SSS    Loss of all vital and vita l DC power      DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9    Hazardous event affecting a SAFETY SYSTEM needed fo r the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 89
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite A    power to esseRtiak mcrgcric) buses.
Operating Mode Applicabi lity: Power Operation, Startup, Hot Standby, Hot Shutdown Eme rgency Actio n Levels:
Note: The emergency director wi ll declare the General Emergency promptly upon determining that 4 hours has been exceeded, or wil l likely be exceeded.
(I)    a.      Loss of ALL offsite and ALL onsite AC power to 4160 YAC EsseRtialf: mcrgency Buses l/2E, l/2F, AND l/2G.
AND
: b.      EITHER of the following:
* Restoration of at least one AC esseRtial emergency bus in less than 4 hours is not like ly.
* Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPY Water Level.
Basis:
This IC addresses a prolonged loss of all power sources to AC esseRtialcrncrgency 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 wi ll 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG I. This will allow add itional time for implementation of offsite protective actions.
Escalat ion of the emerge ncy classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC esseRtialemcrgency bus by the end of the analyzed station blackout copin g 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 esseRtiak mergcncy 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 max imi ze the time avai lable to prepare for, and implement, protective act ions for the public.
The EAL will also require a General Emergency declaration ifthe loss of AC power results in parameters that indi cate an inability to adequately remove decay heat from the core.
90
 
SGS ECL: General Emerge ncy Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptl y upon determining that 15 minutes has bee n exceeded, or will likely be exceeded.
( I)    a.      Loss of ALL offsite and ALL onsite AC power to 4160 V AC EsseAtial Emt!rg.:nc~  Buses l/2E, l/2F, AND l/2G for 15 minutes or longer.
AND
: b.      lndicated vo ltage is less than I 05/210 VDC on ALL 125/250 VDC Bus 1/2 R22-S016 AND l /2R22-S0 17 for 15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. A loss of all AC power compromises the perform ance o f all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, contai nment heat removal/press ure 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 AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
91
 
551 ECL: Site Area Emerge ncy Initiating Condition: Loss of all offsite and all onsite AC power to esseflfiakmcrgcncy buses fo r 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels :
Note: The emergency director will declare the Site Area Emergency promptl y upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( I)    Loss of ALL offs ite and ALL onsite AC power to 4 160 YAC EssentialEmergenc: Buses l/2E, l/2F, AND 1/20 for 15 minutes or longer.
Basis:
This IC addresses a total loss of AC power that compromi ses the pe rfo rmance of all SA FETY SYSTEMS requiring e lectric power including those necessary for emergency core cooling, containment heat removal/press ure contro l, spent fuel heat removal and the ultimate heat sink.
In addition, fi ss ion product barrier monitoring capabilities may be degraded under these conditions. This IC represe nts a condition that invo lves actua l or like ly major failures of plant fu nctions needed for the protection of the public.
Fifteen minutes is a threshold to exclude transient or momentary power losses.
Escalation of the emergency class ification level uses !Cs RGI , FG I or SG I.
92
 
555 EC L : Site A rea Emergency I nitiating C ondition : Inability to shutdown the reactor cau ing a challenge to RPV water level or RCS heat removal.
Operati ng Mode App licability: Power Operation Em erge ncy Action Levels:
( I)    a.      A n auto mati c or manual scram did not shutdown the reactor.
AND
: b.      A ll manual actions to shutdown the reactor have been unsuccessful.
AND
: c.      E ITHE R of the fo ll owing conditions ex ist:
* Reactor vesse l water level cannot be restored and maintained above Minimum Steam Cooling RPY Water Level
* Exceeding the Heat Capac ity Temperature Limit (HCTL) Curve (EO P Graph 2)
Basis :
Thi s IC addresses a fa ilure of the RP to initiate or co mplete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator acti ons to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capabili ty to adequate ly remove heat fro m the core and/or the RCS. This condition will lead to fuel damage if additi ona l miti gation actio ns are unsuccessful and warrants the declaration of a S ite A rea Emerge ncy.
In some instances, the emergency class ification res ulting from this IC/ EAL may be higher than that res ulting from an assess ment of the plant responses and symptoms against the Recognition Catego ry F ICs/ EA Ls. This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a fa ilure to shutdown the reactor. The inclusion of thi s IC and EAL ensures the time ly declaration of a Site Area Emergency in response to prolonged fa iIure to shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalat ion of the emergency classification level uses IC RO I or FG I.
93
 
558 ECL: Site Area Emerge ncy Initiating Condition: Loss of all vital DC power fo r 15 minutes or longe r.
Operating Mode Applicability: Powe r Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: T he emergency director w ill declare the S ite Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( I)      Indicated vo ltage is less than 105/2 10 VDC on ALL 125/250 VDC Bus l/2R22-S0 16 AND 1/2R22-SO17 fo r 15 minutes or longer.
Basis:
T hi s IC addresses a loss of vital DC power which compromises the abili ty to monitor and control SAFETY SYST EMS. In modes above Cold Shutdown, thi s condition invo lves a maj or fa ilure of plant functions needed fo r the protection of the public.
Fi ftee n minutes is a threshold to exclude transient or momentary power losses.
Escalation of the emergency class ification level uses !Cs RO I, FG I or SGS .
94
 
SA1 E C L: A lert Initiati ng C ond ition: Loss of a ll but one AC powe r source to esseRtiak: mcrgc1ic} buses fo r 15 m inutes or longer.
Operating Mode Appl icability: Power Operation, Startup, Hot Standby, Hot Shutdown E me rge ncy Action Levels:
Note: The e merge ncy director w ill declare the Alert promptly upon determining that 15 minutes has bee n exceeded, or w ill li ke ly be exceeded.
( I)    a.      AC power capability to 41 60 YAC faseRtial bncrgenc)' Buses 1/2E, l/2F, AND l/2G is reduced to a sing le power source ( l abk ')I) fo r 15 minutes or longer.        ( Commented [JRB17]: RAJ 13 b rev1s1on AND
: b.      Any addi tional si ng le power source fa ilure wi ll result in a loss of all AC powe r to SAFETY SYSTEMS.
Table SI Unit I                                Uni t 2 Start-up Aux XFMR IC                    Start-up Aux XFMR 2C Start-up Aux XFMR ID                    Start-up Aux XFMR 20 Diesel Generator I A                  Diesel Generator 2A Diesel Generator I 8                  Diesel Generator I 8 Diesel Ge nerator IC                  Diesel Generator 2C Basis:
SAFETY SYSTEM : A system required fo r safe plant operation, coo ling down the plant and/or placing it in the cold shutdown conditi on, including the ECCS . These a re typ ically syste ms class ified as safety-related.
T hi s IC describes a significant degradation o f offsite and onsite AC powe r sources (see Table S 1) such that any additio nal sing le fa ilure wo uld res ult in a loss of all AC power to SAFETY SYSTEMS. ln thi s conditio n, the sole AC power source may be powering one, or more than one, trai n of safety-related equipme nt. This IC provides an escalation path fro m IC SU l .
An " AC power source" is a source recognized in AOPs and EO Ps, and capable of supplying required power to an esseAtialcmergcnc)' bus. So me examples of thi s cond ition are presented below.
* A loss of all offs ite powe r with a concurrent fa ilure of all but one essentialcmergenc)' power source (e.g., an onsite diese l ge ne rator).
* A loss of all offs ite power a nd loss of all esseRtialemergcnc)' powe r sources (e.g., onsite diesel generators) with a single train of esseRtialc mergcncy buses be ing back- fe d from the unit main ge nerator.
95
* A loss of essefltiftlemergcncy power so urces (e.g., onsite diese l ge nerators) with a single train of esseAtialemergenc) buses being back-fed from an offs ite power source.
Fifteen minutes is a threshold to exclude transient or momentary losses of power.
Escalation of the emergency class ification level uses IC SS I.
96
 
SA2 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(!)    a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RPV Water Level RPV Pressure AND
: b.      ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor scram
* ECCS actuation
* Thermal power oscillations greater than 10%
Basis:
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.
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 product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of plant safety.
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.
97
 
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, ifthe 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 is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses !Cs FS I or IC RS I.
98
 
SAS ECL: Alert Initiating Condition: Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels:
Note: A manual 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.
(1)      a. An automatic or manual scram did not shutdown the reactor.
AND
: b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
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 plant safety. An emergency declaration is required even ifthe 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 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. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). 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 the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS 1, an Alert declaration is appropriate for this event.
99
 
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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
100
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(I)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds ( > 35 mph sustained) or tornado strike                          [ Commented [JRB18]: RAJ 13 a reviSJon
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following :
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 IOI
 
condition significantly reduces the margi n to a loss or potential loss ofa fission product barrier, and therefore represents an actual or potential substantial degradation of the level of plant safety.
EAL I. a identifies hazardous eve nts that could result in damage to plant syste ms. A seism ic event is indicated by entry into IC HU2. Flooding is indicated by a signifi cant increase in water levels (exte rnal or internal). High winds are indicated by sustained winds at the site meteorological towe r exceeding 35 mph.
The first threshold for EAL I .b addresses damage to a SAFETY SYSTEM train that is in service/operation since indicat ions 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 threshold for EAL I .b addresses damage to a SAFETY SYSTEM component that is not in service/operation or readil y apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators wi ll make thi s determination based on the totality of avai lable event and dam age report information. This is intended to be a brief assessme nt not requiring lengthy analysis or quantification of the damage.
Escalat ion of the emergency class ification level uses IC FS 1 or RS 1.
102
 
SU1 ECL: Notification of Unusual Event Initiatin g Conditi on: Loss of all offsite AC power capability to esseAtialcmergcm:y buses fo r 15 minutes or longer.
Operating Mode Applicability : Power Operation, Startup, Hot Standby, Hot Shutdo wn Emerge ncy Action Levels:
Note: The emerge ncy director will declare the Unusual Event promptl y upon determining that 15 minutes has been exceeded, or will li kely be exceeded.
( I)    Loss of ALL offs ite AC power capabili ty I able '>2) to 4 160 VAC EsseAtialf:mergcnc:  [ Commented [JRB19] : RAJ 13 b rev ision Buses l/2E, l/2 F, AN D l/2G for 15 minutes or longer.
Table S2 Unit I              I                Unit 2 Start-up Aux XFMR IC        I      Start-up Aux XFMR 2C Start-up Aux XFMR ID        I      Start-up Aux XFMR 20 Basis:
This IC addresses a prolonged loss of offsite power (see Table S2 above). The loss of offsite power sources renders the plant more vulnerable to a com plete loss of power to AC esseAtiale mcrgcnc: buses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offsi te AC powe r source(s) is avail able to the esseAtialemcrgency buses, whether or not the buses are powe red from it.
Fifteen minutes is the thres hold to exclude transient or momentary losses of offsite power.
Escalatio n of the emergency classification leve l uses IC SA I.
103
 
SU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.
Operating Mode Applicability:          Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)      a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minu!es or longer.
Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Basis:
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.
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 plant safety.
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 is reported if it significantly impairs the capability to perform emergency assessments, particularly those 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 104
 
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, ifthe value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS, or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
105
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2)
Note: Use the Unit I or Unit 2 Pretreatment (Flow vs. mR/hr) Graphs to determine ifthe Pretreatment Radiation Monitor exceeds the TV of240,000 &#xb5;Ci/sec.
(!)    Pretreatment Radiation Monitor 1(2)Dl IK601 1(2)Dl IK602 reading greater than 240,000 &#xb5;Ci/sec for greater than 60 minutes.
(2)    Sample analysis indicates that the reactor coolant specific activity is EITHER:
* Greater than 0.2 &#xb5;Ci/gm and less than or equal to 2.0 &#xb5;Ci/gm dose equivalent Irn for greater than 48 hours
* Greater than 2.0 &#xb5;Ci/gm dose equivalent I lJ 1.
Basis:
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 plant safety.
Escalation of the emergency classification level uses !Cs FA I or the Recognition Category R
!Cs.
106
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)    RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)    Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
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 plant safety.
EAL# 1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL# 1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. 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 uses ICs of Recognition Category R or F.
107
 
SU5 ECL: Notification of Unusual Event Ini~iating  Condition: Automatic or manual scram fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (I or 2)
Note: A manual 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.
(I)      a. An automatic scram did not shutdown the reactor.
AND
: b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a. A manual scram did not shutdown the reactor.
AND
: b. EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic scram is successful in shutting down the reactor.
Basis:
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 plant safety.
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 plant's 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 heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
108
 
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, and other concurrent plant conditions. 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 SAS. Depending upon the plant response, escalation is also possible via IC FA I. Absent the plant conditions needed to meet either IC SA5 or FA!, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance will 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 will be evaluated.
* If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.
109
 
SUS ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
(I)    Loss of ALL of the following onsite communication methods:
In plant telephones (includes hardwired and wireless)
Plant Page Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telephone System (FIS) Lines Commercial phones Basis:
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 theNRC.
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).
EAL #I addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 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 of Georgia, Appling County, Jeff Davis County, Tattnall County and Toombs County.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
110
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure ATWS ................................................................................... Anticipated Transient Without Scram BLDG ................................................................................................................................... Building BWR ............................................................................................................. Boiling Water Reactor CB .......................................................................................................................... Control Building CC ......................................................................................................................... Cubic Centimeter CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CPM .................................................................................................................... Counts Per Minute CPS ..................................................................................................................... Counts Per Second DC .............................................................................................................................. Direct Current DEI ............................................................................................................... Dose Equivalent Iodine DW ....................................................................................................................................... Drywell DWRRM .................................................................................... Drywell Wide Range Rad Monitor EAL ........................................................................................................... Emergency Action Level ECCS ............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level ENN ............................................................................................. Emergency Notification Network ENS ................................................................................................ Emergency Notification System EPA ............................................................................................. Environmental Protection Agency EPG ............................................................................................... Emergency Procedure Guideline FAA. .............................................................................................. Federal Aviation Administration FBI .................................................................................................. Federal Bureau of Investigation FEMA ............................................................................. Federal Emergency Management Agency FSAR................................................................................................... Final Safety Analysis Report FTS ......................................................................................... Federal Telecommunications System GE ...................................................................................................................... General Emergency GM ........................................................................................................................................... Gram HCTL .......................................................................................... Heat Capacity Temperature Limit HNP ................................................................................................................... Hatch Nuclear Plant HOO .................................................................................. Headquarters Operations Officer (NRC)
HPCI .............................................................................................. High Pressure Coolant Injection IC ........................................................................................................................ Initiating Condition ISFSI ........................................................................... Independent Spent Fuel Storage Installation LOCA ........................................................................................................ Loss of Coolant Accident MSL ....................................................................................................................... Main Steam Line
&#xb5;Ci ................................................................................................................................. micro-Curie mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man NE ...................................................................................................................................... Northeast NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. North American Aerospace Defense Command (NO)UE .......................................................................................... (Notification Of) Unusual Event OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM ........................................................................................... Offsite Dose Calculation Manual A-I
 
ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG ....................................................................................................... Protective Action Guideline PBX ........................................................................................................... Private Branch Exchange PCIS .................................................................................... Primary Containment Isolation System PRAIPSA .................................... Probabilistic Risk Assessment I Probabilistic Safety Assessment PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCIC ............................................................................................... Reactor Core Isolation Cooling RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RHR ............................................................................................................. Residual Heat Removal RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RWCU .......................................................................................................... Reactor Water Cleanup Rx .......................................................................................................................................... Reactor SAG........................................................................................................ Severe Accident Guideline SAR .............................................................................................................. Safety Analysis Report SC ................................................................................................................ Secondary Containment SCBA ..................................................................................... Self-Contained Breathing Apparatus SE ....................................................................................................................................... Southeast SEP ..................................................................................................................................... Separator SFP ........................................................................................................................... Spent Fuel Pool SNC ....................................................................................................... Southern Nuclear Company SPDS ............................................................................................ Safety Parameter Display System SW ..................................................................................................................................... Southwest TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel TV ........................................................................................................................... Threshold Value VAC ......................................................................................................... Volts Alternating Current VDC .................................................................................................................. Volts Dire ct Current VOIP ................................................................................................... Voice Over Internet Protocol A-2
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA P AG exposure levels offsite for more than the immediate site area.
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.
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 P AG exposure levels.
Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The following are key terms necessary for overall understanding the NEI 99-0 I emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL): 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 effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NOUE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1. Secondary Containment OPERABLE per Technical Specification 3.6.4.1 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 may 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.
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control of HNP Security.
PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
B-2
 
REFUELING PATHWAY: This includes the reactor cavity, the transfer canal, and the spent fuel pool.
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.
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.
UNISOLABLE: 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.
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.
8-3
 
Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 Responses to Requests for Additional Information VOGTLE ELECTRIC GENERATING PLANT UNITS 1AND2 EAL SCHEME MARKED-UP PAGES
 
VOGTLE ELECTRIC GENERATING PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND .................................................................................. 1 1.1 0PERATINGREACTORS .................................................................................................. l 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ************************************* l 1.3 NRC ORDEREA-12-051 ................................................................................................2 1.4 ORGANIZATION AND PRESENTATION OF INFORMATION *********************************************** 3 1.5 IC AND EAL MODE APPLICABILITY .............................................................................. 3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ....................................... 5 2.1 GENERAL CONSIDERATIONS .......................................................................................... 5 2.2 CLASSIFICATION METHODOLOGY ********************************************************************************* 6 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS ****************************************** 6 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ******************************** 6 2.5 CLASSIFICATION OF IMMINENT CONDITIONS *******************.****************************.************** 7 2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ******************* 7 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS ................................................................. 7 2.8 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................................. 7 2.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION **************** 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ......................... 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .............*..... 26 5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ..... ;....... 46 6  FISSION PRODUCT BARRIER ICS/EALS ............................................................... 49 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ........ 65 8  SYSTEM MALFUNCTION ICS/EALS ........................................................................ 88 APPENDIX A - ACRONYMS AND ABBREVIATIONS...................................................... A*1 APPENDIX B - DEFINITIONS ........................................................................................ B*1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections for this document are:
* 10 CFR &sect; 50.47(a)(l)(i)
* 10 CFR &sect; 50.47(b)(4)
* 10 CFR &sect; 50.54(q)
* 10 CFR &sect; 50.72(a)
* 10 CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* 10 CFR &sect; 50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by regulatory guidance documents. Documents of particular relevance to NEI 99-01 include:
NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines 10 CFR &sect; 50. 72 and&sect; 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions also may be directed to the NEI Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NEI 99-0 I is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR &sect; 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR &sect; 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a 10 CFR &sect; 50.47 emergency plan.
The generic !Cs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HU I covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. Additionally, appropriate aspects ofIC HUI and IC HAI will also be included to address a HOSTILE ACTION directed against an ISFSI.
 
The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result ofa tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision-makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel:(!) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool.
* A display in an area accessible following a severe event.
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These 2
 
EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the plan's effectiveness. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.
1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode                R        c      E        F        H        s Power Operations          x                x        x        x        x Startup              x                x        x        x        x Hot Standby            x                x        x        x        x Hot Shutdown            x                x        x        x        x Cold Shutdown            x        x      x                  x Refueling            x        x      x                  x Defueled              x        x      x                  x 3
 
Vogtle Units 1 and 2 Technical Specifications Table 1.1-1 provides the following operating mode definitions:
Reactivity    % Rated Condition    Thermal        Average RCS Mode              Title Temperature (&deg;F)
(Kerr)      PowerC*l 1    Power Operation        2: 0.99          >5                NA 2    Startup                2: 0.99          :::: 5            NA 3    Hot Standby            < 0.99          NA              2: 350 4    Hot ShutdownCb)        < 0.99          NA        350 > Tavg > 200 5    Cold Shutdown<bl      < 0.99          NA              ::::200 6    Refueling< cl            NA            NA                NA (a)    Excluding decay heat.
(b)    All reactor vessel head closure bolts fully tensioned.
(c)    One or more reactor vessel head closure bolts less than fully tensioned.
In addition to these identified modes, "Defueled" is also applicable to the Vogtle EAL scheme, consistent with NEI 99-01 guidance. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Vogtle EALs with no modifications from NEI 99-0 I.
When a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
4
 
2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes and the informing basis information. In the recognition category F matrices, EALs are referred to as fission product barrier thresholds; the thresholds serve the same function as an EAL.
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.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
Indications will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 will be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting in an expected event or condition that 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with the operating license is maintained, provided the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 &sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
5
 
Although the EALs have been developed to address a full spectrum of possible events and conditions that may warrant emergency classification, a provision for classification based on operator/ management experience and judgment is sti ll necessary. The NEI 99-0 I sche me provides the emergency director with the ability to classify events and conditions based on judgment using EA Ls consistent with the emergency classification level (ECL) definitions (refer to Category H). The emergency director will need to determine ifthe effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A sim ilar provision is incorporated into the fission product barrier tables; judgment may be used to determine the status of a fission product barrier.
2.2 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 . An EAL(s) evaluation must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is declared in accordance with plant procedures .
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 NS IR/DPR-ISG-01 .
2.3 CLASSIFICATION OF MULTIPLE EVENTS A '0 CONDITIONS In the event of multiple emergencies or conditions, the user will identify all EA Ls met or exceeded. The highest app licable ECL identified during this review is declared. For example:
If an Alert EAL and a S ite Area Emergency EAL are met, a Site Area Emergency should be declared.
There is no " additive" effect from multiple EALs meeting the same ECL. For example:
If two Alert EALs are met, an Alert will be declared.
Related guidance for classification of rapidly escalating eve nts or conditions is provided in Regulatory Issue Summary (RlS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events .
2.4 CONSIDERATION OF MODE CHANGES D URING CLASSIFICATION The mode in effect at the time an event or condition occurred, and prior to any plant or operator response, determines whether an IC is applicab le. If an event or condition occurs, and resu lts in a mode change before the e mergency is declared, the eme rge ncy classification leve l 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 dil'forent mode i, reached, an: n<.'\\ event or condition. not related to the original event or condition. rt'quiring emergl'llC) classilication should be evaluated against the !Cs and l ~ ALs applicable to the operating mode at thc time of th<' nc\1 event or condition.                                  ( Commented [JRBl] : RAJ I revJSion 6
 
2.5 C LASSIFICATION O F I MM INENT C ONDITIO S A lthough EALs provide specific thres holds, the emerge ncy director must remain alert to events or conditions that co ul d lead to meeting or exceeding an EAL within a relative ly short period of time (i.e., a change in the ECL is IMMlNENT). If, in the j udgment of the emerge ncy director, meeting an EAL is IMMlNENT, the emergency classificati on will be made as though the EAL has bee n met. Wh ile applicable to all emergency class ification leve ls, thi s approach is part icularly important at the higher emergency class ification levels since it prov ides additional time for imple mentation of protective meas ures.
2.6 E M ERGENCY CLASS IFICATION L EVE L UPGRADI NG AND D OWNG RADI NG SNC policy is that once an emergency class ification is made, it cannot be downgraded to a lower class ificat ion. Te rm ination criteri a conta ined in procedure N MP-EP-11 0, Emergency Class ification and Initial Act ions shall be compl eted for an event to be terminated. At terminati o n, on an event specific bas is, the site enter either norm al operating co nditions or a recovery condi tion wi th a recovery organ ization es tab lished for turnover from the ERO. G uidance concerning class ification of rapid ly escalating events or conditions is prov ided in RI S 2007-02 .
2.7 C LASSIFI CATION OF SHORT-LIVE D E VE 'TS Event-based ICs and EA Ls de fine a variety of specific occurrences that have pote ntial or actual safety signifi cance. By the ir nature, some of these events may be short-li ved and end before the emerge ncy class ification assessment can be completed:-. Ifan c1cnt occurs that meets or cxcc<"ds an l~A I , the associated L'CL must be dcclareJ rcuardlcss of its continued prc,;cnce at !he time of declaration. Por example an earthquake, ~r fa ilure      [ Commented [JRB2]: RAJ 2 Revision of the reactor protection system to automat icall y scram/trip the reactor fo ll owed by a successful manual scram/tri p.
2.8 C LASSIFICATION O F TRANS IENT CONDITIONS Many of the ICs and EALs in thi s document empl oy time-based crite ria that require the IC/EAL conditions be present fo r a defi ned peri od of time before an emerge ncy declaration is warranted. In cases where no time-based criterion is specified, so me trans ient conditions may cause an EAL to be met fo r a brief period of time. T he fo llowing guidance sho uld be applied to the class ification of these conditions.
EA L mome ntarily met during expected plant response - When an EAL is briefly met during an expected (norma l) plant respo nse, an emerge ncy declaration is not warranted provided that assoc iated systems and components are operating as expected, and operator actions are perfo rm ed in accordance with procedures.
EA L momentaril v met but the condition is corrected prior to an emergencv declaration -
If an operator takes prompt manual acti on to address a condition, and the action is successful in correcting the condi tion prior to the emergency declaration, then the applicable EAL is not cons idered met and the assoc iated emerge ncy declaration is not required. This exam ple prese nts an ill ustration:
7
 
An ATWS occurs and the auxiliary fee dwate r system fa ils to automat ically start.
Steam generator leve ls rapidl y decrease and the plant enters an inadequate RCS heat removal condition (a potenti al loss o f both the fuel clad and RCS barriers). If an operator manually starts the auxiliary fee dwater system in accordance with an EOP step and clears the inadequate RCS heat removal conditi on prior to an emergency declaration, then the class ification will be based on the ATWS onl y.
It is important to stress that the 15-minute emergency c lass ification assessment period is not a "grace period" during whi ch a class ification may be de layed to allow the performance of a corrective action that would obviate the need to class ify the event; emergency c lass ification assessments must be deliberate and timely, with no undue delays. The prov ision di scussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the emerge ncy di rector completing the review and steps necessary to make the emergency declaration.
This prov ision is included to ensure that any publi c protecti ve actions resulting from the emerge ncy class ification are truly warranted by the plant conditions.
2.9 AFTER-TH E-FACT DISCOVE RY OF AN EMERGENCY EVENT OR CONDITIO N In some cases, an EAL may be met but the emerge ncy class ification was not made at the time of the eve nt or condition. Personnel could di scover an event or condi tion existed that met an EAL, but no emergency was declared, and the event or conditi on no longer exists at the tim e of di scove ry. It may be the eve nt or conditi on was not recognized at the time, or there was an error in the emergency class ification process.
In these cases, no emergency declaration is warranted, but the guidance in NU REG-1022 is applicable. Specifically, the eve nt wi ll be reported to the NRC in accordance with I 0 CFR &sect; 50.72 within one hour of the discove ry of the undeclared event or condition. The licensee will also noti fy appropriate state and local age ncies in accordance with the agreed upon arrangements.
8
 
i 3    ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL                  SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY RGI Release of            RSI Release of            RAI Re lease of          RUI Release of gaseous radioactivity      gaseous radioactivity      gaseous or liquid        gaseous or liquid resulting in offsite      resulting in offsite      radioactivity resulting  radioactivity greater dose greater than 1,000    dose greater than I 00    in offs ite dose greater than 2 times the mrem TEDE or 5,000        mrem TEDE or 500          than IO mrem TEDE        ODCM limits for 60 mrem thyroid CDE.          mrem thyroid CDE .        or 50 mrem thyroid      minutes or longer.
Op. Modes: All            Op. Modes: All            CDE.                    Op. Modes: All Op. Modes: All RG2 Spent fuel pool        RS2      Spent fuel pool  RA2 Significant          RU2    UNPLANNED leve l cannot be          level at -194- 195 foot    lowering of water level  loss of water leve l restored to at least .J-94 level (Level 3).          above, or damage to,    above irradiated fuel. ( Commented [JRB4]: RAJ J.d revision 195 foot l evel ~        Op. Modes: All            irradiated fuel.        Op. Modes: All
~ for 60 minutes or                                  Op. Modes: All                                  ( Commented [JRBl]: RAJ J .d revision longer.
Op. Modes: All RAJ Rad iation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CO E.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* 1fan ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is knowTI to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.
* The pre-calculated effluent monitor values presented in EAL# 1 will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Plant Vent RE-12444E                          50 &#xb5;Ci/cc Turbine Building Vent (SJAE) RE -12839E        2.1 x 103 &#xb5;Ci /cc (2)    Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid COE greater than 5,000 mrem for one hour of inhalation.
Basis:
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 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 monitor reading threshold values are 10
 
determined usi ng a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used to determine the monitor reading thres hold values in !Cs RS I and RA 1. This protocol will maintain intervals between the thres hold values fo r the three class ifications. Since dose are generally not monitored in real-time, a release duration of one hou r is ass umed, and the thres hold values are based on a site boundary (o r beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting.
The TE DE dose is set at the EPA PAO of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 rati o of the EPA PAO fo r TE DE and thyroid CDE.
Class ification based on effluent monitor readings assumes that a release path to the environment is established. lfthe effluent fl ow pa t an effluent monito r is known to have stopped due to actions to isolate the release path, then the e fflu ent monitor reading is no longe r valid fo r class ification purposes.
11
 
RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 4 l 9'i foot leve l (LeYel 3) for 60 minutes or longer.                                                                ( Commented [JRBS] : RAJ J d rev1S1on Operating Mode Applicability: All Emergency Action Levels :
Note: The emergency director will declare the General Emergency promptl y upon determining that 60 minutes has bee n exceeded, or will like ly be exceeded.
( 1)    Spent fuel pool leve l cannot be restored to at least +94- 195 foot level (LeYel 3) for 60 [ Commented [JRB6]: RAJ J d revision minutes or longe r.
Basis:
Th is IC addresses a significant loss of spent fuel pool inventory control and makeup capabili ty leading to a prolonged uncove ry of spent fue l. The spe nt fuel pool level instrument is located outside the control room but in close proximity. 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 we ll after another General Emergency IC was met; however, it is included to provide classification diversity.
12
 
RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE.
Operating Mode Applicabi lity: All Emergency Action Levels: ( I or 2 or 3)
Notes :
* The emergency director wi ll declare the Site Area Eme rge ncy promptly upon determining that the applicable time has been exceeded, or wi ll likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* If the effluent flow past an effluent monitor is known to have stopped due to act ions to iso late the re lease path, then the effl uent monitor reading is no longe r valid for classification purposes .
* The pre-calcul ated effluent monitor va lues presented in EAL# I will be used for emergency class ification assessments until the results from a dose assessme nt using actual meteorology are avail ab le.
( I)      Reading on ANY of the fo llowing radi ation monitors greater than the reading shown for 15 minutes or longer:
Plant Vent RE-12444E                            5.0 &#xb5; Ci/cc Turbine Building Vent (SJAE) RE-12839E            2.1 x I 0 2 &#xb5;Ci/cc (2)      Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CO E at or beyond the site boundary.
(3)      Field survey results indicate EITHER of the following at or beyond the site boundary:
* C losed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid COE greater than 500 mrem for one ho ur of inhalation.
Basis:
T his IC addresses a release of gaseous radioactivity that resu lts in projected or actual offsite doses greater than or equal to 10 percent 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 effl uent EALs are included to provi de a bas is for classifying eve nts and conditions that cannot be read ily 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 acc ident events and condi tions. The monitor reading threshold values are 13
 
determined using a dose assess ment method that back calculates from the dose values specified in the IC. The meteorology and so urce term (noble gases, particul ates, and haloge ns) used is the same as those used to determine the mon itor reading thres hold values in !Cs RGI and RA I. This protocol maintains interva ls between the thres hold values fo r the three classifications. Since doses are generally not moni tored in real-time, a release durati on of one hour is ass umed, and the thres hold values are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour thyro id, whi chever is more limi ting.
The T EOE dose is set at 10 percent of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the I :5 ratio of the EPA PAG for TEOE and thyroid CO E.
Class ification based on e ffluent monitor readings ass umes that a release path to the environment is established. If the e ffluent fl ow past an effluent monitor is known to have stopped due to actions to isolate the re lease path, then the effluent monitor reading is no longe r valid for class ificat ion purposes.
Escalation of the emerge ncy class ification level uses IC RG I.
14
 
RS2 EC L: Site A rea Emergency Initiati ng Condition: Spent fue l pool level at fl 19'\ foot level (Le\'el 3).                    ( Commented [lRB7] : RAl 3 d revision Operating Mode A pplicabi li ty: A ll Emergency Action Levels:
( I)    Lowering of spent fuel pool level to 94- 195 foot level (bevel 3) ,                        ( Commented [JRBB] : RAl 3,d revision Basis:
T hi s IC addresses a significant loss of spent fuel pool inventory control and make up capabil ity leading to IMMIN ENT fuel dam age, The spent fu el pool level instrument is located outs ide the control roo m but in close proximity. This condi tion ste ms fro m maj or fa ilures of plant fun ctions needed to protect the pub lic that warrant a Site Area Emerge ncy declaration.
It is recognized that this IC would li ke ly not be met unti l we ll after another S ite Area Emergency IC was met; however, it is included to provide class ification diversity.
Escalation of the emergency class ification level uses IC RGI or RG2 15
 
RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resul ting in offsite dose greater than 10 mrem TEDE or 50 mrem thyro id CDE.
Operating Mode Applicability : A ll Emergency Action Levels: ( I or 2 or 3 or 4)
Notes:
* The emergency director wi ll declare the Alert promptly upon determining that the applicable time has been exceeded, or w ill li ke ly be exceeded.
* If an ongoing release is detected and the release start ti me is unknown, ass ume that the release duration has exceeded 15 minutes.
* 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 class ification purposes.
* The pre-calcul ated e ffluent monitor values presented in EAL# 1 will be used for emergency classification assess ments until the res ults from a dose assess ment using actual meteorology are avai lable.
( 1)    Reading on ANY of the fo llowing radiation monitors greater than the reading shown fo r 15 minutes or longer:
Plant Vent RE- 12444E                          0.50 &#xb5;Ci/cc Turbine Building Vent (SJAE) RE-128390          2.1 x 101 &#xb5;Ci/cc (2)      Dose assessment using actual meteoro logy indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary.
(3)      Analys is of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than IO mrem TEDE or SO mrem thyroid CDE at or beyond the site boundary fo r one hour of exposure.
(4)      Field survey resu lts indicate EITHER of the following at or beyond the site boundary:
* C losed window dose rates greater than IO mR/hr expected to continue for 60 minutes or longer.
* Analyses of fie ld survey sampl es indicate thyroid C DE greater than SO mrem for one hour of inhalation.
Basis:
This IC addresses a re lease of gaseous or li quid radioactivity that res ults in projected or actual offsite doses greater than or equal to 1perce nt of the EPA Protective Act ion Guides (PA Gs). It includes both monitored and un-monitored releases. Releases of this magni tude represent an actual or potential substantial degradation of the level of plant safety as indicated by a radio log ical release that significantly exceeds regul atory limits (e.g., a significant uncontrolled release).
16
 
Radiological effluent EALs are included to provide a bas is for classifying events and conditions that cannot be readily or appropriately class ified on the bas is of plant conditions alone. The inclusion of both plant condition and radiological effluent EA Ls more fully addresses the spectrum of poss ible acc ident events and conditions. The monitor readi ng thres hold values are determined using a dose assessment method that back calculates from the dose va lues specified in the IC. The meteorology and source term (noble gases, particul ates, and halogens) used is the same as those used to dete rmine the monitor reading threshold values in ICs RG I and RS I. This protocol mainta ins intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (o r beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whicheve r is more limiting.
The TEOE dose is set at I percent of the EPA PAG of 1,000 mrem whil e the 50 mrem thyroid COE was established in consideration of the I :5 ratio of the EPA PAG for TEOE and thyroid COE.
Class ification based on effl uent monitor readings assumes that a release path to the environment is established. If the effl uent flow past an effl uent monitor is known to have stopped due to actions to isolate the release path, then the effl uent monitor readi ng is no longer valid fo r class ification purposes.
Escalat ion of the emergency classification level uses IC RS I.
17
 
RA2 ECL: Alert Initiating Condition: Significant lowering o f water level above, or damage to, irradi ated fuel.
Operating Mode Applicability: A ll Emergency Action Levels: (I or 2 or 3)
(I)    Uncovery of irradiated fue l in the REFUEUNG PATHWAY .
(2)    Damage to irradiated fuel resu lting in a release of radioactivity from the fuel as indicated by l I II< t1 I *\larm on ANY of the following radiation monitors:                            ( Commented [JRB9]: RAJ S revision Fuel Handling Building RE-008 CNMT BLDG Low Range** RE-002/003                **Mode 6 onl y during fu el movement Fuel Handling BLDG EF FL. ARE-2532 A/B Fuel Handling BLDG EFFL. ARE -2533 A/B (3)    Lowering of spent fuel pool level to 204 feet (Level 2).
Basis:
REFUEUNG PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through whi ch irrad iated fuel may be moved, but not including the reactor vessel.
This IC addresses eve nts that have caused IMMINENT or actual damage to an irradi ated fuel asse mbly, 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 rel ease of radioactivity to the environment. As such, they represent an actual or potenti al substantial degradation of the level of plant safety. The spent fuel pool leve l instrument is located outside the control room but in close proximity.
This IC applies to irradiated fuel that is lice nsed 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 IC E-HU I.
Escalation of the emergency is based on e ithe r Recognition Category R or C ICs.
This EAL escalates from RU2. 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), sig nificant changes in water and radiation le ve ls, or other plant parameters . Computational aids may also be used (e.g., a boil-off curve).
Classification of an eve nt usi ng this EAL will be based on the totality of available indi cations, reports and observations.
18
 
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 the fue l is actually uncovered. To the degree possible, readings will be considered in combination with othe r available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vesse l may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes .
This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fue l. Damaging events may include the dropping, bumping or binding of an assemb ly, or dropping a heavy load onto an assembly. A ri se in readings on radiation monitors will be considered in conjunction with in-plant reports or observations ofa potential fuel damaging event (e.g., a fuel handling accident).
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 is a precursor to a loss of the ab ili ty to adequately cool the irradiated fuel assembles stored in the pool.
Escalation of the emergency classification level uses ICs RS I or RS2.
19
 
RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, coo ldown or shutdown.
Operating Mode A pplicability: All Emergency Action Levels: ( I or 2)
Note: 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.
(I)    Dose rate greater than 15 mR/hr in ANY of the following areas:
* Control Room (RE-00 I )
* Central Alarm Station (Survey Only)
(2)    An UNPLANNED event results in radiation leve ls that prohibit or impede access to any Table HI plant rooms or areas:
Table Hl Applic able Building                  Room Number Mode ICB-226, ICB-A45, 3
2CB-223, 2CB-A22 ICB-A77, ICB-B6 1, ICB-B76, ICB-B79 3
2CB-A79, 2CB-BO I Control Building    2CB-B04, 2CB-Bl8 I CB-226, ICB-A45 ICB-B84, 2CB-B85                            4 2CB-223, 2CB-A22 ICB-A48, ICB-A50 4
2CB-A 15, 2CB-A 16 AFW Pump Operation and standby AFW Pump House                                                I, 2, 3 Readiness IAB-A28, 2AB-A72 I, 2, 3 A-level demin vessel valve galleries IAB-A24, 2AB-A77                            3 IAB-A08, 2AB-A I0 I                        3 Auxiliary Building  IAB-C85, IAB-C89 4
2AB-C38, 2AB-C44 IAB-B l5 MEZZ IAB-Bl 9 MEZZ 4
2AB-B 11 7 MEZZ 2AB-B 11 9 MEZZ Basis:
UNPLANNED : A parameter change or an event that is not I ) the res ult of an intended evolut ion 20
 
or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
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 plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be app licable.
For EAL #2, an A lert 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 c lassification is not conti ngent upon whether entry is actually necessary at the time of the increased radiation levels. Access wi ll be considered as impeded if extraordinary measures are necessary to faci litate 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.
* The plant is in an operati ng 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 activ ity th at includes compensatory measures whi ch 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.
Escalation of the emergency class ification level uses Recognition Category R, C or F !Cs.
21
 
RU1 ECL : Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Ap plicability: All Emergency Action Levels: ( I or 2 or 3)
Notes:
* The emergency director wi ll declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
( I)      Reading on ANY of the following effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
SG Slowdown Effluent Line (RE-0021)                2 x release permit setpoint Turbi ne Building Drain Effluent Line (RE-0848)    2 x release permit setpoint
            +llFlliAe BuilEliAg Ve111, 8JAe {Roe 128~9)        2 1* Felease JleFAlil seljleiHI Plant Vent (RE-1 2442C)                            2 x release permit setpoint Plant Vent (RE- 12444C)                            2 x release permit setpoint Turbine Building Vent. <,J.'\E (Rl::-12839Cl                                              [ Commented [JRB10] : RAJ revisi on No Confirmed Primary-Secc>ndary Leakage      1.6 x-, 0*3 ~1Ci!cd                  [ Commented [JRB11]: RAJ 6 revi51on Confirmed Primary-Secondary Leakage      2 x !!lease permll sctpoin~            ( Commented [JRB12]: RAJ 6 revision (2)      Reading on ANY of the fo llowing effl uent radiation monitor greate r than 2 times the alarm setpoint establ ished by a current radioactivity discharge permit fo r 60 minutes or longer.
Liquid Radwaste Effluent Li ne (RE-0018)        2 x release permit setpoint Waste Gas Process Effluent Line (ARE-00 14)      2 x release permit setpoint (3)      Sample analysis fo r a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
This IC addresses a potential decrease in the level of plant safety as indicated by a low-level radio logical re lease 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.
22
 
Nuclear power plants incorporate des ign features intended to contro l the release of radioacti ve e fflu ents to the environm ent. Admini strati ve controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontro lled radioacti ve release to the environment indicates degradation in these features and/or controls.
Radi ological effl uent EALs are included to prov ide a bas is fo r class ify ing events and conditions that cannot be readily or appropriately class ified on the bas is of plant conditions alone. The inclusion of both plant condi tion and radiological e ffluent EA Ls more full y addresses the spectrum of possible acc ident events and condi tions.
Class ification based on effluent monitor readings assumes that a release path to the environment is established. lft he effluent fl ow past an effluent monitor is known to have stopped due to actions to isolate the re lease path, then the effluent monitor reading is no longe r valid for class ification purposes.
Releases will not be prorated or averaged. For exampl e, a release exceeding 4 times release limi ts fo r 30 minutes does not meet the EA L.
EA L # 1 - This EA L addresses norm ally occurring continuous radioacti vity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EA L addresses radioacti vity releases that cause effl uent radiation moni to r readings to exceed 2 times the limit established by a radioacti vity di scharge permit. This EAL will typically be assoc iated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EA L #3 - This EA L addresses uncontro lled gaseous or liquid releases that are detected by sample analyses or environmental surveys, parti cul arl y on unm onitored pathways (e.g., spills of radioacti ve liquids into storm drains, heat exchange r leakage in ri ve r water systems, etc.).
Escalation of the emergency class ification level uses 1C RA 1.
23
 
RU2 ECL: Notification of Unusual Eve nt Initiating Condition: UNPLA              ED loss of water level above irradi ated fuel.
Operating Mode App licability: All Emergency Action Levels:
( l)    a.        UNPLANNED water leve l drop in the REFUELING PATHWAY as indicated by ANY of the fo llowing:
Personnel report of low water level LSHL-0625 eff..seale'>l P low k\'d \lam1 (ALB05                                    [ Commented [JRB13 ]: RA! 3 b rev1S1on E02)
AND
: b.        UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
RE-0008 in the spent fuel pool building RE-0002, -0003, -0004 in contai nment
* RE-00 11 at the seal table
* RE-0005, -0006 in containment *
                    *Not applicable in Modes 1-4 Basis:
UNPLA        ED : A parameter change or an event that is not !) the result of an inte nded evoluti on or 2) an expected plant res ponse to a transient. The cause of the parameter change or event may be known or unknown.
REFUELING PATHWAY : This includes the reactor refuel cavity the fuel transfer canal, and the spe nt fu el poo l, canals and pools through which irradiated fuel may be moved, but not inc luding the reactor vesse l.
This IC addresses a decrease in wate r level above irradi ated fuel sufficie nt to cause e levated radiation levels. This condition can be a precursor to a more se rious eve nt and indicates a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of plant safety.
A water level decrease wi ll be primarily determined by indications from available level instrum entation. Othe r sources of level indications include re ports from plant personnel (e.g.,
from a refue ling crew) or video camera obse rvations (if available). A significant drop in the water level may also cause an increase in the radiation leve ls of adjacent areas that can be detected by monitors in those locations.
24
 
The effects of planned evolutions will be considered. For example, a refueling bridge area radi ation monitor reading may increase due to planned evolutions such as lifting of the reactor vesse l head or move ment of a fuel assemb ly. Note that this EA L is applicable onl y in cases where the elevated reading is due to an UNPLANNED loss of water level.
A drop in wate r level above irradiated fue l within the reactor vessel may be class ified in accordance Recognition Category C during the Cold Shutdown and Re fueling modes.
Escalation of the emergency class ification level uses IC RA2.
25
 
4  COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/ EALS GENERAL                SITE AREA ALERT            NUSUAL EVENT EMERGENCY              EMERGENCY CGI Loss of RPV          CSI Loss of RPV        CAI Loss ofRPV        CUI UNPLANNED inventory affecting      inventory affecting    inventory.            loss of RPV inventory fuel clad integrity with core decay heat        Op. Modes: Cold        for 15 minutes or containment              removal capability. Shutdown, Refueling    longer.
challenged.              Op. Modes : Cold                              Op. Modes: Cold Op. Modes : Cold        Shutdown, Refueling                          Shutdown, Refueling Shutdown, Refueling CA2 Loss of all        CU2 Loss of all but offsite and all onsite one AC power source AC power to            to emergency buses for emergency buses for    15 minutes or longer.
15 minutes or longer. Op. Modes: Cold Op. Modes : Cold      Shutdown, Refi1eling, Shutdown, Refu eling,  Defueled Defi1eled CA3 Inability to      CU3 UNPLANNED maintain the plant in  increase in RCS cold shutdown.        temperature .
Op. Modes : Cold      Op. Modes: Cold Shutdown, Refu eling  Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling cus    Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refi1eling, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Cold Shutdown, Refi1eling 26
 
CG1 ECL : General Emergency Initiatin g Condition: Loss of RP V in ventory affecting fue l clad integri ty with containment challenged.
Operatin g Mode Applica bility: Cold Shutdown, Refu eling Emergency Acti on Levels: ( I or 2)
Note: The emergency director will declare the Ge neral Emergency promptly upon determining th at 30 minutes has been exceeded, or will likely be exceeded.
( I)    a.      RPV level less than 181 '-IO" [TOAF] (63%on RVLI S full range) fo r 30 minutes or longer.
AN D
: b.      A 'Y indication fro m th e Conta inment Challenge Table C I.
(2)    a.      RPV level cannot be monitored fo r 30 minutes or longe r.
AND
: b.      Core uncove ry is indicated by ANY of the fo llowing:
RE-005 OR 006                                                  I ::::: 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Dra in Tank (RCDT) or Waste Holdup Tanks (WHT) levels of suffic ient magnitude to ind icate core uncovery A D
: c.      A Y ind ication from the Containment Challenge Table C I.
Conta inm ent Chall enge Table Cl CONTAINM ENT CLOSURE NOT established*
Explosive mixture inside containment - greater than OR equal to 6% H2 L'\PI <\Nl\I D increase m contamm.:nt pressure                                          { Commented [JRB14]: RAJ 11 reVIsion
* lf CONTAJNME T CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.
Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 142 10-1/2, Containment Building Penetrations Verification - Refueling UNPLANNED: A parameter change or an eve nt that is not I) 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 kn own or unkn own.
27
 
This IC addresses the inability to restore and mai ntain reactor vesse l level above the top of acti ve fue l with containment chall enged. This conditi on represents actual or IMMINEN T substanti al core degradati on or melting with potenti al fo r loss of containment integrity. Releases can be reasonabl y expected to exceed EPA PAG exposure levels offs ite fo r more th an the immedi ate site area.
Foll owin g an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a fu rther reduction in reactor vessel leve l. If RCS/reactor vesse l level cannot be restored, fuel damage is probable.
W ith CONTA IN MENT CLOSURE not established, there is a hi gh potential for a direct and unmonitored release of radioactivity to the environment. If CONTArNM ENT CLO SURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emerge ncy is not required.
The site stieeitie tiressHre at 'Nhieh COJ>ITAIJ>IMEJ>IT is eeAsidered elialleAged may ehaRge based eA the e0Adili0A efthe COJ>JTA[J>IMENT. If the UAit is iA the eeld shHtdewfl made aAd tohe CONTAIJ>IMHIT is fully iAtaet tReA tile site speeifie set peiAt is l:he COJ>ITAIJ>IMEJ>IT desigR press!H'e (52 psig). This is eeAsisteAt with ty13ieal eWHer's greH13s emergeAe) RespeAse Preeedt:lres.
'Nith CONTAIJ>IMEJ>IT CLOSURE established iAteetieRally by the plaAt staffiA preparatiens fer iAsfleetieA, A1aiAteAE!flee, er refueling the set peiAt is based en the 13eAetratieA seals desigA ef 13 psig.1 ( Commented [JRB15]: RAJ 11 revision The ex istence of an explos ive mi xture means, at a minimum, that the containment atm ospheri c hydrogen concentration is sufficient to support a hydrogen bum (i.e., at the lower de tl agrati on limit).
A hydrogen bum will raise contai nment 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 earl y stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explos ive gas mixture in containment. If all installed hydrogen gas monito rs are out-of-servi ce during an event leading to fu el cladding damage, it may not be poss ible to obtain a containment hydrogen gas concentration reading, as ambi ent 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 indi cations to assess wheth er containn1ent is chall enged.
In EAL I .a, R VLI S is used to determine when reactor water level is less than TOA F. RV LI S indi cation is only avai lable during Mode 5 up to the point of reactor head di sassembl y prior to Mode 6 entry. Once RVLIS becomes unava il able classifi cati on of IC CG I is accomplished in accordance with EAL2 .
Jn EAL 2. b, the 30-minute criterion is ti ed to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows suffic ient time to monitor, assess and corre late reactor and pl ant cond itions to determine if core uncovery has actu ally occurred (i.e., to account fo r various accide nt progress ion and instrumentation unce rtainties). It also all ows suffi cient time fo r acti ons to terminate leakage, recover inventory control or makeup equipment, and/or restore level monitoring.
The inabil ity to monitor RPV level may be caused by instrumentation and/or power fa ilures, or water level dropping below the ran ge 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 potenti al sources of wate r fl ow to ensure they indicate leakage from the RPV.
28
 
These EALs address 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.
29
 
CS1 ECL: Site Area Emergency Initiating Condition: Loss ofRPV inventory affecting core decay heat removal capability.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(I)    a.      CONTAINMENT CLOSURE not established.
AND
: b.      RPV level less than 185'-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).
(2)    a.      CONTAINMENT CLOSURE established.
AND
: b.      RPV level less than 181 '-10" [TOAF] (63% on RVLIS full range).
(3)    a.      RPV level cannot be monitored for 30 minutes or longer.
AND
: b.      Core uncovery is indicated by ANY of the following:
RE-005 OR 006                                      I2: 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses a significant and prolonged loss of RPV 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 to protect the public and warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause 30
 
reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage-is probable.
In EALs I .a and 2.a the specified levels represent reactor vessel levels that are lower than the monitoring capability of RCS level instrumentation and therefore must be monitored using RVLIS.
This level will only be observable in Mode 5 with RVLIS in operation. In Mode 6 or when RVLIS is not in operation the IC should be evaluated suing EAL 3.
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 ofEALs l.b and 2.b reflects that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, 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 actions to terminate leakage, recover inventory control or 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 indicate leakage from the RPV.
These EA Ls address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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 uses IC CG 1 or RG 1.
31
 
CA1 ECL: Alert Initiating Condition: Loss ofRPV inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)      Loss ofRPV inventory as indicated by level less than elevation 185'-10" (73% on Full Range RVLIS).
(2)      a.      RPV level cannot be monitored for 15 minutes or longer AND
: b.      UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tank (WHT) levels due to a loss ofRPV inventory.
Basis:
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.
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 EAL #1, a lowering of water level below the bottom ID of the RCS Loop setpoint (187' 6")
indicates that operator actions have not been successful in restoring and maintaining RPV water level.
The 187' 6" level specified in the EAL is the minimum RCS level for RHR operation as outlined in the procedure for mid-loop operations. Below this level, loss of RHR pump net positive suction head (NPSH) may occur resulting in a loss of decay heat removal capability. 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.
Although related, EAL #I 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 Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, 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 indicate leakage from the RPV.
32
 
The 15-minute duration for the loss oflevel indication was chosen because it is halfofthe EAL duration specified in IC CS 1 If the RPV inventory level continues to lower, then escalation to Site Area Emergency uses IC CSL 33
 
CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicabil ity : Cold Shutdown, Refueling, Defueled Emergency Action Levels:
The emergency director wi ll declare the Alert promptly upon determining that 15 minutes has been exceeded, or wi ll likely be exceeded .
(I)      Loss of ALL offsite and ALL onsite AC Power (Table "i I) to BOTH 1(2)AA02 AND                  [ Commented [JRB16]: RAJ 12 revision 1(2)BA03 for 15 minutes or longer.
Table SI Unit I                                      Unit 2 Unit Auxiliary Transformer lNXAA          Unit Auxi liary Transformer 2NXAA Unit Auxiliary Transfo rmer lNXAB          Unit Auxiliary Transformer 2NXAB Reserve Auxiliary Transformer 1NXRA Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer INXRB Reserve Auxiliary Transformer 2NXRB Diesel Generator IA                        Diesel Generator 2A Diesel Generator IB                        Diesel Generator 2B Standby Auxiliary Transformer ANXRA Standby Auxi liary Transformer ANXRA Basis:
This IC addresses a total loss of AC power (see Table S I above) that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary fo r emergency core cooling, containment heat removal/pressure control, spent fue l heat removal and the ultimate heat sink.
When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time availab le to restore an emergency bus to service. Additional time is avai lable due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. When in these modes, this condition represents an actual or potenti al substanti al degradation of the level of plant safety.
Fiftee n minutes is the thres hold to exclude transient or momentary power losses.
Escalation of the emergency class ification level uses IC CS I or RS I.
34
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(1)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F for greater than the duration specified in the following table.
Table C2: RCS Heat-up Duration Thresholds RCS Status              Containment Closure Status            Heat-up Duration Not Intact                    Not Established                  0 minutes (or at reduced inventory)              Established                    20 minutes*
Intact Not applicable                  60 minutes*
(but not at reduced inventory)
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)    UNPLANNED RCS pressure increase greater than 10 psig. (This EAL does not apply during water-solid plant conditions.)
Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling 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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
35
 
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, 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 will allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS I or RS 1.
36
 
CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
(l)      a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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 may 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.
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.
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.
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 37
 
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 plant safety.
The first threshold for EAL l.b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL l.b 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 all 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 uses IC CS 1 or RS I.
38
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss ofRPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( 1)    UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT),
or Waste Holdup Tank (WHT) levels.
Basis:
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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL # 1 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 ifthe 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.
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.
EAL #2 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 39
 
and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they indicate leakage from the RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CAI or CA3.
40
 
CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Co ld Shutdown, Refueling, Defueled Emergency Action Levels:
otc: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or wi ll likely be exceeded.
( I)    a.      AC power capability to BOTH I (2)AA02 AND I (2)BA03 is reduced to a single power source (1 abk 'i I l for 15 minutes or longer.                                          { Commented [JRB17]: RAJ 13 b rev151on AND
: b.      Any additional sing le power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table SI Unit l                                    Unit 2 Unit Auxi liary Transformer INXAA          Unit Auxiliary Transformer 2NXAA Un it Auxiliarv Transformer INXAB          Unit Auxi liary Transformer 2NXAB Reserve Auxiliarv Transformer INXRA        Reserve Auxi liarv Transformer 2NXRA Reserve Aux iliary Transformer INXRB      Reserve Auxiliary Transformer 2NXRB Diesel Generator I A                      Diesel Generator 2A Diesel Generator I B                      Diesel Generator 28 Standby Auxiliarv Transformer ANXRA Standby Auxiliary Transformer ANXRA Basis:
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 .
This IC describes a significant degradation of offsite and onsite AC power sources (see Table S 1 above) where any add itiona l single failure would result in a loss of all AC power to SAFETY SYSTEMS. ln this condition, the so le 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 availab le lo restore another power source to service. Additional time is ava ilab le due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems . When in these modes, this condition is considered to be a potential degradation of the level of plant safety.
An " AC power source" is a source recognized in AOPs and EOPs, and capable of supplyi ng required power to an emergency bus . Examples of this condition include :
41
 
A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g.,
an onsite diesel generator).
A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes is the 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.
42
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F.
(2)    Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
Basis:
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of plant safety. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the emergency director will 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.
EAL# l 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, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators are 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 is the threshold to exclude transient or momentary losses of indication.
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
43
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105 VDC on required 125 VDC buses 1(2)ADI, 1(2)BDI, 1(2)CDI, 1(2)DDI for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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.
This condition is considered to be a potential degradation of the level of plant safety.
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 Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable),
then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
44
 
CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (I or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will be assessed only when extraordinary means are being used 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.).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Georgia and South Carolina; Burke County, Georgia; Aiken County, South Carolina: Barnwell and Allendale, South Carolina; and the Savannah River Site in South Carolina.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
45
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HUI Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 46
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY of the values listed in Table El.
Table El Location of Dose Rate                        Total Dose Rate (Neutron+ Gamma mR/hr)
HI-TRAC 125 Side - Mid-height                              950 Too                                      200 HI-STORM 100 Side - 60 inches below mid-height                        170 Side - Mid-height                                180 Side - 60 inches above mid-height                        110 Center of lid*                                50 Middle of too lid**                              60 Top (outlet) duct                            130 Bottom (inlet) duct                            360
* The center of the too lid represents a 6 in. radius.
          ** The middle of the top lid represents an approximately 4 in. wide cvlindrical "strip" located about mid-distance of the lid.
Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY ofa 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 radiation reading values listed in the table represent 2 times the site-specific cask specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical specification multiple of"2 times" is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of 47
 
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, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under ICs HUI and HAI.
48
 
6 FIS~ION  PRODUCT BARRIER ICS/EALS                      LOSS POTENTIAL LOSS POTENTIAL LOSS FUEL CLAD Re~ognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.                                              - YES      f!ll-1.o~s    of A..'lY T\\o Harri.:l'll Afilll.us~ or Pot.:ntial Lo~s ofThirdllarri.:r FGl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown
                                                                                      ~-------NO SITE AREA EMERGENCY POTENTIAL Loss or Potential Loss of any two barriers. LOSS LOSS LOSS POTENTIAL LOSS I
I                                                      FUEL CLAD                    CONTAINMENT FSl Op. Modes: Power Operation, Hot Standby,
      '      Startup, Hot Shutdown ALERT                                                                          .Efil - Loss or Pot.:nlial Loss of ANY T1111 llarri.::rs I
Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
:FAl                                                        POTENTIAL LOSS                LOSS Op. Modes: Power Operation, Hot Standby,                LOSS FUEL CLAD            RCS Startup, Hot Shutdown
                                                                                                          .EA! -ANY Loss or ANY Poh:ntial Loss offilill.ER Fu.:\ Clad QR RCS 49
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY                            FSl SITE AREA EMERGENCY                              FAl ALERT Loss of any two barriers and Loss or          Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either Potential Loss of the third barrier.                                                      the Fuel Clad or RCS barrier.
Fuel Clad Barrier                                    RCS Barrier                                Containment Barrier LOSS              POTENTIAL LOSS                  LOSS              POTENTIAL LOSS              LOSS              POTENTIAL LOSS I. RCS or SG Tube Leakage                        I. RCS or SG Tube Leakage                      I. RCS or SG Tube Leakage Not Applicable          A. CORE COOLING        A. An automatic or        A. Operation of a    A. A leaking or        Not Applicable CSF-ORANGE              manual ECCS                standby charging    RUPTURED SG is entry conditions        actuation is required      pump is required    FAULTED outside met.                    by EITHER of the          by EITHER of the    of containment.
following:                following:
I. UNTSOLABLE              I. UNISOLABLE RCS leakage                  RCS leakage OR                          OR
: 2. SG tube                2. SG tube RUPTURE.                    leakage.
OR B. RCS INTEGRJTY CSF - RED entry conditions met 50
 
Fuel Clad Barrier                                RCS Barrier                                    Containment Barrier LOSS              POT ENT IAL LOSS              LOSS            POTENTIAL LOSS                LOSS                POTENT IAL LOSS
: 2. Inadequate Heat Remova l                    2. Inadequate Heat Removal                    2. Inadequate Heat Removal A. CORE COOLING A. CORE COOLfNG                Not Applicable        A. HEAT SINK CSF -      Not Appli cabl e          A. CORE COO LING CSF - RE D entry          CSF- ORANGE                                  RED e ntry                                        CSF - RED entry conditions met            entry conditions                              conditions met.                                  conditions met for met                                                                                            15 minutes or longer OR                                      Note* I kat Smk CSI B. HEAT S INK CSF -                            should not he RED entry                                considered RH) if total conditions met                          a\ ailahlc ko.:d\\ atcr tlcm 1s less than 535 Note: I kat Sink CS!*                          gpm due to opcrator should not be                                  action.                                                                    ( Commented [JRB19]: RAI 15 revision considered Rf D 1ftotal aHtilabk ft:cd,\atcr llcl\\ is less than 535 gpm due to operator action.                                                                                                                    { Commented [JRBlB]: RAI 15 revision 51
 
Fuel Clad Barrier                            RCS Barrier                            <'ontainment Barrier LOSS            POTENTIAL LOSS            LOSS            l'OTE"ITIAL LOSS        LOSS              POTE!"\Tl. \L LOS~
: 3. RCS Activity I Containment Radiation  3. RCS Activity I Containment Radiation      3. RCS Activity I Containment Radiation A. Containment            Not Applicable  A. Containment            Not Applicable    Not Applicable          A. Containment radiation monitor                        radiation monitor                                                      radiation monitor RE-005 OR 006                            RE-005 OR 006                                                          RE-005 OR 006
::': 2.6E+5 mR/hr.                      ::': 8.7 E+2 mR/hr.                                                    ::': U E+7 mR/hr.
OR B. Indications that reactor coolant activity is greater than 300 &#xb5; Ci/gm dose equival ent I-131.
52
 
Fuel Clad Barrier                                        RCS Barrier                                        Containment Barrie1*
LOSS              POTENTIAL LOSS                      LOSS              POTENTIAL LOSS                      LOSS              POTENT! \L LOS..,
: 4. Containment Integrity or Bypass                  4. Containment Integ ri ty or Bypass                  4. Containment Integrity or Bypass Not Applicable            Not Appli cable            Not Appli cable            Not Applicable            A Containment iso lation    A CONTAINMENT CSF is required                  RED entry conditions AN D                        met.
EITHER of the                OR fo llowing:              B. CTMT hydrogen I. Containment              concentration greater integrity has been    than 6%
lost based on          OR Emergency Director
: c. I. GGJ>l+.O.ll>I~ 461>1+
GSl<GRAl>IGE>
judgment.
S0R~ t li0R5 OR                        ff!e!{ 'o    !  I,: I ~
: 2. UNISOLABLE                  prt.:>S1      * ,r ti n pathway from the          ~I 'I''                ( Commented [JRB20]: RAI 19 revision contai nment to              AN D the environment
: 2. Less than Four exists.
CTMT fan coolers OR                                and one full train B. Indications of RCS                  ofCTMT Spray leakage outside of                  are operating per containment.                        design for 15 minutes or longer.
: s. Other Indications                                s. Other Indications                                  s. Other Indications Not aoolicable            Not applicable            Not applicable              Not applicabl e          Not aoolicable              Not aoolicable
: 6. Emergency Director Judgment                      6. Emergency Director Judgment                        6. Emergency Director Judgment A. ANY condition in      A . ANY condition in      A. ANY condition in the A. ANY condition in          A. ANY condition in          A. ANY condition in the the opinion of the        the opinion of the        opinion of the              the opinion of the        the opinion of the          opinion of the emergency                emergency director        emergency director          emergency director        emergency director          emergency director director that            that indicates            that indicates loss of      that indicates            that indicates loss of      that indicates indi cates loss of        potenti al loss of the    the RCS Barrier.            potential loss of the    the contai nment            potenti al loss of the the fuel clad            fue l clad barrier.                                    RCS Barrier.              barrier.                    containment barrier.
barrier.
53
 
Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
: 1. RCS or SG Tube Leakage There is no Loss threshold associated with RCS or SG Tube Leakage.
Potential Loss I.A This condition indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
: 2. Inadequate Heat Removal Loss 2.A This condition indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Potential Loss 2.A This condition indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
Potential Loss 2.B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity I Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300&#xb5;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 of2 percent to 5 percent 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.
54
 
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3.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.
Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;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 percent to 5 percent 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.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
55
 
RCS BARRIER THRESHOLDS:
The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I.A will also be met.
Potential Loss I.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment If a leaking steam generator is also FA ULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I.A will also be met.
Potential Loss l.B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock- a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
: 2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss 2.A 56
 
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity I Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
57
 
CONTAINMENT BARRIER THRESHOLDS:
The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss I .A and Loss I .A, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology. This determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, ifthe pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent ofa loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
58
 
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R !Cs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate                      Yes                        No Less than or equal to 25 gpm            No classification          No classification Greater than 25 gpm                Unusual Event per SU4      Unusual Event per SU4 Requires operation of a Site Area Emergency standby charging (makeup)                                            Alert per FA 1 per FSl pump (RCS Barrier Potential Loss)
Requires an automatic or              Site Area Emergency Alert per FA 1 manual ECCS (SI) actuation                  per FSl (RCS Barrier Loss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
: 2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss 2.A This condition represents an IMMINENT core melt sequence that, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing or if reactor vessel level is increasing. Whether the procedure(s) will be effective should be apparent within 15 minutes. The emergency director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it 59
 
is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
: 3. RCS Activity I Containment Radiation There is no Loss threshold associated with RCS Activity I Containment Radiation.
Potential Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent 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 percent 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 containment that would then escalate the emergency classification level to a General Emergency.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. There may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2.
4.A.1 - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the emergency director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Two simplified examples are provided in the middle piping run of Figure 6-F-l. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example is a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FAULTED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of 60
 
containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
4.A.2 - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
See the simplified example in the top piping run ofFigure 6-F-1. The inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
A simplified example is shown in the bottom piping run of Figure 6-F-1. Leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. If there is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. I to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur ifa containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category RI Cs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold I.A.
61
 
Loss 4.B Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment will be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly.
However, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
In the simplified example in the middle piping run of Figure 6-F-1, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. l to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold I .A to be met.
Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. This threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Potential Loss 4.B 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 potential loss of the Containment Barrier.
Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment because containment heat removal/depressurization 62
 
I i I
systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
63
 
Figure 6-F-1: PWR Containment Integrity or Bypass Examples
                                                                                *:
* 4.il."2 *_ Airborne*:*:*:*
                                                                                *
* release from : -: -: * .
* Inside Containment                                                                  . . P.at.h"."~Y.
RCP Seal Cooling 64
 
7    HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                EMERGENCY HGI HOSTILE              HSI HOSTILE              HAI HOSTILE            HUI Confirmed ACTION resulting in      ACTION within the        ACTION within the      SECURITY loss of physical        PROTECTED AREA.          OWNER                  CONDITION or control of the facility. Op. Modes: All            CONTROLLED              threat.
Op. Modes: All                                    AREA or airborne        Op. Modes: All attack threat within 30 minutes.
Op. Modes: All HU2 Seismic event greater than QBE levels.
Op. Modes: All HU3 Hazardous event.
Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HAS Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All HS6 Inability to          HA6 Control Room control a key safety      evacuation resulting in function from outside    transfer of plant the Control Room.        control to alternate Op. Modes: All            locations.
Op. Modes: All HG7 Other                HS7 Other                HA7 Other              HU7 Other conditions exist which  conditions exist which    conditions exist which  conditions exist in the judgment of the  in the judgment of the    in the judgment of the  which in the emergency director      emergency director        emergency director      judgment of the warrant declaration of  warrant declaration of    warrant declaration of  emergency director a General Emergency. a Site Area              an Alert.              warrant declaration of Op. Modes: All          Emergency.                Op. Modes: All          a (NO)UE.
Op. Modes: All                                    Op. Modes: All 65
 
HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by security shift supervision.
AND
: b.      EITHER of the following has occurred:
I.      ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMINENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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 I) 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.
66
 
Timely and accurate communications between security shift supervision 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 NE! 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
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 Security Plan.
67
 
HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a General Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a General Emergency.
68
 
HS1 ECL: Site Area Eme rge ncy Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.
Operating Mode Applicability: All Emergency Action Levels:
( I)      A HOSTIL E ACTION is occurring or has occurred w ithin the PROTECTED AREA as reported by security shift supervision.
Basis:
HOSTILE ACTION : An act toward a nuclear power plant (NPP) or its personnel th at includes the use of violent force to destroy equipme nt, take HOST AGES, and/or intimidate the licensee to achieve an e nd. This includes attack by air, land, or wate r using guns, explos ives, PROJECTILES, ve hicles, or other dev ices used to deli ver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobed ience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such act ivities (i.e., this may include viole nt acts betwee n indi viduals in the owner controlled area (OCA)).
PROTECTED AREA (PA) : The area that encompasses all controlled areas within the security protected area fence .
This IC addresses the occurrence of a HOSTILE ACTION within the PROT ECTED AREA (PA). This event w ill require rapid res ponse and assistance due to the possibili ty for damage to plant equipment.
Timely and accurate communications between security shi ft superv ision and the control room is esse nti al 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 [and Independent Spent Fuel Storage Installation Security Program}.
As time and conditions allow, these events require a he ig htened state of readiness by the plant staff and imple me ntation of onsite protective measures (e.g., evacuation, di spe rsal or shelte ring).
The Site Area Eme rgency declaration wi ll mobilize ORO resources and have them available to develop and implement public protective actions in the unlike ly eve nt that the attack is successful in impairing multiple safety functions .
Thi s IC dees Rat a1313ly te a HOSTILE\ ACTIO!>I diraeted at aR ISi;'Sl PR.OTIICTIID AREA leeated 01:Jtside the 13!a11t PR.OTIICTIID ARIIA (PA); Sl:leh 8:R attaek shel:lld ee assessed H:SiRg IC HA I. It alse does not apply to incidents that are accidental events, acts of civil disobedience, or    ( Commented [JRB21]: RAJ 22 rev151on otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, phys ical di sputes between employees, etc.
Reporting of these types of events is adequate ly addressed by other EA Ls, or the require ments of 10 CF R &sect; 73 .71 or 10 CF R &sect; 50.72.
69
 
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 Security Plan.
Escalation of the emergency classification level uses IC HG I.
70
 
HS6 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    a.      An event has resulted in plant control being transferred from the control room to the remote shutdown panels due to a control room evacuation.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* Core cooling
* RCS heat removal Basis:
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 "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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level uses IC FG I or CG 1.
71
 
HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Site Area Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a Site Area Emergency.
72
 
HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Operating Mode Applicability: All Emergency Action Levels: (I or 2)
(1)    A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by security shift supervision.
(2)      A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control of VEGP security.
This IC addresses the occurrence ofa 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 (PA), or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program}.
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, 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.
73
 
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 C FR &sect; 73.71or10 CFR &sect; 50 .72.
EAL# I is appli cable for any HOSTIL E ACTION occurri ng, or that has occurred, in the OWNER CONTROLL ED AREA (OCA) . Tkis iAel1:1des MJ' aetioA direeted agaiAsl RA ISFSI tilat is loeated 01:1tside tile fllRAt PRon;cnm AREA (PA).                                          [ Commented [JRB22]: RAJ 22 rev1Sion EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of thi s EA L 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 EA L is met when the threat-related information has bee n validated in accordance with station procedures.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe threat invo lves 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 (OCA) was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clari fy thi s point. In thi s case, the appropriate federal agency is intended to be NORAD, FB I, FAA or NRC. The emerge ncy declaration, including one based on other ICs/EALs, shou ld 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 infor mation. This includes info rm ation 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 Security Plan.
Escalation of the emergency class ification leve l uses IC HS I.
74
 
HAS ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(I)    a.      Release of a toxic, corrosive, asphyxiant or flammable gas into any Table HI plant rooms or areas:
AND
: b.      Entry into the room or area is prohibited or impeded.
Table HI Applicable Building                  Room Number Mode lCB-226, 1CB-A45, 3
2CB-223, 2CB-A22 1CB-A77, ICB-B61, 1CB-B76, 1CB-B79 3
2CB-A79, 2CB-BOI Control Building    2CB-B04, 2CB-Bl8 lCB-226, 1CB-A45 I CB-B84, 2CB-B85                          4 2CB-223, 2CB-A22 1CB-A48, 1CB-A50 4
2CB-Al5, 2CB-Al6 AFW Pump Operation and standby AFW Pump House                                                I, 2, 3 Readiness 1AB-A28, 2AB-A72 1, 2, 3 A-level demin vessel valve galleries IAB-A24, 2AB-A77                          3 IAB-A08, 2AB-AIOI                          3 Auxiliary Building    IAB-C85, 1AB-C89 4
2AB-C38, 2AB-C44
                                  !AB-BIS MEZZ IAB-B19 MEZZ 4
2AB-B 117 MEZZ 2AB-B 119 MEZZ Basis:
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 75
 
cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of plant safety.
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 director's 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 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 4.
* The gas release is a planned activity that includes compensatory measures to address the temporary inaccessibility ofa room or area (e.g., fire suppression system testing).
* The action that 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.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19 percent, 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.
Escalation of the emergency classification level uses Recognition Category R, C or F !Cs.
76
 
HAG ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(!)    An event has resulted in plant control being transferred from the control room to the remote shutdown panels due to a control room evacuation.
Basis:
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 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 uses IC HS6.
77
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
78
 
HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3)
(I)      A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by security shift supervision.
(2)      Notification of a credible security threat directed at VEGP.
(3)      A validated notification from the NRC providing information of an aircraft threat.
Basis:
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and represent a potential degradation in the level of plant safety. Security events that do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.7I or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HG!.
Timely and accurate communications between security shift supervision 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 OROs.
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 [and Independent Spent Fuel Storage Installation Security Program}.
EAL #I references security shift supervision 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 &sect; 2.39 information.
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is -
79
 
assessed in accordance with station procedures.
EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HA 1.
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HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels.
Operating Mode Ap plicability: All Emergency Action Levels:
(I)      Seismic event greater than Operating Basis Earthquake (OBE) as indicated by the Seismic Monitoring System                                      1iAElieatiAg gFeateF thaA 0. 12 g aeeeleratian.                                                                                [ Commented [JRB23] : RAl 23 rev ision Basis:
This IC addresses a seismic event that results in accelerations at the plant site greater than those specified fo r an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant 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., perfo rms walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and full y understand any impacts, this event represents a potential degradation of the level of plant safety.
If a seismic event occurs that exceed~ the OHL an audibk alarm \\ill be sounded and the OBL I XCCl'danee Indicator on the Seismic \fonitoring ">yst.:m Panel \\ill change from Clrcen 10 Red        ! Commented [JRB24]: RAl 23  rev 1s1on Event verification wi th external sources should not be necessary during or fo llowing an OBE.
Earthquakes of this magnitude should readily be felt by on-s ite personnel and recognized as a se ismic event (e.g., typical lateral accelerations are in excess of0.08g). The Shift Manager or emergency director may seek external ve rification if deemed appropriate (e.g., a call to the U G or check of internet news sources); 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 uses IC CA6 or SA9.
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HU3 ECL : Notification of Unus ual Event Initiating Condition : Hazardous event.
Operating Mode Applicab ili ty: A ll Emerge ncy Actio n Leve ls: ( I or 2 or 3 or 4 or 5)
N ote: EA L #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
( I)      A tornado strike within the PROTECTED AREA .
(2)      Internal room or area fl ooding of a magnitude sufficient to require manual or automatic e lectrical isolation of a SAFETY SYSTEM component needed fo r the current operating mode.
(3)      Move ment of personnel within the PROTECTE D AREA (PA) is impeded due to an offsite event in volving hazardous materi als (e.g. , an offsite chemi cal spill or toxic gas release).
(4)      A hazardous event that res ults in on- site co ndi tions suffic ient to prohibit the plant staff from access ing the site via personal vehi cles.
(5)      Susta ined hurricane force winds greater than 74 mph fo recast to be at the plant site in the next fo ur hours.
Basis:
PROTECTED AREA (PA) : The area that enco mpasses all controlled areas within the security protected area fence.
SA FETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condi tion, including the ECCS. These are typ ically syste ms class ified as safety-related.
This 1C addresses hazardous events that are considered to represe nt a potential degradation of the level of plant safety.
EA L # 1 addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 addresses fl ooding of a building room or area that res ults in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Class ificati on is also required if the water level or re lated wetting causes an automatic isolation of a SAFETY SYSTEM component from its powe r source (e.g., a breaker or relay trip). To warrant class ification, operability of the affected component must be required by Techn ical Specifications for the current operating mode.
EAL #3 addresses a hazardous materi als event originating at an offs ite location and of suffic ient magnitude to impede the movement of personne l within the PROTECTED AREA (PA).
82
 
EAL #4 addresses a hazardous event that causes an on-site imped iment to ve hi cle move ment 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, or darn fa ilure, or an on-site train derailment blocking the access road.
This EA L is not intended to apply to routine impediments such as fog, snow, ice, or vehi cle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on T urkey Point in 1992, the fl ooding aro und the Cooper Station during the Mid west flood s of 1993, or the fl ooding around Ft. Calhoun Station in 2011 .
EAL #5 addresses the phenomena of the hurri cane based on the severe weather mitigation procedure.
Escalation of the emerge ncy classification leve l is based on ICs in Recognition Categories A, F, Sor C.
83
 
HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.
Operating Mode App licab ili ty: All Emerge ncy Action Levels: ( I or 2 or 3 or 4)
Note: The emergency director will declare the Unusual Eve nt promptly upon determining that the applicable time has been ex ceeded, or will likely be exceeded.                      I (I)    a.      A FIRE is NOT extinguished within 15-minutes of ANY of the fo llowing FIRE detection indications:
* Report from the field (i .e., vis ual observation)
* Rece ipt of mul tiple (more than I) fire alarms or indications
* Field ve ri fication of a single fi re alarm AND
: b.      The FIRE is located within ANY of the Table H2 plant rooms or areas.
(2)    a.      Receipt of a single fire alarm (i.e., no other indications of a FIRE).
AND
: b.      The FIRE is located wi thin ANY of the Table H2 plant rooms or areas.
AND
: c.      The existence of a FIRE is not ve rified within 30-minutes of alarm receipt.
(3)    A FIRE within the plant PROTECTED AREA (PA) er ISFSI PROTECTE:D AREA ,not                  [ Commented [JRB25] : RA! 22 revisi on )
extinguished within 60-minutes of the initial report, alarm or indication.
(4)    A FIRE within the plant PROT ECTED AREA (PA) er JSFSI PROTECTE:D ARE:A that                ( Commented [JRB26] : RAJ 22 revision  )
requires firefighting support by an offs ite fire response agency to extinguish.
Table H2 Containment Build ing NSCW Coolino To wers Diesel Generator Bui lding Auxiliar Buildino Fuel Handlino Bui lding Control Buildin Diesel Fuel Oil Storage Tank Pumphouse Auxiliar Feedwater Pumphouse 84
 
Basis:
FIRE : Combustion character ized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRE . Observation of fl ame is preferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA (PA): The area that encompasses all controlled areas wi thin the security protected area fe nce.
This IC addresses the magnitude and extent of FIRES that may be indicati ve of a potential degradation of the level of plant safety.
The intent of the IS-minute duration is to size the FIRE and to di scriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a Fl RE include a drop in fire mai n pressure, automatic act ivation of a suppress ion system, etc.
Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment, the emerge ncy declaration clock starts at the time that the initial alarm, indi cation, or report was received, and not the time that a subsequent ve rifi cation action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indicatio n or report.
This EAL addresses receipt of a si ng le fire alarm, and the ex istence 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 validi ty of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initi al alarm was rece ived, and not the time that a subseq uent verification action was performed.
A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious act ivation, and not an actual FIRE. For thi s reason, additi onal time is allowed to ve rify the validity of the alarm. The 30-minute period is a reasonable an10unt of time to determine ifan actual FIRE exists; however, after that time, and absent information to the contrary, it is ass umed that an actual FIRE is in progress.
If an actual FIRE is ve rified by a re port from the field, then EAL # l is immediate ly applicable, and the emergency must be declared ifthe FIRE is not exting uished within IS-minutes of the report. If the alarm is ve rified to be due to an equipment fa ilure or a spurious activation, and thi s ve rifi cation occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.
In addition to a FIRE addressed by EA L # l or EAL #2, a FIRE within the plant PROTECTE D AREA (PA) not extinguished within 60-minutes may also potentiall y degrade the level of plant safety. Tkis besis e1cteAes tea FIRE. eee1miAg witkiA tke PROTE.CTED ARE.A efBfl 18f81 leeatee e1:1tsiee the plaAl PROTECTE.D ARE.A (PA).                                                      ( Commented [JRB27] : RAJ 22 rev1Sion 8S
 
lfa FIRE within the plant or 18F81 PROTECTED AREA is of sufficient size to require a                ~mented [JRB28] : RAJ 22 revision response by an offsite firefi ghting agency (e.g., a local town Fi re Department), then the level of plant safety is potentially degraded. The di spatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to active ly support firefighting efforts because the fire is beyond the capability of the Fire Brigade to exti nguish. Declaration is not necessary ifthe agency resources are placed on stand-by, or supporting post-extinguishment recove ry or investigation actions.
Basis- Related Requirements from Appendix R Appendix R to IO CFR 50, states in part:
Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be des igned and located to minimize, consistent with other safety req uirements, the probabi li ty and effect of fires and explosions."
When considering the effects of fire, those syste ms associated with achieving and mainta ining 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 syste ms and because the loss of function of systems used to mitigate the consequences of design bas is acc idents under post-fi re conditions does not per se impact public safety, the need to limit fire damage to syste ms required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the con equences of design basis acc idents.
Append ix R to 10 CFR 50, requires, among other considerations, the use of I-hou r fire barriers fo r the enclos ure of cable and eq uipment and assoc iated non-safety ci rcuits of one redundant train (G .2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case I-hour time period .
Depending upon the plant mode at the time of the event, escalation of the emergency class ification level uses IC CA6 or SA9.
86
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a (NO)UE.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a NOUE.
87
 
8    SYSTEM MALFUNCTION ICS/EALS GENERAL                  SITE AREA ALERT              UNUSUAL EVENT EMERGENCY                  EMERGENCY SGI Prolonged loss of      SSl    Loss of all offsite    SAI Loss of all but one  SUI Loss of all offsite all offsite and all onsite and all onsite AC power        AC power source to        AC power capability to AC power to emergency      to emergency buses for        emergency buses for 15    emergency buses for 15 buses.                    15 minutes or longer.          minutes or longer.        minutes or longer.
Op. Modes: Power          Op. Modes: Power              Op. Modes: Power          Op. Modes: Power Operation, Startup, Hot    Operation, Startup, Hot        Operation, Startup, Hot  Operation, Startup, Hot Standby, Hot Shutdown      Standby, Hot Shutdown          Standby, Hot Shutdown    Standby, Hot Shutdown SA2 UNPLANNED            SU2 UNPLANNED loss of Control Room      loss of Control Room indications for 15        indications for 15 minutes or longer with a  minutes or longer.
significant transient in  Op. Modes: Power progress.                Operation, Startup, Hot Op. Modes: Power          Standby, Hot Shutdown Operation, Startup, Hot Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS    Inability to          SAS Automatic or          SUS Automatic or shutdown the reactor          manual trip fails to      manual trip fails to causing a challenge to        shutdown the reactor, and shutdown the reactor.
core cooling or RCS heat      subsequent manual        Op. Modes: Power removal.                      actions taken at the      Operation Op. Modes: Power              reactor control consoles Operation                      are not successful in shutting down the reactor.
Op. Modes: Power Operation 88
 
GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY SU6 Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Failure to isolate containment or loss of containment pressure control.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS Loss of all AC        SSS Loss of all Vital and Vital DC power        DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 89
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
(I)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND 1(2)BA03.
AND
: b.      EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* CORE COOLING CSF - RED conditions met.
Basis:
This IC addresses a prolonged loss of all power sources to AC emergency 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG I. 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 AC emergency 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 emergency bus will be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success will 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 ifthe loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.
90
 
SGS ECL: General Emergency Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND 1(2)BA03 for 15 minutes or longer.
AND
: b.      Indicated voltage is less than 105 VDC on ALL 125 VDC busses 1(2)AD1, 1(2)801, 1(2)CDI, 1(2)DD1for15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. 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 loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
91
 
551 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND 1(2)BA03 for 15 minutes or longer.
Basis:
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 is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses !Cs RGI, FG I or SG I.
92
 
555 ECL: Site Area Emergency Init iating Co nd it ion: Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.
O peratin g Mode A pplica bili ty: Power Operati on E m er ge ncy Actio n Leve ls:
otc:      I k.11 '>111K ( '>t ~ 10uld 111'1h1,;1:01 J~t"d RI D ti lot.ii, v 11!,1bh.: 1,,1,;thhll~r tlcm JS k:~'
tl*n i '15 gpm du" to np.;r.itor ,11.:lllHl                                                            [ Commented [JRB29] : RAJ 25 rev1S1on
( I)    a.          An automat ic or manual trip did not shutdown the reacto r.
AND
: b.        A ll manual actions to shutdown the reactor have been unsuccessful.
AND
: c.          EIT HE R of the fo ll owing conditi ons ex ist:
* Core Cooling CSF - RED conditions met
* Heat Sink CSF - RED cond itions met Bas is :
T hi s IC addresses a fa ilure of the RP S to initiate or complete an automatic or manual reactor trip that re ults in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are un successful , and continued power generation is challenging the capability to adequately remove heat from the core and/o r the RCS . This condition will lead to fuel damage if additional mitigation actions are unsuccess ful and thus warrants the declaration of a Site Area Emergency.
In some instances, the emergency classificati on resulting from th is IC/EAL may be higher than that resulting from an assess ment of the plant responses and symptoms against the Recognition Category F ICs/ EALs. This is appropriate because the Recognition Category F ICs/ EA Ls do not address the additional threat posed by a fa ilure to shutdown the reactor. The inclusion of thi s IC and EAL ensures the timely declaration of a Site A rea Emerge ncy in re ponse to prolonged fa ilure to shutdown the reactor.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalat ion of the emergency class ification level uses IC RG I or FG I.
93
 
558 ECL: Site Area Emergency Initiating Condition: Loss of all vital DC power for 15 minutes or longer.
Operating Mo de A pplicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director wi ll declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or wi ll likely be exceeded.
( 1)    Indicated voltage is less than 105 VOC on ALL 125 VOC busses 1(2)A0 1, 1(2)80 1, 1(2)C0 1, 1(2)DOI for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition invo lves a major fa ilure of plant funct ions needed for the protection of the public.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RG 1, FG l or SGS.
94
 
SA1 ECL: Alert Initiating Co ndition: Loss of a ll but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Powe r Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels :
Note: The emergency director will decla re the A lert promptly upon determining that 15 min utes has been exceeded, or will likel y be exceeded .
( I)    a.        AC power capabi li ty to BOTH 1(2)AA02 AND 1(2)BA03 is reduced to a single power source {I ubl..: SI) for 15 minutes or longer.                                      ( Commented [JRB30]: RAI 13 .b revision AN D
: b.        Any addi tional sing le powe r source fa ilure wi ll result in a loss of a ll AC powe r to SAFETY SYSTEMS.
Table S I Unit I                                    Unit 2 Unit Auxiliary Transformer INXAA          Unit Auxi liary Transformer 2NXAA Unit Auxiliary Transformer INXAB          Unit Auxiliary Transformer 2N XAB Reserve Aux ili ary Transformer INXRA      Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer INXRB        Reserve Auxi liary Transformer 2NXRB Diesel Generator IA                        Diesel Ge nerator 2A Diesel Generator I B                      Diesel Ge nerator 28 Standby Auxiliary Transfom1er ANXRA Standby Auxiliary Transforme r ANXRA Basis :
SAFETY SYSTEM: A system required for safe pl ant operation, coo ling down the pl ant and/or pl aci ng it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-re lated.
This IC describes a significant degradation of offs ite and onsite AC power sources (see Table S 1 above) where any add itio na l sing le failure wou ld resu lt in a loss of a ll AC power to SAFETY SYSTEMS. In thi s condition, the sole AC power source may be powering one, or more than one, train of safety-related equipme nt. This IC provides an escalation path from IC SU 1.
An " AC power source" is a source recognized in AOPs and EOPs and capable of supplying required power to an e me rgency bus. Some examples of this condition are presented be low.
* A loss of a ll offs ite power with a concurrent failure of a ll but one emergency power source (e.g., an onsite diesel gene rator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diese l generators) with a s ing le train of emergency buses being back-fed from the unit main generator.
95
* A loss of emergency power ources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes is the threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level uses IC SS I.
96
 
SA2 ECL: Alert Initiating Conditio n: UNPLANNED loss of Control Room ind ications for 15 minutes or longer with a significant transient in progress .
Operati ng Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emerge ncy Ac tion Levels:
Note: The emerge ncy director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    a.      An UNPLANNE D event res ults in the inability to monitor one or more of the foll owing parameters from within the contro l room fo r 15 minutes or longer.
Reactor Power RCS Level RCS Press ure In-Core/Core Exit Temperature Wide Range Level in at least one steam generator Steam Generator Main or Auxiliary Feed Water Flow AND
: b.      ANY of the following trans ient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor trip
* ECCS actuation Basis :
UNP LA      ED: A parameter change or an event that is not I ) the res ult of an intended evolution or 2) an expected plant res ponse to a transient. The cause of the parameter change or event may be known or unknown.
This JC addresses the difficulty associated with monitoring rap idl y changing pl ant conditions during a transient without the ab ility to obtain SAFETY SY TEM parameters fro m within the control room. During this condition, the margin to a potential fission product barrier challenge is reduced. It represents a potential substantial degradation in the leve l of plant safety.
As used in this EAL, an " inabi lity to monitor" means that values fo r one or more of the listed parameters cannot be determined from within the control room. This situation wo uld require a loss of all of the control room sources for the given parameter(s). For example, the reactor power level cannot be determined fro m any analog, digital and recorder source within the control room.
Various instrumcn1at1011 is also u .cJ to Jctcrmim, RCS I .ncl RV! IS, prcs,,u111cr level, J1g1tal or recorJcrs. A loss of all contrnl room sources for this parameter \\OulJ also apply.              [ Commented [JRB31] : RAJ 26 revision 97
 
An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with IO CFR 50.72 (and associated guidance in NUREG-1022) to determine ifan NRC event report is required. The event is 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, core cooling 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses !Cs FS I or IC RS 1.
98
 
SAS ECL: Alert Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Note: A manual 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.
Emergency Action Levels:
(1)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 plant safety. An emergency declaration is required even ifthe 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 consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
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 consoles (e.g., locally opening breakers). 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".
The plant response to the failure of an automatic or manual reactor trip 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 shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS I, an Alert declaration is appropriate for this event.
99
 
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F !Cs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
100
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 101
 
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 plant safety.
The first threshold for EAL 1.b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL 1.b 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 all 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 uses IC FS I or RS 1.
102
 
SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Eme rge ncy Action Levels:
Note: The emergency director wi ll declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( 1)    Loss of ALL offs ite AC power capability ~ labk: '>2) to BOTH 1(2)AA02 AN D            ( Commented [JRB32] : RAI 13.b revision l (2)BA03 fo r 15 minutes or longer.
Table S2 Unit I                                Unit 2 Reserve Auxiliary Transformer INXRA Reserve Aux iliary Transformer 2NXRA Reserve Auxiliary Transformer INXRB Reserve Aux iliary Transformer 2NXRB Standby Auxi liary Transformer ANXRA Standby Aux iliary Transformer ANXRA Basis:
This IC addresses a prolonged loss ofoffsite power. The loss ofoffsite power ources (see Table S2 above) renders the plant more vu lnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is avail able to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes is the threshold to exclude transient or momentary losses ofoffsite power.
Escalation of the emergency classification level uses IC SA I.
103
 
SU2 ECL: Notification of Unusual Event Initiating Cond it ion: UNPLANN ED loss of Control Room indications for 15 minutes or longer.
Ope rating Mode Applicability:          Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Eve nt promptly upon determining that 15 minutes has been exceeded, or wi ll likely be exceeded.
( I)      a. An UN PLA        ED event res ults in the inability to monitor one or more of the fo llowing parameters from within the control room for I 5 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Level in at least one steam generator Steam Generator Main or Auxiliarv Feed Water Flow Basis:
UNPLA        ED: A parameter change or an event that is not I) the re ult 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 .
This IC addresses the difficulty associated with monitoring normal plant conditions without the ab ility 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 plant safety.
As used in this EAL, an " inab ility to monitor" means that values fo r 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.
Vm*ious instrumentation i-; also us1:d lo dcl.:nninc RCS l1:\t'I- R\ LIS. pr1,;ssuri11,;r lcvd digital or recorders. A loss of all control room sourc.:s for this pararnctl'r ''ould abu apply                [ Commented [JRB33] : RAJ 26 revtsion 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 is reported if it significantly impairs the capabi lity to perform emergency assessments, particularl y those necessary to implement abnormal operating procedures, emergency operating proced ures, and emergency plan implementing procedures address ing emergency class ification, accident assessment, or protective action decision-making.
This EA L is focused on a se lected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RC heat removal. The loss of the abil ity to determine one or more of these parameters from within the control room is considered to be 104
 
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 reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
105
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2)
(1)    RE-48000 reading greater than 5.0 &#xb5;Ci/cc.
(2)    RCS coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits as indicated by ANY of the following:
Dose Equivalent 1-131 greater than I &#xb5;Ci/gm for greater than 48 hours Dose Equivalent 1-131 greater than Technical Specification figure 3.4.16-1 limits RCS specific activity greater than I 00/E &#xb5;Ci/gm gross radioactivity Basis:
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 plant safety.
Escalation of the emergency classification level uses ICs FA! or the Recognition Category R I Cs.
106
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)      RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)      RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)      Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
This IC addresses RCS leakage that could 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 plant safety.
EAL #1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL# I uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification is required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level uses I Cs of Recognition Category R or F.
i07
 
SUS ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (1 or 2)
Note: A manual 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.
(1)      a.      An automatic trip did not shutdown the reactor.
AND
: b.      A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a.      A manual trip did not shutdown the reactor.
AND
: b.      EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic trip is successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip 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 plant safety.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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 plant's decay heat removal systems.
If an initial manual reactor trip 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 trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
108
 
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 trip).
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".
The plant response to the failure of an automatic or manual reactor trip 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 SA5. Depending upon the plant response, escalation is also possible via IC FA!. Absent the plant conditions needed to meet either IC SA5 or FA 1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip 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 trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and will be evaluated.
* If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicable and no classification is warranted.
109
 
SU6 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 theNRC.
This IC will be assessed only when extraordinary means are being used 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).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OR Os of an emergency declaration. The OROs referred to here are the states of Georgia and South Carolina; Burke County, Georgia; Aiken County, South Carolina; Barnwell and Allendale, South Carolina; and the Savannah River Site in South Carolina.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
110
 
SU7 ECL: Notification of Unusual Event Initiating Condition: Failure to isolate containment or loss of containment pressure control.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: ( I or 2)
( I)    a.      Failure of containment to isolate when required by an actuation signal.
AND
: b.      ALL required penetrations are not closed within 15 minutes of the actuation signal.
(2)      a.      Containment pressure greater than ~21 . 5 psig.                                    ( Commented [lRB34]: RAJ 28 revision AND
: b.      Less than 4 CTMT fan coolers and one full train of CTMT spray is operating per design for 15 minutes or longer.
Basis:
This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety.
For EAL# I, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - will be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one fu ll train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event will escalate to a ite Area Emergency in accordance with IC FS I ifthere is a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
Ill
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ........... ... .. ..... ....................... ........ ................... .. .................................. ........... Alternating C urrent AOP .......... .......... ..... ... ... ... .. .... .............. ..................................... ...... Abnormal Operating Procedure ATWS ......... ........... ....... .. ......... .. ..... .. ............... ..................... Anticipated Transient Without Scram BLDG ....... ........ .............. .... .... ... ....... ....................... ...... ............ .... .. ........ .. ...... .. .... ..... .......... Building COE .. ..... ....... .. ........ ... ........... .... .... ...... .. ....... ...... ..... ... ... ...... ....... ......... .. Committed Dose Equivalent CFR ... ......... ........... ... .. ...... ... ..... ................................................... ......... Code of Federal Regul at ions CTMT/CNMT .. .... .. ............... .. ......... ... ... ......... ... .. .... ................. ...... ................. ... ... ... .. .. Containment CSF .. .. ...... .......... ........ .... ... .... ...................... ...... ... ..... ....... .. .. ....... ........ .. .... .. Critical Safety Function CSFST .. .......... .... ..... .. .... .. .. .. ......... ... ................ ........ .. ...... .. ....... Critical Safety Function Status Tree OBA ....... .. ............... .......................................... ....... ............................... ...... Design Basis Accident DC ... ... ...... ... ....... ... ... ........... ...... ....... .... .......... ... .. .. .. ... .. ..... ... ..... .... .... ..... ... .. ... ...... ... ... Direct Current EAL. ..................................................................... ............... ...................... Emergency Action Level ECCS ............. ... .. .... .. .................. ............... .... ..... .. ... ...... ...... ......... Emergency Core Cooling System ECL .... ...... .... ..... ..... .... ......... ... ... ..... ......... .... .. .. .... ... .... ....... ...... .. .... . Emergency C lassification Level EFFL ............................... ...................................... .. ..... ..... ..... ............ ........... ..... .................. Effluent ENN ........................... ... ............ ............ .... .. ................ ... .... ..... ..... Emergency Notification Network ENS .... .... ...... ... .. ..... .... .... ........ ... ........ ... .... ..... ................. ... ... ... ..... .. Emergency Not ification System EOF .... .. .. .... ... ... .... .. ..... .. .......... ... ........ ... ... ................. .... ................... Emergency Operations Facility EOP ........ .. .... ..... .... ...... .... ......................... ... ... .............. ... .. ........ .. .. Emergency Operating Procedure EPA ... .... .. ... ........ .. .. ... ... .. ......... .... ... .. .... ... .. .. .... ... ......................... Environmental Protection Agency FAA. .... ...... ... ........... ..... ... .... ... ... ... .. .... ..... ........ ... .... .. ..... ... ............. Federal Aviation Adm inistration FBI .... ...... ... ........ .......... ...... .... .... ..... ........ .... ...... .. ... .... ... ....... ..... ...... Federal Bureau of Investigation FEMA ........... ..... .... ........ .... .. ........ ...... ...... ... ... .... .... ...... ... Federal Emergency Management Agency FTS .......... ................... ............ .. ..... ........ .. .... ..... ... ... .. .. ... .. .. .... . Federal Telecommunications System GA .............. ...... ... ........... ... .. ......... .... ......................... ....... ........................... .... ......... ............. Georgia GE ...................................... .................................... .... .................................. ...... General Emergency HOO ........ .... ...... ... ......... ............. .. .......... .... ....................... Headquarters Operations Officer (NRC)
IC ................ ....... .... .............................. ................................ .... ... ... .............. ....... Initiating Condition ID ......... ... ............. .......................... .......................................................................... Inside Diameter ISFSI ..... .. .... ... ..... ... .... ... ....... .... ... ............... ........ ....... .. Independent Spent Fuel Storage Installation Keff ... .... ... .... ... ... .... .... ......... .. ... ... ................... .. .. .... .... ........ Effective Neutron Multiplication Factor mR, mRem, mrem, mREM ..... .... ............................................ .. ..... milli-Roentgen Equivalent Man NE! ... ... .... .. .. ... .. ......... ......... .... .... ........... .. ........ .... .... ........................ ... ... ..... N uclear Energy Institute NPP ........................................... ................ ......... .. ............ .. .... ........ ..... ... .......... Nuc lear Power Plant NRC ...... ... ... .. ... .. ..... ... .. ........ ..... ..... ......... ... .. .... ... ........ .. ................ Nuclear Regulatory Commiss ion ORAD ... .... ... ........... .... ......... ..... ... ....................... North American Aerospace Defense Command (NO)UE ............. ......... ..................... .......... ................. ...... .............. (Notiftcation Of) Unusual Event OBE ............. ................................... .. ..... .. .... ... ... .. .... ... ......... ............. ..... Operating Basis Earthquake OCA .. ... ...... ... ...... .... ... .. ...... ........ .... .. .. .. ..... ...... .. ..... .... ... ............ ....... ... ........ .Owner Controlled Area ODCM ...... ... .. .. .. ... ............ .. .... .... .... .... .. ........... .... ..... .. .. .... ............. Offsite Dose Calculation Manual PX .................................................................................... .. .. ................... .... .. Off Pren1ise E>.el'laRge ORO ......... ........................................ ..... ..... ...... .. .. ....... ....... ....... ...... Off-site Response Organization PA .......... ............. ... .... .. ... ...... ...... ...... ............... .. ... ... ........ .... ........ ...... ... .... ... ... ........... Protected Area PAG ............ .... ...... .. .... .................. ........... .............................................. Protective Action Guideline BX ............ ........ .... .. ............... ......... ..... ........ ............... .. ............... .............. PriYate BraRel'l e 1iel'l aRge A-I
 
PWR ........................................................................................................ Pressurized Water Reactor PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCDT ................................................................................................... Reactor Coolant Drain Tank RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RVLIS ...................................................................... Reactor Vessel Level Instrumentation System SAE ................................................................................................................. Site Area Emergency SC .............................................................................................................................. South Carolina SCBA ..................................................................................... Self-Contained Breathing Apparatus SG ........................................................................................................................... Steam Generator SI .............................................................................................................................. Safety Injection SJAE ............................................................................................................... Steam Jet Air Ejector SNC ....................................................................................................... Southern Nuclear Company SPDS ............................................................................................ Safety Parameter Display System TED E ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel VDC .................................................................................................................. Volts Direct Current VEGP ............................................................................................ Vogtle Electric Generating Plant VOIP ................................................................................................... Voice Over Internet Protocol WHT ................................................................................................................. Waste Holdup Tank A-2
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
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.
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.
Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The following are key terms necessary for overall understanding the NE! 99-01 emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL): 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 (I) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NODE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control ofVEGP security.
B-2
 
PROJECTILE: An object directed toward an NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
RUPTURE(D): The condition ofa steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
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.
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.
UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.
UNPLANNED: A parameter change or an event that is not I) 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.
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.
B-3
 
Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant Units 1 and 2; Edwin I. Hatch Nuclear Plant Units 1 and 2; Vogtle Electric Generating Plant Units 1 and 2; License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Responses to Requests for Additional Information ENCLOSURE 4 EAL SCHEMES CLEAN COPIES
 
Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Responses to Requests for Additional Information JOSEPH M. FARLEY NUCLEAR PLANT EAL SCHEME CLEAN COPIES
 
FARLEY NUCLEAR PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALVES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND ..................................................................................... 1 1.1 OPERATING REACTORS ****...*****.*****.*.***.***...*..*.........********.......**......*********.*...............*.... 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ***************......*........**.**.* 1 1.3 NRC ORDEREA-12-051 *******.*..........*.........**..........*********************.**************........*.*..**.**. 2 1.4 ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION .*********.....*.....***.*****. 3 1.5 IC AND EAL MODE APPLICABILITY .............................................................................. 3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 5 2.1 GENERAL CONSIDERATIONS .......................................................................................... 5 2.2 CLASSIFICATION METHODOLOGY ................................................................................. 6 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .......................................... 6 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ................................ 6 2.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................... 7 2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ................... 7 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS .....*****....*....**..........***.***..*....**......*..*........ 7 2.8 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................................. 7 2.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ................ 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS .......................... 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 26 5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 46 6  FISSION PRODUCT BARRIER ICS/EALS ..................................*...........*..*.....*.......... 49 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......... 65 8  SYSTEM MALFUNCTION ICS/EALS ..........**..*..........................................................*. 88 APPENDIX A - ACRONYMS AND ABBREVIATIONS ........................................................A-1 APPENDIX B - DEFINITIONS ...............................................................**..........................8-1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections for this document are:
* 10 CFR &sect; 50.47(a)(l)(i)
* 10 CFR &sect; 50.47(b)(4)
* 10 CFR &sect; 50.54(q)
* 10 CFR &sect; 50.72(a)
* 10 CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* 10 CFR &sect; 50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by regulatory guidance documents. Documents of particular relevance to NEI 99-01 include:
NUREG-0654/FEMA-REP-1, Criteriafor Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines 10 CFR &sect; 50.72 and&sect; 50.73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions also may be directed to the NEI Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NEI 99-0 I is applicable to licensees electing to use their I 0 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR &sect; 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR &sect; 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a 10 CFR &sect; 50.47 emergency plan.
The generic ICs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HUI covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. . Additionally, appropriate aspects of IC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI.
1
 
The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
1.3 NRC    ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to ModifY Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all U.S. nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;
* A display in an area accessible following a severe event; and
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To ModifY Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These 2
 
EALs are included within existing IC RA2, and new !Cs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the effectiveness of the plan. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.
1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode                R        c      E      F        H        s Power Operations          x                x      x        x        x Startup              x                x      x        x        x Hot Standby            x                x      x        x        x Hot Shutdown            x                x      x        x        x Cold Shutdown            x        x      x                x Refueling            x        x      x                x Defueled              x        x      x                x 3
 
Farley Units 1 and 2 Technical Specifications Table 1.1-1 provides the following operating mode definitions:
Reactivity      % Rated Average RCS Mode            Title          Condition      Thermal Power(a)    Temperature (&deg;F)
(Keff) 1    Power Operation          ~0.99            >5              NA 2      Startup                  ~0.99            :s 5            NA 3      Hot Standby              <0.99            NA            ~  350 4      Hot Shutdown<bJ          <0.99            NA        350 > Tavg > 200 5      Cold Shutdown<bJ          <0.99            NA            _::::200 6      Refueling<cJ                NA            NA              NA (a)  Excluding decay heat.
(b)  All reactor vessel head closure bolts fully tensioned.
(c)  One or more reactor vessel head closure bolts less than fully tensioned In addition to these defined modes, "Defueled" is also applicable to the Farley EAL scheme, consistent with NEI 99-01. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Farley EALs with no modifications from NEI 99-01.
When a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
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2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes and the informing basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.
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 provides guidance on implementing this requirement in NSIRIDPR-ISG-01, Interim Staff Guidance, Emergency Planningfor Nuclear Power Plants.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
Indication will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 will be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting in an expected event or condition that 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with 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 &sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
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Although the EALs have been developed to address a full spectrum of possible events and conditions that may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-0 I scheme provides the emergency director with the ability to classify events and conditions based upon judgment using EALs consistent with the emergency classification level (ECL) definitions (refer to Category H). The emergency director will determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.
2.2 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. An EAL(s) evaluation must be consistent with the related operating mode applicability and notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is declared in accordance with plant procedures.
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.
2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS In the event of multiple emergencies or conditions, 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, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
If two Alert EALs are met, an Alert will be declared.
Related guidance for classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events.
2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time an event or condition occurred, and prior to any plant or operator response, determines whether 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.
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2.5 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 will be made as though 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.
2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification. Termination criteria contained in procedure NMP-EP-110, Emergency Classification and Initial Actions shall be completed for an event to be terminated. At termination, on an event specific basis, the site will enter either normal operating conditions or a recovery condition with a recovery organization established for turnover from the ERO.
2.7 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 end 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. For example, an earthquake, or failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
2.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and EALs in this document employ time-based criteria that require 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, some transient conditions may cause an EAL to be met for a brief period of time. The following guidance will be applied to the classification of these conditions.
EAL momentarily met during expected plant response - When 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 applicable EAL is not considered met and the associated emergency declaration is not required. This example presents an illustration:
An A TWS occurs and the auxiliary feedwater system fails to automatically start.
Steam generator levels rapidly decrease and the plant enters an inadequate RCS 7
 
heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification will be based on the A TWS only.
It is important to note that the 15-minute emergency classification assessment period is not a "grace period" to delay a classification in order to perform 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 rapidly evolving situations in which an operator is able to take corrective action before the emergency director completes the review and necessary steps to make the emergency declaration. This provision ensures any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
2.9 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. Personnel could discover that an event or condition existed that met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. It may be the event or condition was not recognized at the time, or there was an error in the emergency classification process.
In these cases, no emergency declaration is warranted; but, the guidance in NUREG-1022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR &sect; 50.72 within one hour of the undeclared event or condition is discovered. The licensee will also notify appropriate State and local agencies in accordance with the agreed upon arrangements.
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3    ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                EMERGENCY RGl Release of          RSl Release of          RAl Release of          RUl Release of gaseous radioactivity    gaseous radioactivity    gaseous or liquid      gaseous or liquid resulting in offsite    resulting in offsite    radioactivity resulting radioactivity greater dose greater than 1,000  dose greater than I 00  in offsite dose greater than 2 times the mrem TEDE or 5,000      mrem TEDE or 500        than 10 mrem TEDE      ODCM limits for 60 mrem thyroid CDE.        mrem thyroid CDE.        or 50 mrem thyroid      minutes or longer.
Op. Modes: All          Op. Modes: All          CDE.                    Op. Modes: All Op. Modes: All RG2 Spent fuel pool      RS2 Spent fuel pool      RA2 Significant        RU2 UNPLANNED level cannot be          level at 131 feet.      lowering of water level loss of water level restored to at least 131 Op. Modes: All          above, or damage to,    above irradiated fuel.
feet for 60 minutes or                            irradiated fuel.        Op. Modes: All longer.                                          Op. Modes: All Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #1 will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Readings on ANY of the following radiation monitors greater than the reading shown below for 15 minutes or longer:
Steam Jet Air Ejector RE-I SC              130 &#xb5;Ci/cc (130 R/hr)
Plant Vent Stack RE-29B (NG)                0.8 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
Basis:
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 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified 10
 
in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used to determine the monitor reading threshold values in ICs RS 1 and RA 1. This protocol will maintain intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting.
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.
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RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 131 feet for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
( 1)    Spent fuel pool level cannot be restored to at least 131 feet for 60 minutes or longer.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. The spent fuel level instrument is located outside the Control Room but in close proximity. 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.
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RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1or2 or 3)
Notes:
* The emergency director will declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #1 will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Steam Jet Air Ejector RE-15C                13 &#xb5;Ci/cc (13 R/hr)
Plant Vent Stack RE-29B (NG)                0.08 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10 percent 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 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 monitor reading threshold values are 13
 
determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in ICs RG 1 and RA 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at 10 percent 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 uses IC RG 1.
14
 
RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at 131 feet.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Lowering of spent fuel pool level to 131 feet.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition stems from major failures of plant functions needed to protect the public that warrant a Site Area Emergency declaration. The spent fuel pool level instrument is located outside the Control Room but in close proximity.
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 uses via IC RGI or RG2.
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RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Notes:
* The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL # 1 will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Steam Jet Air Ejector RE-1 SC              1.3 &#xb5;Ci/cc (1.3 R/hr)
Plant Vent Stack RE-29B (NG)                0.008 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary.
(3)    Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary for one hour of exposure.
(4)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1 percent 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 plant safety as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
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Radiological effluent EALs are 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the re. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in res RG 1 and RS 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at 1 percent of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration_ of the I :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 uses re RS 1.
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RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
(1)    Uncovery of irradiated fuel in the REFUELING PATHWAY.
(2)    Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on ANY of the following radiation monitors:
Spent Fuel Pool Ventilation Monitor RE-25A ORB Spent Fuel Pool Area Radiation Monitor RE-5 Containment Purge Ventilation Monitor RE-24A ORB (3)    Lowering of spent fuel pool level to 140 feet (Level 2).
Basis:
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly. 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 plant safety.
This IC 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 ICE-HUI.
Escalation of the emergency is based on either Recognition Category R or CI Cs.
EAL#l This EAL escalates from RU2. 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), 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 will 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 18
 
indication of whether the fuel is actually uncovered. To the degree possible, readings will be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.
EAL#2 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 will be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Containment Purge Ventilation Monitors are not available during all modes.
EAL#3 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. The spent fuel pool level instrument is located outside the Control Room but in close proximity. This condition reflects a significant loss of spent fuel pool water inventory and is 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 uses !Cs RSI or RS2.
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RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
Note: 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.
(1)      Dose rate greater than 15 mR/hr on RE-IA, Control Room Radiation Monitor.
(2)      An UNPLANNED event results in radiation levels that prohibit or impede access to any Table HI plant rooms or areas:
Table Hl Mode      Room Name                          Room Number Electrical Penetration Room        334, 333, 347 I 2334,2333,2347 Hallway Outside Filter Room        312, 332/
3 IA I 2A MCC areas                2312,2332 Sample Room and Primary CHM labs  323, 324 I 2323,2324 Sample Room and Primary CHM labs  323, 324 I 2323,2324 4
RHRHxRoom.                        128/
2128 Basis:
UNPLANNED: A parameter change or an event that is not I) 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.
This IC addresses elevated radiation levels in certain plant rooms or 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 plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL #1, the area requiring continuous occupancy is the control room and the central alarm station. The central alarm station is in the control room envelope. The value of l 5mR/hr is derived from the GDC 19 value of 5 Rem in 30 days with adjustment for expected occupancy times.
For EAL #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 20
 
necessary at the time of the increased radiation levels. Access will 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.
* 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 to 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.
Escalation of the emergency classification level uses Recognition Category R, C or FI Cs.
21
 
RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
(1)    Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
Liquid Effluents Steam Generator Blowdown Effluent Line RE-23B      2.80 x 103 cpm Gaseous Effluents Steam Jet Air Ejector RE-15                        3.5 x 102 cpm Plant Vent Gas R-14                                      3.2 x 104 cpm RE-22                                      4.0 x 102 cpm RE-29B (NG)                                8.9 x 104 &#xb5;Ci/cc (2)    Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
Li uid Radwaste Effluent Line RE-18 Plant Vent Gas R-14 (3)    Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
This IC addresses a potential decrease in the level of plant safety 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.
22
 
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment indicates degradation in these features and/or controls.
Radiological effluent EALs are 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 will not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EAL #3 - 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 uses IC RAl.
23
 
RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Personnel report of low water level Annunciator EH2 "SFP LVL HI/LO" AND
: b.      UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
RE-5 in the spent fuel pool building RE-2 in containment RE-27 A ORB in containment Basis:
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
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.
This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition can be a precursor to a more serious event and indicates a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of plant safety.
A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications 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 will be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor 24
 
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 with Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level uses IC RA2.
25
 
4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY              EMERGENCY CGl Loss ofRPV          CSl Loss ofRPV        CAl Loss ofRPV          CUl UNPLANNED inventory affecting      inventory affecting    inventory.              loss of RPV inventory fuel clad integrity with core decay heat        Op. Modes: Cold          for 15 minutes or containment              removal capability. Shutdown, Refueling      longer.
challenged.              Op. Modes: Cold                                Op. Modes: Cold Op. Modes: Cold          Shutdown, Refueling                            Shutdown, Refueling Shutdown, Re.fueling CA2 Loss of all          CU2 Loss of all but offsite and all onsite  one AC power source AC power to              to emergency buses for emergency buses for      15 minutes or longer.
15 minutes or longer. Op. Modes: Cold Op. Modes: Cold          Shutdown, Refueling, Shutdown, Refueling,    Defueled Defueled CA3 Inability to        CU3 UNPLANNED maintain the plant in    rise in RCS cold shutdown.          temperature.
Op. Modes: Cold        - Op. Modes: Cold Shutdown, Refueling      Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling CVS Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refueling, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Cold Shutdown, Refueling 26
 
CG1 ECL: General Emergency Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged.
Operating Mode AppHcability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(1)    a.        RPV level less than ANY of the following for 30 minutes or longer:
* 0% RVLIS (Mode 5)
* 119' Temporary Level Indicator (Mode 6)
AND
: b.      ANY indication form the Containment Challenge Table Cl.
(2)    a.      Reactor vessel level cannot be monitored for 30 minutes or longer.
AND
: b.      Core uncovery is indicated by ANY of the following:
* Containment High Range Radiation Monitor RE27A or 27B reading greater than or equal to 100 R/hr.
* Erratic source range monitor indication
* UNPLANNED rise in Containment Sump, or Reactor Coolant Drain Tank (RCDT), or Waste Holdup Tank (WHT) levels of sufficient magnitude to indicate core uncovery AND
: c.      ANY indication from the Containment Challenge Table C 1.
Containment Challenge Table Cl
* Greater CONTAINMENT CLOSURE not established*
* UNPLANNED  than or equal to 6 % H2 exists inside containment
          ** If CONTAINMENTincrease    in containment pressure CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
Basis:
CONTAINMENT CLOSURE: Per FNP-1(2)-STP-18.4, "Containment Integrity Verification and Closure".
27
 
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.
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 Emerge~cy 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 containment is challenged.
In EAL 2.b, 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 actions to terminate leakage, recover inventory control or 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 indicate leakage from the RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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.
28
 
CS1 ECL: Site Area Emergency Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(1)    a. CONTAINMENT CLOSURE not established.
AND
: b. RVLIS (Mode 5) level less than 16% (121 '0").
(2)    a. CONTAINMENT CLOSURE established.
AND
: b. RPV level less than ANY of the following:
* 0% RVLIS (Mode 5)
* 119' Temporary Level Indicator (Mode 6)
(3)    a. RPV level cannot be monitored for 30 minutes or longer.
AND
: b. Core uncovery is indicated by ANY of the following:
* Containment High Range Radiation Monitor RE27A or 27B reading greater than or equal to 100 R/hr
* Erratic source range monitor indication
* UNPLANNED rise in Containment Sump, or Reactor Coolant Drain Tank (RCDT), or Waste Holdup Tank (WHT) levels of sufficient magnitude to indicate core uncovery Basis:
CONTAINMENT CLOSURE: Per FNP-1(2)-STP-18.4, "Containment Integrity Verification and Closure".
UNPLANNED: A parameter change or an event that is not I) 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.
29
 
This IC addresses a significant and prolonged loss of RPV 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 to protect the public and 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 level specified in EAL 1.b represents a level in the RPV that is 6 inches below the bottom ID of the reactor vessel penetration. This level is lower than the RPV monitoring capability of RCS level instrumentation and therefore must be monitored using RVLIS. This level will only be observable in Mode 5 with RVLIS operable. In Mode 6, when RVLIS is not operable, this IC should be evaluated using EAL #2.b or 3.
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 specified RCS/reactor vessel levels of EAL l .b and 2.b reflect that without CONTAINMENT CLOSURE established, there is a higher probability of a fission product release to the environment.
In EAL 3.a, 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 actions to terminate leakage, recover inventory control or 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 indicate leakage from the RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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 uses IC CG 1 or RG 1.
30
 
CA1 ECL: Alert Initiating Condition: Loss ofRPV inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss ofRPV inventory as indicated by level less than 122' 11".
(2)    a.      RPV level cannot be monitored for 15 minutes or longer AND
: b.      UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tank (WHT) levels due to a loss ofRPV inventory.
Basis:
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.
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 EAL #1, a lowering of water level below 122' 11" indicates that operator actions have not been successful in restoring and maintaining RPV water level. The 122' 11" level specified in EAL # 1 is the minimum RCS level for RHR operation provided in procedure for mid loop operations. Below this level, loss of RHR pump net positive suction head (NPSH) may occur resulting in a loss of decay heat removal capability. 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.
Although related, EAL #1 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 residual heat removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, 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 indicate leakage from the RPV.
31
 
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1.
If the RPV inventory level continues to lower, then escalation to Site Area Emergency uses IC CSL 32
 
CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC Power (Table Sl) to BOTH 4I60V ESF busses I (2)F AND 1(2)G for 15 minutes or longer.
Table Sl Unit 1                                        Unit2 Start-up Aux XFMR IA                            Start-up Aux XFMR 2A Start-up Aux XFMR lB                            Start-up Aux XFMR 2B Diesel Generator I-2A                          Diesel Generator 1-2A Diesel Generator 1B                            Diesel Generator 2B Diesel Generator IC                            Diesel Generator 1C Diesel Generator 2C                            Diesel Generator 2C Basis:
This IC addresses a total loss of AC power (see Table SI above) 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.
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 emergency 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. When in these modes, this condition represents an actual or potential substantial degradation of the level of plant safety.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses IC CS I or RS I.
33
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
                                                                ~~~~~~~~~~~~~~
(1)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F for greater than the duration specified in Table C2.
Table C2: RCS Heat-up Duration Thresholds
                                                                ~~~~~~~~~~~~~----<
RCS Status            Containment Closure Status            Heat-up Duration
                                                                ~~~~t--~~~~~~~~----1 Not Intact                    Not Established
                                                                ~~~~t--~~~~~~~~----1 0 minutes (or at reduced inventory)              Established                    20 minutes*
Intact Not applicable                  60 minutes*
(but not at reduced inventory)
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)    UNPLANNED RCS pressure increase greater than 10 psig. (This EAL does not apply during water-solid plant conditions).
Basis:
CONTAINMENT CLOSURE: Per FNP-1 (2)-STP-18.4, "Containment Integrity Verification and Closure".
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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment 34
 
atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, 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 will allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS 1 or RS 1.
35
 
CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 36
 
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 plant safety.
The first threshold for EAL l .b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL 1.b 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 all 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 uses IC CS 1 or RS 1.
37
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss ofRPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note:    The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    UNPLANNED loss ofreactor coolant results in RPV level less than a required lower limit for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED rise in Containment sump, Reactor Coolant Drain Tank, or Waste Holdup Tank levels.
Basis:
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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL # 1 recognizes that the minimum required zyv 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.
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.
EAL #2 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 38
 
sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they indicate leakage from the RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CA 1 or CA3.
39
 
CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The. emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      AC power capability to BOTH 4 I 60V ESF busses I (2)F AND I (2)G is reduced to a single power source (Table SI) for I 5 minutes or longer.
AND
: b.      Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                                  Unit2 Start-up Aux XFMR IA                    Start-up Aux XFMR 2A Start-up Aux XFMR IB                    Start-up Aux XFMR 2B Diesel Generator 1-2A                  Diesel Generator 1-2A Diesel Generator IB                      Diesel Generator 2B Diesel Generator IC                      Diesel Generator IC Diesel Generator 2C                      Diesel Generator 2C Basis:
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.
This IC describes a significant degradation of offsite and onsite AC power sources where 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. When in these modes, this condition is considered to be a potential degradation of the level of plant safety.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus (see Table SI above). Examples of this condition include:
40
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from an offsite power source.
Fifteen minutes is the 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.
41
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      UNPLANNED increase in RCS temperature to greater than 200 &deg;F.
(2)    Loss of ALL RCS temperature AND RPV level indication for 15 minutes or longer.
Basis:
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.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level. It represents a potential degradation of the level of plant safety. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the emergency director will 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.
EAL # 1 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, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators are 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 is the threshold to exclude transient or momentary losses of indication.
42
 
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
43
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105 VDC on Technical Specification required 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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. This condition is considered to be a potential degradation of the level of plant safety.
As used in this EAL, "required" means the vital DC busses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, ifTrain A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
44
 
CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (1or2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will be assessed only when extraordinary means are used 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).
EAL # 1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Alabama, Georgia, and Florida; Houston and Henry Counties, Alabama; and Early County, Georgia.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
I 45
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HUl Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 46
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY of the values listed in Table E 1.
Table El Location of Dose Rate                  Total Dose Rate (Neutron+ Gamma mR/hr)
HI-TRAC 125 Side-Mid-height                        1360 Top                                260 HI-STORM 100 Side - 60 inches below mid-height                340 Side - Mid- height                        350 Side - 60 inches above mid-height                170 Center of lid                          50 Middle of top lid                        60 Top (outlet) duct                      160 Bottom (inlet) duct                      460 Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
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 that could cause challenges in removing the cask or fuel from storage.
The existence of "damage" is determined by radiological survey. The radiation reading values listed in the table represent 2 times the site-specific cask specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical specification multiple of "2 times" 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 a?d not the magnitude of the associated dose or dose rate. It is 47
 
recognized that in the case of extreme damage to a loaded cask, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under ICs HUI and HAI.
48
 
6 FISSION PRODUCT BARRIER ICS/EALS                        LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS FUEL CLAD              RCS                      CONTAINMENT Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY                                            ~
Loss of any two barriers and Loss or Potential Loss of the third barrier.                                L . ./~sofatleast2
                                                                                              ',, Barriers?
                                                                                                                  -  YES-    ffi! - Loss of ANY Two Bmicrs Alill Loss or Potential Loss of Third Barrier FGl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown                                                                    ~------NO--~
SITE AREA EMERGENCY POTENTIAL                                        POTENTIAL Loss or Potential Loss of any two barriers. LOSS LOSS LOSS LOSS FUEL CLAD                                      CONTAINMENT FSl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown ALERT                                                                                              FSI - Loss or Potential Loss of ANY Two Barriers Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
FAl                                                          POTENTIAL          POTENTIAL LOSS                LOSS Op. Modes: Power Operation, Hot Standby,              LOSS                LOSS FUEL CLAD              RCS Startup, Hot Shutdown FA I - ANY Loss or ANY Potential Loss of EITHER Fuel Clad !IB RCS 49
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGl GENERAL EMERGENCY                        FSl SITE AREA EMERGENCY                              FAlALERT Loss of any two barriers and Loss or          Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either Potential Loss of the third barrier.                                                      the Fuel Clad or RCS barrier.
  .          Fuel,Clad Barrier                                  RCS: Barrier                                Containment Barrier LOSS            POTENTIAL LOSS                  LOSS            POTENTIAL LOSS                LOSS              POTENTIAL LOSS
: 1. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage Not Applicable        A. CORE COOLING        A. An automatic or        A. Operation of a      A. A leaking or        Not Applicable CSF-ORANGE            manual ECCS                standby charging      RUPTURED SG is entry conditions      actuation is required      pump is required      FAULTED outside met.                    by EITHER of the          by EITHER of the      of containment.
following:                following:
* UNISOLABLE
* UNISOLABLE RCS leakage                RCS leakage
* SGtube
* SG tube RUPTURE.                    leakage.
OR B. RCS INTEGRITY CSF - RED entry conditions met 50
 
Fuel Clad Barrier                                  RCS Barrier                                  Containment Barrier LOSS              POTENTIAL LOSS                  LOSS            POTENTIAL LOSS                LOSS              POTENTIAL LOSS
: 2. Inadequate Heat Removal                      2. Inadequate Heat Removal                      2. Inadequate Heat Removal A. CORE COOLING A. CORE COOLING                  Not Applicable          A. HEAT SINK CSF -      Not Applicable          A. CORE COOLING CSF - RED entry          CSP-ORANGE                                      RED entry                                      CSF - RED entry conditions met            entry conditions                                conditions met.                                conditions met for 15 met                                                                                              minutes or longer OR                                          NOTE: Heat Sink CSF B. HEAT SINK CSF -                              should not be RED entry                                    considered RED if total conditions met                                AFW flow is less than 395 gpm due to NOTE: Heat Sink CSF                              operator action.
should not be considered RED if total AFW flow is less than 395 gpm due to operator action.
: 3. RCS Activity I Containment Radiation          3. RCS Activity I Containment Radiation          3. RCS Activity I Containment Radiation A. Containment          Not Applicable          A. Containment          Not Applicable          Not Applicable          A. Containment radiation monitor                                radiation monitor                                                        radiation monitor RE-27 AorB                                        RE-2 greater than 1                                                      RE-27 A orB greater than 600                                  R/Hr                                                                      greater than 8000 R/Hr.                                            OR                                                                        R/Hr.
OR                                                Containment B. Indications that                                  radiation monitor reactor coolant                                  RE-7 greater than activity is greater                              500 mR/Hr.
than 300 &#xb5;Ci/gm dose equivalent I-131.
51
 
Fuel Clad Barrier                        RCS Barrier                                Containment Barrier LOSS            POTENTIAL LOSS          LOSS            POTENTIAL LOSS            LOSS              POTENTIAL LOSS
: 4. Containment Integrity or Bypass    4. Containment Integrity or Bypass      4. Containment Integrity or Bypass Not Applicable          Not Applicable  Not Applicable          Not Applicable  A. Containment isolation    A. CONTAINMENT CSF is required                RED entry conditions AND                        met.
EITHER of the              OR following:              B. Containment Hydrogen
* Containment              concentration greater integrity has been    than 6%
lost based on        OR I
Emergency
: c. I. Containment Director pressure greater than judgment.
27 psig.
* UNISOLABLE                  AND pathway from the containment to        2. Less than one the environment          CTMT fan coolers exists.                  and one full train of OR                            CTMT Spray is operating per design B. Indications of RCS              for 15 minutes or leakage outside of              longer.
containment as indicated by alarms on any of the following instruments:
* RE-10
* RE-14
* RE-21
* RE-22 Note: Increases in sump levels, temperatures, pressures, flow rates and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
: 5. Other Indications                    5. Other Indications                      5. Other Indications Not aoolicable          Not applicable  Not applicable          Not applicable  Not applicable              Not applicable 52
 
Fuel Clad Barrier                                      RCS -Barrier                                    Containment Barrier LOSS            POTENTIAL LOSS                    LOSS                POTENTIAL LOSS                  LOSS              POTENTIAL LOSS
: 6. Emergency Director Judgment                    6. Emergency Director Judgment                      6. Emergency Director Judgment A. ANY condition in    A. ANY condition in      A. ANY condition in the A. ANY condition in          A. ANY condition in      A. ANY condition in the the opinion of the      the opinion of the        opinion of the              the opinion of the      the opinion of the        opinion of the emergency              emergency director        emergency director          emergency director      emergency director        emergency director director that          that indicates            that indicates loss of      that indicates          that indicates loss of    that indicates indicates loss of        potential loss of the    the RCS barrier.            potential loss of the    the containment          potential loss of the the fuel clad          fuel clad barrier.                                    RCS barrier.            barrier.                  containment barrier.
barrier.
53
 
Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The fuel clad barrier consists of the cladding material that contains the fuel pellets.
: 1. RCS or SG Tube Leakage There is no Loss threshold associated with RCS or SG Tube Leakage.
Potential Loss I .A This condition indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
: 2. Inadequate Heat Removal Loss 2.A This condition indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Potential Loss 2.A This condition indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
Potential Loss 2.B NOTE: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 gpm due to operator action.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the fuel clad barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS barrier potential loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity/Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300&#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater 54
 
than that expected for iodine spikes and corresponds to an approximate range of 2 percent to 5 percent 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 Joss threshold 3 .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.
Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2 percent to 5 percent 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.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no potential loss threshold associated with RCS activity/containment radiation.
: 4. Containment Integrity or Bypass Not applicable (included for numbering consistency)
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
55
 
RCS BARRIER THRESHOLDS:
The RCS barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the containment barrier loss threshold I .A will also be met.
Potential Loss 1.A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS (SI) actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FA UL TED outside of containment, the declaration escalates to a Site Area Emergency since the containment barrier loss threshold I .A will also be met.
Potential Loss 1.B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
: 2.      Inadequate Heat Removal There is no loss threshold associated with inadequate heat removal.
56
 
Potential Loss 2.A NOTE: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 g m due to o erator action.
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to fuel clad barrier potential loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity/Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for fuel clad barrier loss threshold 3.A since it indicates a loss of the RCS barrier only.
There is no potential loss threshold associated with RCS activity/containment radiation.
: 4. Containment Integrity or Bypass Not applicable (included for numbering consistency)
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
57
 
CONTAINMENT BARRIER THRESHOLDS:
The containment barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost second~ry side isolation valve. Containment barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      RCS or SG Tube Leakage Loss l.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss l .A and Loss 1.A, respectively. This condition represents a bypass of the containment barrier.
FAUL TED is a defined term within the NEI 99-0 I methodology. This determination is not necessarily dependent on entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FA ULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e.,-RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (similar to a FAULTED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may
* occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
58
 
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAUL TED Outside of Containment?
P-to-S Leak Rate                      Yes                          No Less than or equal to 25 gpm          No classification            No classification Greater than 25 gpm                Unusual Event per SU4      Unusual Event per SU4 Requires operation of a Site Area Emergency standby charging (makeup)                                          Alert per FA 1 per FSI pump (RCS barrier potential loss)
Requires an automatic or            Site Area Emergency Alert per FA 1 manual ECCS (SI) actuation                  per FSl (RCS barrier loss)
There is no potential loss threshold associated with RCS or SG Tube Leakage.
: 2. Inadequate Heat Removal There is no loss threshold associated with inadequate heat removal.
Potential Loss 2.A This condition represents an IMMINENT core melt sequence that, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS barrier and the fuel clad barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the containment barrier.
The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing or ifreactor vessel level is increasing. Whether the procedure(s) will be effective should be apparent within 15 minutes. The emergency director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it 59
 
is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
: 3. RCS Activity/Containment Radiation There is no loss threshold associated with RCS activity/containment radiation.
Potential Loss 3 .A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent 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 percent 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 containment that would then escalate the emergency classification level to a General Emergency.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. There may be accident and release conditions that simultaneously meet both thresholds 4.A. l and 4.A.2.
4.A.1 - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the emergency director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment).
Two simplified examples are provided in the middle piping run of Figure 6-F-l. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example is a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FA UL TED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of 60
 
containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
4.A.2- Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
See a simplified example in the top piping run of Figure 6-F-1. The inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
A simplified example is shown in the bottom piping run of Figure 6-F-l. Leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. Ifthere is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.1 to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R I Cs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using loss threshold I .A.
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Loss 4.B Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment will be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly.
However, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
In the simplified example in the middle piping run of Figure 6-F-l, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. l to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS loss and/or potential loss threshold I .A to be met.
Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the containment barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and fuel clad barriers would already be lost. This threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Potential Loss 4.B 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 potential loss of the containment barrier.
Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment because containment heat removal/depressurization 62
 
systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
: 5. Other Indications Not applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the containment barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
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Figure 6-F-1: PWR Containment Integrity or Bypass Examples I- - - - - - - - - - :
:I!.,. ________
Effluent_ I' Inside Containment              Auxiliary Building      , Monitora Damper I
: Process
: Monitor :
I              I
                                                    ~---*--~                      Closed Cooling Water System Cooling 64
 
7    HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT              UNUSUAL EVENT EMERGENCY                EMERGENCY HGl HOSTILE              HSl HOSTILE                HAl    HOSTILE          HUl    Confirmed ACTION resulting in      ACTION within the          ACTION within the        SECURITY loss of physical control PROTECTED AREA.            OWNER                    CONDITION or threat.
of the facility.        Op. Modes: All              CONTROLLED AREA          Op. Modes: All Op. Modes: All                                      or airborne attack threat within 30 minutes.
Op. Modes: All HU2    Seismic event greater than OBE levels.
Op. Modes: All HU3    Hazardous event.
Op. Modes: All HU4    FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HAS    Gaseous release impeding access to equipment necessary for normal plant operations, cooldown, or shutdown.
Op. Modes: All HS6    Inability to        HA6    Control Room control a key safety        evacuation resulting in function from outside      transfer of plant control the Control Room.          to alternate locations.
Op. Modes: All              Op. Modes: All HG7 Other conditions    HS7    Other conditions    HA7    Other conditions  HU7    Other conditions exist which in the      exist which in the          exist which in the        exist which in the judgment of the          judgment of the            judgment of the          judgment of the emergency director      emergency director          emergency director        emergency director warrant declaration of a warrant declaration of a    warrant declaration of    warrant declaration of a General Emergency.      Site Area Emergency.        an Alert.                NOUE.
Op. Modes: All          Op. Modes: All              Op. Modes: All            Op. Modes: All 65
 
HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the on shift Security Captain or designee.
AND
: b.      EITHER of the following has occurred:
: 1.      ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMINENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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 supervision and the control room is essential for proper classification of a security-related event.
66
 
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 [and Independent Spent Fuel Storage Installation Security Program].
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 Security Plan.
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HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a General Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a General Emergency.
68
 
HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.
Operating Mode Applicability: All Emergency Action Levels:
(1)      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the on shift Security Captain or designee.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses the occurrence of a HOSTILE ACTiON within the PROTECTED AREA (PA). This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between Security shift supervision and the control 'room is essential for proper classification of a security-related event.
Security plans and terminology are based on the guidance provided byNEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
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 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 &sect; 73.71or10 CFR &sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate security-sensitive information. This includes information that may be 69
 
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 Security Plan.
Escalation of the emergency classification level uses IC HGI.
70
 
HS6 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      An event has resulted in plant control being transferred from the control room to the remote shutdown panel.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* Core cooling
* RCS heat removal Basis:
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 "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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level uses IC FG 1 or CG I.
71
 
HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Site Area Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, 1) toward site personnel or equipment that could lead to the likely failure of or, 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because cm;iditions exist that are believed by the emergency director to fall under the emergency classification level description for a Site Area Emergency.
72
 
HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
(1)      A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the on shift Security Captain or designee.
(2)      A validated notification from NRC of an aircraft attack threat within 3 0 minutes of the site.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control ofFNP security.
This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA (OCA) 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 (PA),
or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program}.
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, 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.
73
 
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71or10 CFR &sect; 50.72.
EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA (OCA).
EAL #2 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 station procedures.
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 (OCA) 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, will 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 Security Plan.
Escalation of the emergency classification level uses IC HS 1.
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HAS ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(1)    a.      Release of a toxic, corrosive, asphyxiant or flammable gas into any Table Hl plant rooms or areas:
Table Hl Mode      Room Name                              Room Number Electrical Penetration Room            334, 333, 347 I 2334,2333,2347 Hallway Outside Filter Room            312,332/
3                                              2312,2332 IA I 2A MCC areas Sample Room and Primary CHM labs      323, 324 I 2323,2324 Sample Room and Primary CHM labs      323, 324 I 2323,2324 4
RHRHxRoom.                            128/
2128 AND
: b.      Entry into the room or area is prohibited or impeded.
Basis:
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 plant safety.
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 director's 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.
75
 
Access should be considered as impeded if extraordinary 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 de~laration 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 4.
* The gas release is a planned activity that includes compensatory measures to 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.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19 percent, 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.
Escalation of the emergency classification level uses Recognition Category R, Cor F ICs.
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HA6 ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(1)    An event has resulted in plant control being transferred from the control room to the remote shutdown panel.
Basis:
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 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 uses IC HS6.
77
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOST AGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for an Alert.
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HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3)
(1)    A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the on shift Security Captain or designee.
(2)    Notification of a credible security threat directed at FNP.
(3)      A validated notification from the NRC providing information of an aircraft threat.
Basis:
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and 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 IO CFR &sect; 73.71 or IO CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGI.
Timely and accurate communications between Security Shift Supervision 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 OROs.
Security plans and terminology are based on the guidance provided by NEI 03-I2, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program}.
EAL #I references site security force 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 I 0 CFR &sect; 2.39 information.
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is 79
 
assessed in accordance with station procedures.
EAL #3 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HAI.
80
 
HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels.
Operating Mode Applicability: All Emergency Action Levels:
(1)    Seismic event greater than Operating Basis Earthquake (OBE) as indicated by seismic switch activation with the seismic system computer indicating EITHER of the following:
* Cumulative Absolute Velocity (CAY) greater than 0.160 g-sec AND Spectral Acceleration greater than 0.200 g.
* Cumulative Absolute Velocity (CAY) greater than 0.160 g-sec AND Spectral Velocity greater than 15.240 cm/sec.
Basis:
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 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 plant safety.
Event verification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should readily be felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are 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 or check of 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 uses IC CA6 or SA9.
81
 
HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
(1)    A tornado strike within the PROTECTED AREA.
(2)    Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
(3)    Movement of personnel within the PROTECTED AREA is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
(4)    A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
(5)    Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours.
Basis:
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of plant safety.
EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 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.
EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA (PA).
82
 
EAL #4 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, or dam failure, etc., or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses phenomena of the hurricane based on the severe weather mitigation procedure.
Escalation of the emergency classification level is based on I Cs in Recognition Categories A, F, sore.
83
 
HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Note: The emergency director will declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(1)    a.      A FIRE is NOT extinguished within IS-minutes of ANY of the following FIRE detection indications:
* 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
: b.      The FIRE is located within ANY Table H2 rooms or areas.
(2)    a.      Receipt of a single fire alarm (i.e., no other indications of a FIRE).
AND
: b.      The FIRE is located within ANY Table H2 rooms or areas AND
: c.      The existence of a FIRE is not verified within 30-minutes of alarm receipt.
(3)    A FIRE within the plant PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
(4)    A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
Table H2 Auxiliary Building Diesel Generator Building Service Water Intake Structure (SWIS)
Containment RWST CST 84
 
Basis:
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 (PA): The area that encompasses all controlled areas within the security protected area fence.
This JC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of plant safety.
EAL#l 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 include a drop in fire main pressure, automatic activation of a suppression system.
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.
EAL#2 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 alann 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 EAL #1 is immediately applicable, and the emergency must be declared ifthe 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.
EAL#3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA (PA) not extinguished within 60-minutes may also potentially degrade the level of plant safety.
EAL#4 85
                                                                                                      -~I
 
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 not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
Basis-Related Requirements from Appendix R Appendix R to I 0 CPR 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.
Appendix R to I 0 CPR 50, requires, among other considerations, the use of I-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 EAL #2, the 30-minutes to verify a single alarm is well within this worst-case I-hour time period.
Depending upon the plant mode at the time of the event, escalation of the emergency classification level uses IC CA6 or SA9.
86
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a NOUE.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a NOUE.
87
 
8    SYSTEM MALFUNCTION ICS/EALS GENERAL                  SITE AREA ALERT              UNUSUAL EVENT EMERGENCY                  EMERGENCY SGl Prolonged loss of      SSl Loss of all offsite  SAl Loss of all but one SUl Loss of all offsite all offsite and all onsite and all onsite AC power  AC power source to        AC power capability to AC power to emergency      to emergency buses for    emergency buses for 15    emergency buses for 15 buses.                    15 minutes or longer. minutes or longer.        minutes or longer.
Op. Modes: Power          Op. Modes: Power          Op. Modes: Power          Op. Modes: Power Operation, Startup, Hot    Operation, Startup, Hot  Operation, Startup, Hot  Operation, Startup, Hot Standby, Hot Shutdown      Standby, Hot Shutdown    Standby, Hot Shutdown    Standby, Hot Shutdown SA2 UNPLANNED            SU2 UNPLANNED loss of Control Room      loss of Control Room indications for 15        indications for 15 minutes or longer with a  minutes or longer.
significant transient in  Op. Modes: Power progress.                Operation, Startup, Hot Op. Modes: Power          Standby, Hot Shutdown Operation, Startup, Hot Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS Inability to          SAS Automatic or          SUS Automatic or shutdown the reactor      manual trip fails to      manual trip fails to causing a challenge to    shutdown the reactor, and shutdown the reactor.
core cooling or RCS heat  subsequent manual        Op. Modes: Power removal.                  actions taken at the      Operation Op. Modes: Power          reactor control consoles Operation                are not successful in shutting down the reactor.
Op. Modes: Power Operation 88
 
GENERAL                  SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY SU6 Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Failure to isolate containment or loss of containment pressure control.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS Loss of all AC        SSS Loss of all vital and vital DC power        DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 89
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 4 l 60V ESF busses 1(2)F AND 1(2)G.
AND
: b.      EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* CORE COOLING CSF - RED conditions met.
Basis:
This IC addresses a prolonged loss of all _power sources to AC emergency busses. 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. 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 AC emergency 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 emergency bus will be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success will 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.
90
 
SGS ECL: General Emergency Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 4160V ESF busses 1(2)F AND 1(2)G for 15 minutes or longer.
AND
: b.      Indicated voltage is less than 105 VDC on ALL 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. 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 loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
91
 
551 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC power to BOTH 4160V ESF busses 1(2)F AND 1(2)0 for 15 minutes or longer.
Basis:
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 is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses !Cs RG 1, FG 1 or SG 1.
92
 
555 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.
Operating Mode Applicability: Power Operation Emergency Action Levels:
Note: Heat Sink CSF should not be considered RED if total AFW flow is less than 395 gpm due to operator action.
(1)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      All manual actions to shutdown the reactor have been unsuccessful.
AND
: c.      EITHER of the following conditions exist:
* Core Cooling CSF - RED conditions met
* Heat Sink CSF - Red conditions met Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 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 because the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level uses IC RG 1 or FG 1.
93
 
558 ECL: Site Area Emergency Initiating Condition: Loss of all vital DC power for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105 VDC on ALL 125 VDC vital busses for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RG 1, FG 1 or SGS.
94
 
SA1 ECL: Alert Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      AC power capability to BOTH 4160V ESF busses 1(2)F AND 1(2)G is reduced to a single power source (Table SI) for 15 minutes or longer.
AND
: b.      Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                                    Unit2 Start-up Aux XFMR IA                      Start-up Aux XFMR 2A Start-up Aux XFMR lB                      Start-up Aux XFMR 2B Diesel Generator 1-2A                    Diesel Generator 1-2A Diesel Generator I B                      Diesel Generator 2B Diesel Generator 1C                      Diesel Generator 1C Diesel Generator 2C                      Diesel Generator 2C Basis:
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.
This IC describes a significant degradation of offsite and onsite AC power sources where 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 SUI.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus (see Table SI above). Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
95
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency busses being back-fed from an offsite power source.
Fifteen minutes is the threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level uses IC SS 1.
96
 
SA2 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Levels in at least one steam generator Steam Generator Auxiliary Feed Water Flow AND
: b.      ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor trip
* ECCS actuation Basis:
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.
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 product barrier challenge is reduced. It represents a potential substantial degradation in the level of plant safety.
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. Various instrumentation is also used to determine RCS Level -RVLIS, pressurizer level, digital or recorders. A loss of all Control Room sources for this parameter would also apply.
97
 
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 is 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 ofreactivity control, core cooling 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses !Cs FSl or IC RSI.
98
 
SA5 ECL: Alert Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Emergency Action Level:
(1)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 plant safety. 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 consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
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 consoles (e.g., locally opening breakers). 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".
The plant response to the failure of an automatic or manual reactor trip 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 shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS 1, 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
99
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 100
 
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 plant safety.
The first threshold for EAL l .b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL l .b 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 all 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 uses IC FS 1 or RS 1.
101
 
SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      Loss of ALL offsite AC power capability (Table S2) to BOTH 4160V ESP busses I (2)F AND 1(2)G for 15 minutes or longer.
Table S2 Unit 1                                      Unit2 Start-up Aux XFMR IA                        Start-up Aux XFMR 2A Start-up Aux XFMR I B                        Start-up Aux XFMR 2B Basis:
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 emergency busses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency busses (see Table S2 above), whether or not the busses are powered from it.
Fifteen minutes is the threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level uses IC SAL 102
 
SU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.
Operating Mode Applicability:          Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The Emergency Director should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Level in at least one steam generator Steam Generator Auxiliary Feed Water Flow Basis:
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.
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 plant safety.
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. Various instrumentation is also used to determine RCS Level -RVLIS, pressurizer level, digital or recorders. A loss of all Control Room sources for this parameter would also apply.
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 is 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.
103
 
This EAL is focused on a selected subset of plant parameters associated with the key safety functions ofreactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the control room is considered 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
104
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(I)    RCS coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits as indicated by ANY of the following:
Dose Equivalent I-131 greater than 0.5 &#xb5;Ci/gm for greater than 48 hours Dose Equivalent I-131 greater than Technical Specification figure 3 .4.16-1.
IF less than 20% power, THEN use the Dose Equivalent I-131 20% power limit on Technical Specification figure 3.4.16-1 RCS gross specific activity greater than 100/E &#xb5;Ci/gm.
Basis:
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 plant safety.
Escalation of the emergency classification level uses ICs FAI or the Recognition Category R I Cs.
105
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
( 1)    RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)      RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)      Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
This IC addresses RCS leakage that could 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 plant safety.
EAL #1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification is required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level uses ICs of Recognition Category R or F.
106
 
SU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (1 or 2)
(1)      a.      An automatic trip did not shutdown the reactor.
AND
: b.      A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a.      A manual trip did not shutdown the reactor.
AND
: b.      EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic trip is successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and represents a potential degradation of the level of plant safety.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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 plant's decay heat removal systems.
If an initial manual reactor trip 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 trip using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
A manual action at the reactor control consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
This action does not include manually driving in control rods or implementation of boron 107
 
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".
The plant response to the failure of an automatic or manual reactor trip 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 SAS. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SAS or F Al, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance will be applied.
* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and will be evaluated.
* If the signal does not cause a plant transient and the trip 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.
108
 
SU6 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2 or 3)
(I)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will be assessed only when extraordinary means are used 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).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Alabama, Georgia, and Florida; Houston and Henry Counties, Alabama; and Early County, Georgia.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
109
 
SU7 ECL: Notification of Unusual Event Initiating Condition: Failure to isolate containment or loss of containment pressure control.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)
(1)    a.      Failure of containment to isolate when required by an actuation signal.
AND
: b.      ALL required penetrations are not closed within 15 minutes of the actuation signal.
(2)    a.      Containment pressure greater than 27 psig.
AND
: b.      Less than one CTMT fan cooler AND one full train of CTMT spray is operating per design for 15 minutes or longer.
Basis:
This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety.
For EAL #1, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - will be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. At Farley, a single CTMT fan cooler along with one train of CTMT spray is required per design basis. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event will escalate to a Site Area Emergency in accordance with IC FS 1 if there were a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
                                                , 110
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure ATWS ................................................................................... Anticipated Transient Without Scram CC ......................................................................................................................... Cubic Centimeter CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CHM ................................................................................................................................. Chemistry CPM .................................................................................................................... Counts Per Minute CTMT/CNMT ............................................................................................................... Containment CSF ............................................................................................................. Critical Safety Function DC .............................................................................................................................. Direct Current EAL ........................................................................................................... Emergency Action Level ECCS ............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level ENN ............................................................................................. Emergency Notification Network ENS ................................................................................................ Emergency Notification System EOP ............................................................................................... Emergency Operating Procedure EPA ............................................................................................. Environmental Protection Agency ESF .......................................................................................................... Engineered Safety Feature FAA ............................................................................................... Federal Aviation Administration FBI .................................................................................................. Federal Bureau of Investigation FEMA ............................................................................. Federal Emergency Management Agency FNP .................................................................................................................. Farley Nuclear Plant FTS ......................................................................................... Federal Telecommunications System GE ...................................................................................................................... General Emergency HOO .................................................................................. Headquarters Operations Officer (NRC)
Hx ............................................................................................................................. Heat Exchanger IC ........................................................................................................................ Initiating Condition ID ............................................................................................................................. Inside Diameter ISFSI ........................................................................... Independent Spent Fuel Storage Installation Kerr...................................................................................... Effective Neutron Multiplication Factor MCB .................................................................................................................. Main Control Board MCC ................................................................................................................ Motor Control Center
&#xb5;Ci ................................................................................................................................. micro-Curie mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. North American Aerospace Defense Command NOUE .............................................................................................. Notification Of Unusual Event OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM ........................................................................................... Offsite Dose Calculation Manual ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG ....................................................................................................... Protective Action Guideline PBX ........................................................................................................... Private Branch Exchange PWR ........................................................................................................ Pressurized Water Reactor A-1
 
PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCDT ................................................................................................... Reactor Coolant Drain Tank RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RHR ............................................................................................................. Residual Heat Removal RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RVLIS ...................................................................... Reactor Vessel Level Instrumentation System SAE ................................................................................................................. Site Area Emergency SCBA . ................. ... .......... .... ........... ..... .. ...... ..... ........ ... .......... Self-Contained Breathing Apparatus SFP ........................................................................................................................... Spent Fuel Pool SG ........................................................................................................................... Steam Generator SI .............................................................................................................................. Safety Injection SNC ....................................................................................................... Southern Nuclear Company SPDS ............................................................................................ Safety Parameter Display System TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel TV ........................................................................................................................... Threshold Value VDC .................................................................................................................. Volts Direct Current VOiP ................................................................................................... Voice Over Internet Protocol WHT ................................................................................................................. Waste Holdup Tank A-2
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
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.
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.
Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL ): 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 effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NOUE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT CLOSURE: Per FNP-1(2)-STP-18.4, "Containment Integrity Verification and Closure".
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 may require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The area that encompasses all controlled areas within the FNP site boundary but outside the security protected area fence.
B-2
 
PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
REFUELING PATHWAY: This includes the reactor refuel cavity, the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
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.
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.
UNISOLABLE: 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.
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.
B-3
 
Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 Responses to Requests for Additional Information EDWIN I. HATCH NUCLEAR PLANT EAL SCHEME CLEAN COPIES
 
HATCH NUCLEAR PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND ..................................................................................... 1 1.1 OPERATING REACTORS .................................................................................................. 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) ..................................... l 1.3 NRC ORDEREA-12-051 ................................................................................................ 2 1.4 ORGANIZATION AND PRESENTATION OF INFORMATION ............................................... 3 1.5 IC AND EAL MODE APPLICABILITY .............................................................................. 3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 5 2.1 GENERAL CONSIDERATIONS .......................................................................................... 5 2.2 CLASSIFICATION METHODOLOGY ................................................................................. 6 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .......................................... 6 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ................................ 6 2.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................... 7 2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ................... 7 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS ................................................................. 7 2.8 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................................. 7 2.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ................ 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS ........................*. 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 26 5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .............. 47 6  FISSION PRODUCT BARRIER ICS/EALS .................................................................. 50 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......... 64 8  SYSTEM MALFUNCTION ICS/EALS ............................................................................ 88 APPENDIX A - ACRONYMS AND ABBREVIATIONS ............*...........................................A-1 APPENDIX B - DEFINITIONS .....................................*.............................*.......................8-1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title 10, Code of Federal Regulations (CPR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections of this document are:
* 10 CPR&sect;    50.47(a)(l)(i)
* 10 CPR&sect;    50.47(b)(4)
* 10 CPR&sect;    50.54(q)
* 10 CPR&sect;    50.72(a)
* 10 CPR&sect;    50, Appendix E, IV.B, Assessment Actions
* 10 CPR&sect;    50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by various regulatory guidance documents.
Documents of particular relevance to NEI 99-01 include:
NUREG-0654/FEMA-REP-1, Criteriafor Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines 10 CFR &sect; 50. 72 and&sect; 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions may also be directed to the NEI Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CPR 50 emergency plan to fulfill the requirements of 10 CPR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CPR&sect; 50 and the guidance in NUREG 0654/FEMA-REP-l. The initiating conditions germane to a 10 CPR &sect; 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a 10 CPR&sect; 50.47 emergency plan.
The generic ICs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HUl covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. Additionally, appropriate aspects of IC HU 1 and IC HA 1 will also be included to address a HOSTILE ACTION directed against an ISFSI.
 
The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, and ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision-makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool.
* A display in an area accessible following a severe event.
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These 2
 
EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is IO CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the plan's effectiveness. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.
1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode              R          c        E          F        H          s Power Operations          x                    x          x          x          x Startup            x                    x          x          x          x Hot Shutdown          x                    x          x          x          x Cold Shutdown          x          x        x                    x Refueling          x          x        x                    x Defueled            x          x        x                    x 3
 
Hatch Units 1 and 2 Technical Specifications Table 1.1-1 provides the following mode definitions:
Average Reactor Reactor Mode Mode            Title                                                  Coolant Switch Positon Temperature (&deg;F) 1      Power Operation    Run                                        NA 2      Startup            Refuel(a) or Startup/Hot Standby            NA 3      Hot Shutdown(a)    Shutdown                                  > 212 4      Cold ShutdownCbJ    Shutdown                                  ::;212 5      RefuelingCbl        Shutdown or Refuel                          NA (a)  All reactor vessel head closure bolts fully tensioned.
(b)  One or more reactor vessel head closure bolts less than fully tensioned.
In addition to these defined modes, "Defueled" is also applicable to the Hatch EAL scheme, consistent with NEI 99-01. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Hatch EALs with no modifications from NEI 99-01.
When a unit is defueled, the _Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
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2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes, and the informing basis information. In the recognition category F matrices, EALs are referred to as fission product barrier thresholds; the thresholds serve the same function as an EAL.
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 Planningfor Nuclear Power Plants.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The indications will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a: stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 r~lease_start time is unknown, it will be assumed~that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with the operating license is maintained, provided the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 &sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
5
 
Although 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 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 ifthe effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the fission product barrier tables; judgment may be used to determine the status of a fission product barrier.
2.2  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 ifthe EAL has been met or exceeded. An EAL(s) evaluation must be consistent with the related operating mode applicability and notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is declared in accordance with plant procedures.
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.
2.3  CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS In the event of multiple emergencies or conditions, 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, a Site Area Emergency will be declared.
There is no "additive" effect froin multiple EALs meeting the same ECL. For example:
If two Alert EALs are met, an Alert will be declared.
Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary,(RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events.
2.4  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, determines whether 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.
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2.5 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 will 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.
2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification. Termination criteria contained in procedure NMP-EP-110, Emergency Classification and Initial Actions shall be completed for an event to be terminated. At termination, on an event specific basis, the site can either enter normal operating conditions or enter a recovery condition with a recovery organization established for turnover from the ERO.
2.7 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 end before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated EAL must be declared regardless of its continued presence at the time of declaration. For example, an earthquake, or failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
2.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the ICs and/or EALs in this document employ time-based criteria that require 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, some transient conditions may cause an EAL to be met for a brief period of time. The following guidance will be applied to the classification of these conditions.
EAL momentarily met during expected plant response - When 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 applicable EAL is not considered met and the associated emergency declaration is not required. This example presents an illustration:
An ATWS occurs and RCIC fails to automatically start. RPV level rapidly decreases and the plant enters an inadequate RPV Water Level condition (a potential loss of the fuel clad barrier and a loss of the RCS barrier). If an operator 7
 
manually starts RCIC in accordance with an EOP step and clears the inadequate RPV Water Level condition prior to an emergency declaration, then the classification should be based on the ATWS only.
It is important to note that the 15-minute emergency classification assessment period is not a "grace period" to delay a classification in order to perform 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 where an operator is able to take corrective action before the emergency director completes the review and necessary steps to make the emergency declaration. This provision ensures any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.
2.9  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. Personnel could discover 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. It may be the event or condition was not recognized at the
* time, or there was an error in the emergency classification process.
In these cases, no emergency declaration is warranted, but the guidance contained in NUREG-1022 is applicable. Specifically, the event will be reported to the NRC in accordance with 10 CFR &sect; 50.72 within one hour of the discovery of the undeclared event or condition. The licensee will also notify appropriate State and local agencies in accordance with the agreed-on arrangements.
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3    ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL                SITE AREA UNUSUAL EVENT ALERT EMERGENCY              EMERGENCY RGl Release of          RSl    Release of      RAl    Release of      RUl    Release of gaseous radioactivity  gaseous radioactivity  gaseous or liquid      gaseous or liquid resulting in offsite    resulting in offsite    radioactivity resulting radioactivity greater dose greater than 1,000 dose greater than 100  in offsite dose greater than 2 times the mrem TEDE or 5,000      mrem TEDE or 500        than 10 mrem TEDE      ODCM limits for 60 mrem thyroid CDE.      mrem thyroid CDE.      or 50 mrem thyroid      minutes or longer.
Op. Modes: All          Op. Modes: All          CDE.                    Op. Modes: All Op. Modes: All RG2 Spent fuel pool    RS2 Spent fuel pool    RA2      Significant    RU2    UNPLANNED level cannot be        level at Level 3.      lowering of water level loss of water level restored to at least    Op. Modes: All          above, or damage to,    above irradiated fuel.
Level 3 for 60 minutes                          irradiated fuel.        Op. Modes: All or longer.                                      Op. Modes: All Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
1---
RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded -15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL # 1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reactor Building Vent Accident Range Monitor:
1Dl 1-P601 (feeding 1Dl 1-R631, Rx Bldg Vent Wide Range)    2.6 &#xb5;Ci/cc 2Dl 1-P601(feeding2Dl 1-R631, Rx Bldg Vent Wide Range)      2.6 &#xb5;Ci/cc Main Stack Accident Range Monitor:
                      ~- ~~Dl 1:-P007 (~eedi~g 1011-R~~l, Main Stack Wide Range)      8.1_ x 1_0 3 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
Basis:
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 (PA Gs). 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 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 10
 
inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used to determine the monitor reading threshold values in ICs RS I and RA 1. This protocol will maintain intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting.
The TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.
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.
11
 
RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least Level 3 for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
(l)      Spent fuel pool level cannot be restored to at least Level 3 for 60 minutes or longer.
Basis:
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 1.vell after another General Emergency IC \Vas met: however, it is included to provide classification diversity.
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RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reactor Building Vent Accident Range Monitor:
lDl l-P601 (feeding lDl l-R631, Rx Bldg Vent Wide Range)    2.6 x 10- 1 &#xb5;Ci/cc 2Dl 1-P601(feeding2Dl l-R631, Rx Bldg Vent Wide Range)        2.6 x 10- 1 &#xb5;Ci/cc Main Stack Accident Range Monitor:
lDl l-P007 (feeding lDl l-R631, Main Stack Wide Range)      8.1 x 102 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to IO-percent 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 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 13
 
inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in I Cs RG 1 and RA 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at IO-percent 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 uses IC RG 1.
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RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at Level 3.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Lowering of spent fuel pool level to Level 3.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition stems from major failures of plant functions needed for protection of the public and warrant a Site Area Emergency declaration.
lt is recognized that this IC would likely not be met until \vell after another Site Area Emergency JC was met; however, it is included to provide classification diversity.
Escalation of the emergency classification level uses IC RGI or RG2.
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RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Notes:
* The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #I should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(I)      Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Reactor Building Vent Accident Range Monitor:
1Dl l-P601(feeding1Dl 1-R631, Rx Bldg Vent Wide Range)          2.6 x 10-2 &#xb5;Ci/cc 2Dl l-P601 (feeding 2Dl 1-R63 l, Rx Bldg Vent Wide Range)      2.6 x 10-2 &#xb5;Ci/cc Main Stack Accident Range Monitor:
1Dl l-P007 (feeding 1Dl 1-R631, Main Stack Wide Range)          8.1x10 1 &#xb5;Ci/cc (2)      Dose assessment using actual meteorology indicates doses greater than I 0 mrem TEDE or 50 mrem thyroid COE at or beyond the site boundary.
(3)      Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid COE at or beyond (site-specific dose receptor point) for one hour of exposure.
(4)      Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid COE greater than 50 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to I-percent 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 plant safety as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled 16
 
release).
Radiological effluent EALs are 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in ICs RO 1 and RS 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and that the threshold values are based on a site boundary (or beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at I-percent of the EPA PAO of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAO 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 uses IC RS 1.
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RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode App licability: All Emergency Action Levels: (1 or 2 or 3)
(1)    Uncovery of irradiated fuel in the REFUELING PATHWAY.
(2)    Damage to irrad iated fuel resulting in a release of radioactivity from the fuel as indicated by alarms on A NY Table RI radiation monitors:
Table Rl Refuel Floor Area Radiation Monitors Unit 1                                            Unit2 ID21-K601 A - Rx Head Laydown Area                  2D2l-K601 A - Rx Head Laydown Area ID21-K601 B - Refueling Floor Stairway              2D21-K601 M - Spent Fuel/Fuel Pool Areas 1D21-K601 D - Refuel Floor                          2D21-K601 E - Dryer/Separator Pool 1D21-K601 E - Drywell Shield Plug                  2D21-K61 l K - RPV Refuel Floor 228' ID21-K601 M - Spent Fuel Pool and New Fuel          2D21-K61 l L - RPV Refuel Floor 228' Storage area Refuel Floor Ventilation Monitors Unit 1                                            Unit2 1Dll-K609 A -D - Rx Bldg. Potential                2Dl l-K609 A-D - Rx Bldg. Potential Contaminated Area Vent Exhaust Rad Monitor          Contaminated Area Vent Exhaust Rad Monitor 1Dll-K61 l A -D - Refuel Floor Vent Exhaust        2Dl l-K611 A-D - Refuel Floor Vent Exhaust
                                  '*                          2Dl l-K634 A-D - Refuel Floor Rx Well Vent.
Exhaust
                                                          "  2Dl l-K635 A-D - Refuel Floor DW/Sep. Vent.
                                    ' '                      Exhaust (3)    Lowering of sp ent fuel pool level to Level 2.
Basis:
REFUELING PA THW A Y: This includes the reactor cavity, the transfer canal, and the spent fuel pool.
This IC addresses even ts 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 sa:fiety challenges to plant personnel and are precursors to a release of radioactivity to the env ironment. As such, they represent an actual or potential substantial degradation of the leve 1 of plant safety.
This IC applies to irrad iated fuel that is licensed for dry storage up to the point that the loaded 18
 
storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-HUl.
Escalation of the emergency is based on either Recognition Category R or C ICs.
EAL#l This EAL escalates from RU2. 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), 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 will be considered in combination with other available indications of inventory loss.
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.
EAL#2 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 will be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
EAL#3 Spent fuel pool \Vater 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 is 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 uses ICs RSI or RS2 19
 
RAJ ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
Note: 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.
(1)    Dose rate greaterthan 15 mR/hr in ANY of the following areas:
Control Room area radiation monitor ID21-K600 B or C Central Alarm Station (by survey)
(2)    An UNPLANNED event results in radiation levels that prohibit or impede access to any Table Hl plant rooms or areas:
Table Hl Rooms                      A  licable Modes Diesel                      All                                                All Unit 1/2 130'                                      All Reactor building            Unit 1/2 SE Diagonals RHR)                          All Unit 1/2 NE Diagonals (RHR)                        All Basis:
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.
This IC addresses elevated radiation levels in certain plant rooms or 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. It represents an actual or potential substantial degradation of the level of plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL #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.
20
* 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 I 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., nonnal 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.
Escalation of the emergency classification level uses Recognition Category R, C or FI Cs.
21
 
RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
(1)      Reading on ANY effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
Reactor Building Vent Normal Range Monitor:
1Dl 1-K619 A(B) 2Dl 1-K636 A(B)
Main Stack Normal Range Monitor:
1Dl 1-K600 A(B)
Liquid Radwaste Effluent Line Monitor:
1Dl 1-K604 2Dl 1-K604 Service Water System Effluent Line Monitor:
1Dl 1-K605 2Dl 1-K605 (2)      Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
(3)      Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
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.
22
 
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment indicates 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.
EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EAL #3 - 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 uses IC RAl.
23
 
RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels:
(1)    a.      UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Personnel report of low water level SFP low level alarm annunciator - Spent Fuel Storage Pool Level Low 654-022-1/2 AND
: b.      UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
I D2 l-K60 I A - Rx Head Laydown Area ID21-K601 D - Refuel Floor 1D2l-K601 E -Drywell Shield Plug 1D21-K601 M - Spent Fuel Pool and New Fuel Storage area 2D21-K601 A- Rx Head Laydown Area 2D21-K601 M- Spent Fuel/Fuel Pool Areas 2D21-K601 E - Dryer/Separator Pool 2D2 l-K611 K - RPV Refuel Floor 228' 2D21-K61 I L - RPV Refuel Floor 228' Basis:
REFUELING PATHWAY: This includes the reactor cavity, the transfer canal, and the spent fuel pool 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.
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 plant safety.
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 24
 
(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 with Recognition Category C during the Cold Shutdown and Refueling modes.
Escalation of the emergency classification level uses IC RA2.
25
 
4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL                SITE AREA ALERT          UNUSUAL EVENT EMERGENCY              EMERGENCY CGl Loss ofRPV          CSl Loss ofRPV        CAl Loss ofRPV        CUl UNPLANNED inventory affecting      inventory affecting    inventory.            loss of RPV inventory fuel clad integrity with core decay heat        Op. Modes: Cold        for 15 minutes or containment              removal capability. Shutdown, Refueling    longer.
challenged.              Op. Modes: Cold                              Op. Modes: Cold Op. Modes: Cold          Shutdown, Refueling                          Shutdown, Refueling Shutdown, Refueling CA2 Loss of all        CU2 Loss of all but offsite and all onsite one AC power source AC power to            to emergency buses for emergency buses for    15 minutes or longer.
15 minutes or longer. Op. Modes: Cold Op. Modes: Cold        Shutdown, Refueling, Shutdown, Refueling,  Defueled Defueled CA3 Inability to      CU3 UNPLANNED maintain the plant in  increase in RCS cold shutdown.        temperature.
Op. Modes: Cold        Op. Modes: Cold Shutdown, Refueling    Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling CVS Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refueling, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Cold Shutdown, Refueling 26
 
CG1 ECL: General Emergency Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(1)    a.      RPV level less than -155" (TAP) for 30 minutes or longer.
AND
: b.      ANY indication from the Containment Challenge Table C 1.
(2)    a.      RPV level cannot be monitored for 30 minutes or longer.
AND
: b.      Core uncovery is indicated by ANY of the following:
* A reading greater than 9.5 x 103 mR/hr on ANY of the following radiation monitors:
Unit 1                                      Unit2 1D2l-K601      A - Rx Head Laydown Area          2D21-K60 I A - Rx Head Lay down Area 1Dl 1-K601    D - Refuel Floor                  2D21-K601  M- Spent Fuel/Fuel Pool Areas I D2 l-K60 I  E - Drywell Shield Plug          2D21-K601  E - Dryer/Separator Pool I D2l-K601    M - Spent Fuel Pool and New Fuel  2D21-K611  K- RPV Refuel Floor 228' Storage Area                  2D21-K611  L- RPV Refuel Floor 228'
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps          Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps      Turbine Building Floor Drain Sumps Torus                              Rad Waste Tanks Torus Room Sumps                                  ~ *.
AND
: c.      ANY indication from the Containment Challenge Table C 1.
27
 
Containment Challenge Table Cl Containment H2 greater than or equal to 6% AND 02 greater than or equal to 5%
UNPLANNED increase in Primary Containment Pressure Secondary CONTAINMENT INTEGRITY NOT established*
Secondary Containment radiation monitors greater than Max Safe values (SC EOP - Table 6)
* If Secondary CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.
Basis:
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1. Secondary Containment OPERABLE per Technical Specification 3.6.4.1 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.
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.
Water level for top of active fuel is calculated at -15 8.44". Although slightly more conservative, the -155" EOP value for top of active fuel is provided for this EAL to aid in operator recognition of the event.
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 reactor vessel level cannot be restored, fuel damage is probable.
With Secondary CONTAINMENT INTEGRITY not established, there is- a high potential for a direct and unmonitored release of radioactivity to the environment. If Secondary CONTAINMENT INTEGRITY 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 bum (i.e., at the lower deflagration limit). A hydrogen bum 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.
In EAL 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total 28
 
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 actions to terminate leakage, recover inventory control or 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 RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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.
29
 
CS1 ECL: Site Area Emergency Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(1)    a. Secondary CONTAINMENT INTEGRITY not established.
AND
: b. RPV level less than - 41" (6" below the Level 2 actuation setpoint).
(2)    a. Secondary CONTAINMENT INTEGRITY established.
AND
: b. RPV level less than -155" (T AF).
(3)    a. RPV level cannot be monitored for 30 minutes or longer.
AND
: b. Core uncovery is indicated by ANY of the following:
* A reading greater than 9.5 x l 03 mR/hr on ANY of the following radiation monitors:
Unit 1                                          Unit2 1D21-K601  A- Rx Head Laydown Area            2D21-K601  A - Rx Head Laydown Area 1Dl 1-K601  D - Refuel Floor                    2D21-K60 I  M - Spent Fuel/Fuel Pool Areas 1D21-K601  E- Drywell Shield Plug              2D2l-K601  E - Dryer/Separator Pool 1D21-K601  M- Spent Fuel Pool and New Fuel    2D2 l-K61 l K - RPV Refuel Floor 228' Storage Area                    2D2 l -K611 L - RPV Refuel Floor 228'
* UNPLANNED level increase in any of the following of sufficient magnitude to indicate core uncovery:
Drywell Floor Drain Sumps            Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps        Turbine Building Floor Drain Sumps Torus                                Rad Waste Tanks
                                                                                    ', ~
Torus Room Sumps                                            ,,
30
 
Basis:
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3 .6.1.1. Secondary Containment OPERABLE per Technical Specification 3 .6.4.1.
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.
This IC addresses a significant and prolonged loss of RPV 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 warrant a Site Area Emergency declaration.
Water level for top of active fuel is calculated at -158.44". Although slightly more conservative, the -155" EOP value for top of active fuel is provided for this EAL to aid in operator recognition of the event.
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 reactor vessel level cannot be restored, fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying Secondary CONTAINMENT INTEGRITY following a loss of heat removal or RCS inventory control functions. The difference in the specified reactor vessel levels of EALs l.b and 2.b reflects that with Secondary CONTAINMENT INTEGRITY established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, 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 uncover:Y'has acfuallY occurred~ (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for actions to terminate leakage, recover inventory control or 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 RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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 uses IC CG 1 or RG 1.
31
 
CA1 ECL: Alert Initiating Condition: Loss of RPV inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of RPV inventory as indicated by level less than -35" (Level 2 actuation setpoint).
(2)    a.      RPV level cannot be monitored for 15 minutes or longer AND
: b.      UNPLANNED level increase in any of the following due to a loss of RPV inventory:
Drywell Floor Drain Sumps              Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps          Turbine Building Floor Drain Sumps Torus                                  Rad Waste Tanks Torus Room Sumps Basis:
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.
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 EAL #1, a lowering of water level below -35" (Level 2 actuation setpoint) 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.
Although related, EAL # 1 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 residual heat removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, 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 32
 
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 RPV.
The 15-minute duration for the loss oflevel indication was chosen because it is half of the EAL duration specified in IC CS I If the RPV inventory level continues to lower, then escalation to Site Area Emergency uses IC CSI.
33
 
CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC Power (Table SI) to 4160 VAC Emergency Buses l/2E, l/2F, AND l/2G for 15 minutes or longer.
Table Sl Unit 1                                Unit2 Start-up Aux XFMR 1C                  Start-up Aux XFMR 2C Start-up Aux XFMR lD                  Start-up Aux XFMR 2D Diesel Generator 1A                    Diesel Generator 2A Diesel Generator 1B                    Diesel Generator 1B Diesel Generator 1C                    Diesel Generator 2C Basis:
This IC addresses a total loss of AC power (see Table S 1 above) that compromises the performance of all SAFETY SYSTEMS requiring electric power, including those necessary for essential core cooling, containment heat removal/pressure control, spent fuel heat removal, and the ultimate heat sink.
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 emergency 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. When in these modes, this condition represents an actual or potential substantial degradation of the level of plant safety.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses IC CS 1 or RS 1.
34
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(I)      UNPLANNED increase in RCS temperature to greater than 212 &deg;F for greater than the duration specified in Table C2.
Table C2: RCS Heat-up Duration Thresholds Secondary CONTAINMENT RCS Status                                                    Heat-up Duration INTEGRITY Status Not Established                  0 minutes*
Not Intact Established                    20 minutes Intact                        Not applicable                  60 minutes*
* If RHR is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)      UNPLANNED RCS pressure increase greater than 10 psig.
Basis:
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3 .6.1.1. Secondary Containment OPERABLE per Technical Specification 3 .6.4.1.
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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact, and Secondary CONTAINMENT INTEGRITY is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because I) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
35
 
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when Secondary CONTAINMENT INTEGRITY is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, the RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of Secondary CONTAINMENT INTEGRITY 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.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS 1 or RS 1.
36
 
CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds(> 35 mph sustained) or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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 may 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.
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.
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.
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 37
 
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 plant safety.
EAL 1.a identifies hazardous events that could result in damage to plant systems. A seismic event is indicated by entry into IC HU2. Flooding is indicated by a significant increase in water levels (external or internal). High winds are indicated by sustained winds at the site meteorological tower exceeding 35 mph.
The first threshold for EAL 1.b 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 threshold for EAL l.b 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 all 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 uses IC CS 1 or RS 1.
38
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    UNPLANNED loss of reactor coolant results in RPV level less than the lower limit of the controlling level band for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED level increase in any of the following:
Drywell Floor Drain Sumps                Reactor Building Floor Drain Sumps Drywell Equipment Drain Sumps            Turbine Building Floor Drain Sumps Torus                                    Rad Waste Tanks Torus Room Sumps Basis:
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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL # 1 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 ifthe 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.
39
 
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.
EAL #2 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 RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CAI or CA3.
40
 
CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      AC power capability to 4160 VAC Emergency Buses I/2E, 1/2F, AND 1/2G is reduced to a single power source (Table S 1) for 15 minutes or longer.
AND
: b.      Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                          Unit2 Start-up Aux XFMR 1C            Start-up Aux XFMR 2C Start-up Aux XFMR ID            Start-up Aux XFMR 2D Diesel Generator 1A              Diesel Generator 2A Diesel Generator 1B              Diesel Generator 1B Diesel Generator 1C              Diesel Generator 2C Basis:
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.
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 (see Table SI above) 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. When in these modes, this condition is considered to be a potential degradation of the level of plant safety.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Examples of this condition include:
* A loss of all offsite power with a concurrent failure of all but one emergency power source 41
 
(e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes is the 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.
42
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      UNPLANNED increase in RCS temperature to greater than 212 &deg;F.
(2)      Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
Basis:
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.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of plant safety. If the RCS is not intact and secondary CONTAINMENT INTEGRITY is not established during this event, the emergency director will 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.
EAL #1 involves a ioss of decay heat removai capability, or an addition of heat to the RCS in excess of that which can currently be removed, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions, and operators are 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 is a threshold to exclude transient or momentary losses of indication.
 
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
44
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)    Indicated voltage is less than 105/210 VDC on Technical Specification required 125/250 VDC buses 1/2R22-SO 16 AND l/2R22-SO 17 for 15 minutes or longer.
Basis:
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. This condition is considered to be a potential degradation of the level of plant safety.
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 Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train Bis in-service (operable), then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
45
 
CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
Plant telephones (Includes hardwired and wireless)
Plant page Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will 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).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 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 of Georgia, Appling County, Jeff Davis County, Tattnall County and Toombs County.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
46
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 47
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY value listed on Table E 1.
Table El Location of Dose Rate                Total Dose Rate (Neutron+ Gamma mR/hr)
HI-TRAC 125 Side - Mid- height                      450 Top                              110 HI-STAR 100 or HI-STORM 100 Side - 60 inches below 80 mid-height Side - Mid- height                        80 Side - 60 inches above 30 mid- height Center of lid                        10 Middle of top lid                      20 Top (outlet) duct                    40 Bottom (inlet) duct                    140 Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
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 that could cause challenges in removing the cask or fuel from storage.
The existence of "damage" is determined by radiological survey. The radiation reading values listed in the table represent 2 times the site-specific cask-specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical 48
 
specification multiple of "2 times" 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, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under ICs HUI and HAI.
49
 
6 FISSION PRODUCT BARRIER ICS/EALS                          LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS FUEL ClAD              RCS            CONTAINMENT Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.                                                            -  YES**  FGI -Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier FGl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown
                                                                                                      ~------NO--~
SITE AREA EMERGENCY Loss or Potential Loss of any two barriers. LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS FUELClAD                RCS            CONTAINMENT FSl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown ALERT                                                                                        FS1 - Loss or Potential Loss of ANY Two Barriers Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
FAl POTENTIAL            POTENTIAL LOSS Op. Modes: Power Operation, Hot Standby,                  LOSS LOSS LOSS FUEL ClAD Startup, Hot Shutdown                                                      RCS 50
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY                            FSl SITE AREA EMERGENCY                                  FAlALERT Loss of any two barriers and Loss or            Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either the Potential Loss of the third barrier.                                                          Fuel Clad or RCS barrier.
Fuel Clad Barrier                                  RCS Barrier                                  Containment Barrier LOSS                POTENTIAL                LOSS              POTENTIAL LOSS                    LOSS                POTENTIAL LOSS                                                                                                LOSS
: 1. RCS Activity                              1. Primary Containment Pressure                  1. Primary Containment Conditions A. Activity of 300    Not Applicable        A. Primary              Not Applicable            A. UNPLANNED rapid          A. Primary
    &#xb5;Ci/gm DEl131                              containment                                        drop in primary              containment pressure greater                                  containment pressure          pressure than 1.85 psig due                                following primary            greater than 56 to RCS leakage.                                    containment pressure          psig rise                        OR OR                      B. Greater than or B. Primary containment            equal to 6% H2 pressure response not        AND5%02 consistent with LOCA          exists inside conditions.                  primary containment OR
: c. HCTL exceeded.
: 2. RPV Water Level                          2. RPV Water Level                                2. RPV Water Level A. SAG entry is        A. RPV water level    A. RPV water level      Not Applicable            Not Applicable              A. SAG entry is required.              cannot be            cannot be restored                                                              required.
restored and        and maintained maintained          above -155 inches above -155          or cannot be inches or cannot    determined.
be determined.
51
                                                                                                                                                --- _ _ _J
 
Fuel Clad Barrier                        RCS Barrier                                Containm~nt  Barrier LOSS            POTENTIAL          LOSS            POTENTIAL LOSS                    LOSS                POTENTIAL LOSS                                                                                      LOSS
: 3. Not Applicable                    3. RCS Leak Rate                              3. Primary Containment Isolation Failure Not Applicable        Not Applicable A. UNISOLABLE        A. UNISOLABLE          A. UNISOLABLE direct        Not Applicable break in Main          primary system          downstream pathway Steamline, HPCI,      leakage that results    to the environment Feedwater,            in exceeding            exists after primary R WCU, or RCIC        EITHER of the          containment isolation OR                    following:              signal B. Emergency RPV        1. Max Normal            OR Depressurization.        Operating        B. Intentional primary Temperature          containment venting OR                  per EOPs
: 2. Max Normal            OR Operating Area    C. UNISOLABLE Radiation Level. primary system leakage that results in exceeding EITHER of the following:
: 1. Max Safe Operating Temperature.
OR
: 2. Max Safe Operating Area Radiation Level.
: 4. Primary Containment Radiation      4. Primary Containment Radiation              4. Primary Containment Radiation A. DWRRM              Not Applicable A. D WRRM greater    Not Applicable            Not Applicable              A. D WRRM greater greater than                        than 40 R/hr.                                                              than 26,000 R/hr.
1,400 R/h.
52
 
Fuel Clad Barrier                                RCS Barrier                                Containment Barrier LOSS            POTENTIAL                LOSS            POTENTIAL LOSS                  LOSS              POTENTIAL LOSS                                                                                          LOSS
: 5. Other Indications                      5. Other Indications                          5. Other Indications A. Not Applicable. Not Applicable.      A. Drywell Fission    Not Applicable.          Not Applicable.          Not Applicable.
Product Monitor reading 5.0 x 105 cpm.
: 6. Emergency Director Judgment            6. Emergency Director Judgment                6. Emergency Director Judgment A. ANY condition A. ANY condition A. ANY condition in            A. ANY condition in the  A. ANY condition in the  A. ANY condition in the opinion of    in the opinion      the opinion of the    opinion of the          opinion of the            in the opinion of the emergency        of the              emergency            emergency director      emergency director        the emergency director that        emergency            director that        that indicates          that indicates loss of    director that indicates loss of    director that        indicates loss of    potential loss of the    the Containment          indicates the fuel clad        indicates            the RCS Barrier.      RCS Barrier.            Barrier.                  potential loss of barrier.              potential loss of                                                                            the Containment the fuel clad                                                                                Barrier.
barrier.
53
 
l I
I Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The fuel clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets.
: 1.      RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;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 percent 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.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no potential loss threshold associated with RCS Activity.
: 2.      RPV Water Level Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.
Potential Loss 2.A 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 2.A. 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.
This threshold is considered 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 54
 
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 A TWS/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 SA5 or SS5 will dictate the need for emergency classification.
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.
: 3. Not Applicable (included for numbering consistency between barrier tables)
: 4. Primary Containment Radiation Loss 4.A 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 &#xb5;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 percent 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.
: 5. Other Indications Not Applicable (included for numbering consistency between barrier tables)
: 6. Emergency Director Judgment 55
 
Loss 6.A This threshold addresses any other factors that used by the emergency director in determining whether the fuel clad barrier is lost.
Potential Loss 6.A 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 will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
56
 
RCS BARRIER THRESHOLDS:
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.
: 1.      Primary Containment Pressure Loss I.A The greater than 1.85 psig primary containment pressure 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.
: 2.      RPV Water Level Loss 2.A 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 2.A. 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 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 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 beyond the limit. The threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
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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 SAS or SSS will dictate the need for emergency classification.
There is no RCS potential loss threshold associated with RPV water level.
: 3. RCS Leak Rate Loss Threshold 3.A 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.
Loss Threshold 3.B 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). 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.
Potential Loss Threshold 3.A 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 that 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.
: 4. Primary Containment Radiation Loss 4.A S8
 
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.
: 5. Other Indications Loss 5A A Drywell Fission Products Monitor reading 5.0 x 10 5 cpm indicates a breach of the RCS as an effluent. The monitor value calculated in Calculation SMNH-13-021, Rev 1, was 1.008 x 106 cpm; however, the top of the scale for the monitor is 1 x 106 cpm. Therefore, the EAL threshold value has been established at one half decade below top of scale to aid the operator in distinguishing between a loss of RCS event and an instrument failure resulting in the monitor reading high off scale. No radiation monitors capable of indicating a potential loss of the RCS barrier were identified.
There is no Potential Loss Threshold associated with Other Indications.
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
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CONTAINMENT BARRIER THRESHOLDS:
The primary containment barrier includes the drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Containment barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      Primary Containment Conditions Loss I .A and I .B 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. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.
These thresholds rely 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.
Potential Loss I .A 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 represents a potential loss of the containment barrier.
Potential Loss l .B 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.
Potential Loss I .C The heat capacity temperature limit (HCTL) is the highest suppression pool temperature from which emergency RPV depressurization will not raise:
Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
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The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is used 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 l.oss of containment.
: 2. RPV Water Level There is no loss threshold associated with RPV water level.
Potential Loss 2.A The potential loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The potential loss requirement for entry into the Severe Accident Guidelines indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs specify the conditions when the EPGs are exited and SAGs are entered.
Entry into SA Gs is a logical escalation in response to the inability to assure adequate core cooling.
PRA studies indicate that the condition of this potential loss threshold is a core melt sequence that, 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.
: 3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation (automatic or manual) that allows an UNISOLABLE direct release to the environment. A release path is 'direct' if it allows for the migration of radioactive material from the containment to the environment in a generally uninterrupted manner (e.g., little or no holdup time); therefore, within the context of a containment barrier loss or potential loss threshold, a release path through the wetwell is a direct release path.
Loss 3.A 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 R ICs.
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Loss 3.B 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 will also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas 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 pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.
Loss 3.C The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: I) equipment necessary for the safe shutdown of the plant will fail, nor 2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs use these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
The temperatures and radiation levels will 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 that connect directly to the RPV ensuring 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 3.A this threshold would result in a Site Area Emergency.
There is no potential loss threshold associated with Primary Containment Isolation Failure.
: 4. Primary Containment Radiation There is no loss threshold associated with primary containment radiation.
Potential Loss 4.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20-percent 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-percent 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 containment that would then escalate the emergency classification level to a General Emergency.
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: 5. Other Indications Not Applicable (included for numbering consistency between barrier tables)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the emergency director in determining whether the containment barrier is lost.
Potential Loss 6.A This threshold addresses any other factors that may be used by the emergency director in determining whether the containment barrier is potentially lost. The emergency director will also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
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7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                EMERGENCY HGl HOSTILE              HSl HOSTILE              HAl HOSTILE            HUl Confirmed ACTION resulting in      ACTION within the        ACTION within the      SECURITY loss of physical        PROTECTED AREA.          OWNER                  CONDITION or control of the facility. Op. Modes: All          CONTROLLED              threat.
Op. Modes: All                                    AREA or airborne        Op. Modes: All attack threat within 30 minutes.
Op. Modes: All HU2 Seismic event greater than OBE levels.
Op. Modes: All HU3 Hazardous event.
Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HA5 Gaseous release impeding
                                                - access to equipment-.
necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All HS6 Inability to        HA6 Control Room control a key safety    evacuation resulting in function from outside    transfer of plant the Control Room.        control to alternate Op. Modes: All          locations.
Op. Modes: All 64
                                                                                                  ----__:ii
 
GENERAL              SITE AREA ALERT          UNUSUAL EVENT EMERGENCY              EMERGENCY HG7 Other              HS7 Other                HA7 Other              HU7 Other conditions exist which conditions exist which    conditions exist which conditions exist in the judgment of the in the judgment of the    in the judgment of the which in the emergency director    emergency director        emergency director    judgment of the warrant declaration of warrant declaration of    warrant declaration of emergency director a General Emergency. a Site Area              an Alert.              warrant declaration of Op. Modes: All        Emergency.                Op. Modes: All        a (NO)UE.
Op. Modes: All                                  Op. Modes: All 65
 
HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA (PA) as reported by the on shift Security Captain or designee.
AND
: b.      EITHER of the following has occurred:
: 1.      ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* RPV water level
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMINENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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) or 2) loss of spent fuel pool integrity so that sufficient water level cannot be maintained.
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Timely and accurate communications between Security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program].
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 Security Plan.
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HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a General Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a General Emergency.
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__ _J
 
HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.
Operating Mode Applicability: All Emergency Action Levels:
(1)      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA (PA) as reported by the on shift Security Captain or designee.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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.
Timely and accurate communications between Security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program].
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 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 a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA 1.
It also 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, or physical disputes between employees.
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
69
 
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 Security Plan.
Escalation of the emergency classification level uses IC HGl.
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HS6 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      a.      An event has resulted in plant control being transferred from the control room to remote shutdown panels.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* RPV water level
* RCS heat removal Basis:
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 relativejy
    . - -- .sh.ort 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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location.
Escalation of the emergency classification level uses IC FG 1 or CG 1.
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HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Site Area Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, 1) toward site personnel or equipment that could lead to the likely failure of or, 2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a Site Area Emergency.
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HA1 .
ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2)
(1)      A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA (OCA) as reported by the on shift Security Captain or designee.
(2)      A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control of HNP Security This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification bf an aircraft* attack threat.* This event *will req*uire rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA (PA), or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between Security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program].
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, 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, or physical disputes between employees.
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Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71or10 CFR &sect; 50.72.
EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA.
EAL #2 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HS 1.
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HAS ECL: Alert Initiating Conditfon: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(1)    a.      Release of a toxic, corrosive, asphyxiant or flammable gas into any Table HI plant rooms or areas:
Table Hl Buildin2                          Rooms                Annlicable Modes Diesel generator building    All                                        All Unit 112 130'                              All Reactor building            Unit 112 SE Diagonals (RHR)                All Unit 112 NE Diagonals (RHR)                All AND
: b.      Entry into the room or area is prohibited or impeded.
Basis:
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 plant safety.
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 director's 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 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.
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* 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 4.
* 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.
An asphyxiant is a gas capable ofreducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19-percent, 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 uses Recognition Category R, C or FICs.
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HA6 ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(1)    An event has resulted in plant control being transferred from the control room to remote shutdown panels.
Basis:
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 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 uses IC HS6.
77
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for an Alert.
78
 
HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
(1)      A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the on shift Security Captain or designee.
(2)      Notification of a credible security threat directed at HNP.
(3)      A validated notification from the NRC providing information of an aircraft threat.
Basis:
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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 that do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71or10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGl.
Timely and accurate communications between Security shift supervision 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 OROs.
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 [and Independent Spent Fuel Storage Installation Security Program].
EAL #1 references the Security Shift Captain or designee 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 &sect; 2.39 information.
79
 
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with station procedures.
EAL #3 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HA 1.
80
 
HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels Operating Mode Applicability: All Emergency Action Levels:
(1)    Seismic event greater than Operating Basis Earthquake (OBE) as indicated by ANY of the following
* Unit One "Seismic Peak Shock Recorder High G Level" (657-066) alarm
* Unit Two "Seismic Instrumentation Triggered" (657-048) alarm
* A 12.7 Hz amber light illuminated in the N/S OR E/W column on panel 1Hl 1-P701
* A 12.7 Hz red light illuminated in the N/S OR E/W column on panel 1Hl 1-P701 Basis:
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 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 plant safety.
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., typical lateral accelerations are 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 or check internet news sources); 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 uses IC CA6 or SA9.
81
 
HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
(1)    A tornado strike within the PROTECTED AREA (PA).
(2)    Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
(3)    Movement of personnel within the PROTECTED AREA (PA) is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
(4)    A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site in personal vehicles.
(5)      Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours.
Basis:
PROTECTED AREA (PA): The area which encompasses all controlled areas within the security protected area fence.
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.
This IC addresses hazardous events that are considered to represent a potential degradation of the level of plant safety.
EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 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.
EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA (PA).
82
 
EAL #4 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, or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses the phenomena of the hurricane based on the severe weather mitigation procedure.
Escalation of the emergency classification level is based on ICs in Recognition Categories A, F, sore.
83
 
HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Note: The emergency director will declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(I)    a. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
* 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
: b.      The FIRE is located within ANY Table H2 rooms or areas (2)    a. Receipt of a single fire alarm (i.e., no other indications of a FIRE)
AND
: b.      The FIRE is located within ANY Table H2 rooms or areas AND
: c.      The existence of a FIRE is not verified within 30-minutes of alarm receipt.
(3)    A FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.
(4)    A FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.
TableH2 Building                            Rooms CB 147' Cable Spreading Room Control Building U 1/2 CB 112' Station Battery Rooms A,B Diesel generator building  All Primary containment        All Unit 1/2 130' Unit 1/2 SE Diagonals (RHR)
Unit 1/2 NE Diagonals (RHR)
Reactor building Unit 1 SW Diagonals (RCIC)
Unit 2 NW Diagonals (RCIC)
Unit 1/2 HPCI Rooms Intake structure            All 84
 
Basis:
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 (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of plant safety.
EAL#l 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 or automatic activation of a suppression system.
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.
EAL#2 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 EAL# 1 is immediately applicable, and the emergency must be declared ifthe 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.
EAL#3 In addition to a FIRE addressed by EAL #I or EAL #2, a FIRE within the plant PROTECTED AREA (PA) not extinguished within 60 minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA (PA).
85
 
EAL#4 If a FIRE within the plant PROTECTED AREA (PA) or ISFSI PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency, 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 not necessary ifthe agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
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."
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.
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 EAL #2, 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 uses IC CA6 or SA9.
86
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Notification of Unusual Event (NOUE).
Operating Mode Applicability: All Emergency Action Levels:
(1)
* Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
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 NOUE.
87
 
8    SYSTEM MALFUNCTION ICS/EALS GENERAL                  SITE AREA ALERT              UNUSUAL EVENT EMERGENCY                  EMERGENCY SGl Prolonged loss of      SSl Loss of all offsite  SAl Loss of all but one SUl Loss of all offsite all offsite and all onsite and all onsite AC power    AC power source to        AC power capability to AC power to emergency      to emergency buses for    emergency buses for 15    emergency buses for 15 buses.                    15 minutes or longer.      minutes or longer.        minutes or longer.
Op. Modes: Power          Op. Modes: Power          Op. Modes: Power          Op. Modes: Power Operation, Startup, Hot    Operation, Startup, Hot    Operation, Startup, Hot  Operation, Startup, Hot Standby, Hot Shutdown      Standby, Hot Shutdown      Standby, Hot Shutdown    Standby, Hot Shutdown SA2 UNPLANNED              SU2 UNPLANNED loss of Control Room      loss of Control Room indications for 15        indications for 15 minutes or longer with a  minutes or longer.
significant transient in  Op. Modes: Power progress.                  Operation, Startup, Hot Op. Modes: Power          Standby, Hot Shutdown Operation, Startup, Hot, Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS Inability to          SAS Automatic or          SUS Automatic or shutdown the reactor      manual scram fails to    manual scram fails to causing a challenge to    shutdown the reactor, and shutdown the reactor.
RPV water level or RCS    subsequent manual        Op. Modes: Power heat removal.              actions taken at the      Operation Op. Modes: Power          reactor control consoles Operation                  are not successful in shutting down the reactor.
Op. Modes: Power Operation 88
 
GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY SU6    Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS    Loss of all AC    SSS    Loss of all vital and vital DC power        DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 89
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
(I)    a.      Loss of ALL offsite and ALL onsite AC power to 4160 VAC Emergency Buses 1/2E, 1/2F, AND 1/2G.
AND
: b.      EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level.
Basis:
This IC addresses a prolonged loss of all power sources to AC emergency 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. 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 AC emergency 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 emergency 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.
90
 
SGS ECL: General Emergency Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to 4160 V AC Emergency Buses 1/2E, 1/2F, AND 1/2G for 15 minutes or longer.
AND
: b.      Indicated voltage is less than 105/210 VDC on ALL 125/250 VDC Bus 1/2R22-SO 16 AND 1/2R22-SO 17 for 15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. 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 loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
91
 
SS1 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC power to 4160 V AC Emergency Buses l/2E, l/2F, AND 1/20 for 15 minutes or longer.
Basis:
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 is a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RO I, FO I or SO I.
92
 
555 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.
Operating Mode Applicability: Power Operation Emergency Action Levels:
(1)      a. An automatic or manual scram did not shutdown the reactor.
AND
: b.      All manual actions to shutdown the reactor have been unsuccessful.
AND
: c.      EITHER of the following conditions exist:
* Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level
* Exceeding the Heat Capacity Temperature Limit (HCTL) Curve (EOP Graph 2)
Basis:
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 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 shutdown 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level uses IC RG 1 or FG 1.
93
 
SSS ECL: Site Area Emergency Initiating Condition: Loss of all vital DC power for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105/210 VDC on ALL 125/250 VDC Bus 1/2R22-S016 AND 1/2R22-SO 17 for 15 minutes or longer.
Basis:
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.
Fifteen minutes is a threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RGI, FGl or SGS.
94
 
SA1 ECL: Alert Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      AC power capability to 4160 VAC Emergency Buses 1/2E, 1/2F, AND 1/20 is reduced to a single power source (Table SI) for 15 minutes or longer.
AND
: b.      Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                              Unit2 Start-up Aux XFMR IC                Start-up Aux XFMR 2C Start-up Aux XFMR lD                Start-up Aux XFMR 2D Diesel Generator 1A                Diesel Generator 2A Diesel Generator 1B                Diesel Generator I B Diesel Generator 1C                Diesel Generator 2C Basis:
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.
This IC describes a significant degradation of offsite and onsite AC power sources (see Table S 1) 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 SU 1.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
95
 
Fifteen minutes is a threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level uses IC SS 1.
96
 
SA2 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature AND
: b.      ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor scram
* ECCS actuation
* Thermal power oscillations greater than 10%
Basis:
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.
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 product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of plant safety.
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.
97
 
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, ifthe 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 is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses ICs FS 1 or IC RS 1.
98
 
SAS ECL: Alert Initiating Condition: Automatic or manual scram fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels:
Note: A manual 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.
(1)      a. An automatic or manual scram did not shutdown the reactor.
AND
: b.      Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
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 plant safety. 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 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. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers). 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 the failure to shutdown the reactor is prolonged enough to cause a challenge to the RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS5 or FS 1, an Alert declaration is appropriate for this event.
99
 
It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F !Cs; however, this IC and EAL are included to ensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
100
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)    a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds (> 35 mph sustained) or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 101
 
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 plant safety.
EAL 1.a identifies hazardous events that could result in damage to plant systems. A seismic event is indicated by entry into IC HU2. Flooding is indicated by a significant increase in water levels (external or internal). High winds are indicated by sustained winds at the site meteorological tower exceeding 35 mph.
The first threshold for EAL 1.b 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 threshold for EAL l .b 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 uses IC FSl or RSl.
102
 
SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite AC power capability (Table S2) to 4160 VAC Emergency Buses l/2E, 1/2F, AND 1/20 for 15 minutes or longer.
Table S2 Unit 1                                Unit2 Start-up Aux XFMR IC                  Start-up Aux XFMR 2C Start-up Aux XFMR ID                  Start-up Aux XFMR 2D Basis:
This IC addresses a prolonged loss of offsite power (see Table S2 above). The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes is the threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level uses IC SAL 103
 
SU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.
Operating Mode Applicability:          Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer.
Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Basis:
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.
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 plant safety.
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 is reported if it significantly impairs the capability to perform emergency assessments, particularly those 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 ofreactivity 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 104
 
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, ifthe value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS, or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is a threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
105
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)
Note: Use the Unit 1 or Unit 2 Pretreatment (Flow vs. mR/hr) Graphs to determine if the Pretreatment Radiation Monitor exceeds the TV of 240,000 &#xb5;Ci/sec.
(1)    Pretreatment Radiation Monitor 1(2)Dl 1K601 1(2)Dl 1K602 reading greater than 240,000 &#xb5;Ci/sec for greater than 60 minutes.
(2)    Sample analysis indicates that the reactor coolant specific activity is EITHER:
* Greater than 0.2 &#xb5;Ci/gm and less than or equal to 2.0 &#xb5;Ci/gm dose equivalent Irn for greater than 48 hours
* Greater than 2.0 &#xb5;Ci/gm dose equivalent Irn.
Basis:
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 plant safety.
Escalation of the emergency classification level uses ICs FA 1 or the Recognition Category R I Cs.
106
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)    RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)    Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
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 plant safety.
EAL #1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. 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 uses ICs of Recognition Category R or F.
107
 
SU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual scram fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (1 or 2)
Note: A manual 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.
(1)      a. An automatic scram did not shutdown the reactor.
AND
: b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a. A manual scram did not shutdown the reactor.
AND
: b. EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic scram is successful in shutting down the reactor.
Basis:
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 plant safety.
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 plant's 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 heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
108
 
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, and other concurrent plant conditions. 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 SA5. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA5 or FA 1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing),
the following classification guidance will 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 will 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.
109
 
SU6 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones (includes hardwired and wireless)
Plant Page Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telephone System (FTS) Lines Commercial phones Basis:
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).
EAL #I addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 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 of Georgia, Appling County, Jeff Davis County, Tattnall County and Toombs County.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
110
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure A TWS ................................................................................... Anticipated Transient Without Scram BLDG ................................................................................................................................... Building BWR ............................................................................................................. Boiling Water Reactor CB .......................................................................................................................... Control Building CC ......................................................................................................................... Cubic Centimeter CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CPM .................................................................................................................... Counts Per Minute CPS ..................................................................................................................... Counts Per Second DC .............................................................................................................................. Direct Current DEI ............................................................................................................... Dose Equivalent Iodine DW ....................................................................................................................................... Drywell DWRRM .................................................................................... Drywell Wide Range Rad Monitor EAL ........................................................................................................... Emergency Action Level ECCS ............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level ENN ............................................................................................. Emergency Notification Network ENS ................................................................................................ Emergency Notification System EPA ............................................................................................. Environmental Protection Agency EPG ............................................................................................... Emergency Procedure Guideline FAA ............................................................................................... Federal Aviation Administration FBI .................................................................................................. Federal Bureau of Investigation FEMA ............................................................................. Federal Emergency Management Agency FSAR ................................................................................................... Final Safety Analysis Report FTS ......................................................................................... Federal Telecommunications System GE ...................................................................................................................... General Emergency GM ........................................................................................................................................... Gram HCTL .......................................................................................... Heat Capacity Temperature Limit HNP ................................................................................................................... Hatch Nuclear Plant HOO .................................................................................. Headquarters Operations Officer (NRC)
HPCI .............................................................................................. High Pressure Coolant Injection IC ........................................................................................................................ Initiating Condition ISFSI ........................................................................... Independent Spent Fuel Storage Installation LOCA ........................................................................................................ Loss of Coolant Accident MSL ....................................................................................................................... Main Steam Line
&#xb5;Ci ................................................................................................................................. micro-Curie mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man NE ...................................................................................................................................... Northeast NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. North American Aerospace Defense Command (NO)UE .......................................................................................... (Notification Of) Unusual Event OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM ........................................................................................... Offsite Dose Calculation Manual A-I
 
ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG ....................................................................................................... Protective Action Guideline PBX ........................................................................................................... Private Branch Exchange PCIS .................................................................................... Primary Containment Isolation System PRA/PSA .................................... Probabilistic Risk Assessment I Probabilistic Safety Assessment PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCIC ............................................................................................... Reactor Core Isolation Cooling RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RHR ............................................................................................................. Residual Heat Removal RPS ......................................................................................................... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RWCU .......................................................................................................... Reactor Water Cleanup Rx .......................................................................................................................................... Reactor SAG ........................................................................................................ Severe Accident Guideline SAR .............................................................................................................. Safety Analysis Report SC ................................................................................................................ Secondary Containment SCBA ..................................................................................... Self-Contained Breathing Apparatus SE ....................................................................................................................................... Southeast SEP ..................................................................................................................................... Separator SFP ........................................................................................................................... Spent Fuel Pool SNC ....................................................................................................... Southern Nuclear Company SPDS ............................................................................................ Safety Parameter Display System SW ..................................................................................................................................... Southwest TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel TV ........................................................................................................................... Threshold Value VAC ......................................................................................................... Volts Alternating Current VDC .................................................................................................................. Volts Direct Current VOiP ................................................................................................... Voice Over Internet Protocol A-2
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
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.
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.
Notification of Unusual Event (NODE): 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.
The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL): 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 effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NODE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3 .6.1.1. Secondary Containment OPERABLE per Technical Specification 3 .6.4.1 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 may 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.
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control ofHNP Security.
PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
B-2
 
REFUELING PATHWAY: This includes the reactor cavity, the transfer canal, and the spent fuel pool.
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.
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.
UNISOLABLE: 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.
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.
B-3
 
Southern Nuclear Operating Company License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 Responses to Requests for Additional Information VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 EAL SCHEME CLEAN COPIES
 
VOGTLE ELECTRIC GENERATING PLANT EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS
 
TABLE OF CONTENTS 1  REGULATORY BACKGROUND ************************************************************************************* 1 1.1 OPERATINGREACTORS **..**...*....**...**.*...*...**.*.**.***.****.*******..**.....*******..............*.*.***..*..** 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSl) **..***.*****...**...**..****...***.. l 1.3 NRC ORDER EA-12-051 *...*..****..****...*....*....**.***...***.........****..**.*****...**********......**...*******2 1.4 0RGANIZATION AND PRESENTATION OF INFORMATION *...*********...***.*********.*.*....***..**.** 3 1.5 IC AND EAL MODE APPLICABILITY .*..**....*...**.*.*******.**.*.....***..*.*....**....*..*********...**...**.. 3 2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS ........................................ 5 2.1 GENERAL CONSIDERATIONS *.****...***...**..**....***..**...**....*....***...***...********...**...*...****..*..*** 5 2.2 CLASSIFICATION METHODOLOGY *****...**...**...***.***..****.****.*****...***...***..**...**...**....*...***** 6 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .......................................... 6 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION *..*...**.*..****.***.******.... 6 2.5 CLASSIFICATION OF IMMINENT CONDITIONS ............................................................... 7 2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING .*.*.**....*...***. 7 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS ................................................................. 7 2.8 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................................. 7 2.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION ................ 8 3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS .......................... 9 4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS .................... 26 5  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS ************** 45 6  FISSION PRODUCT BARRIER ICS/EALS ****************************************************************** 48 7  HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......... 64 8  SYSTEM MALFUNCTION ICS/EALS ***************************************************************              11 . . . . . . . . . . . . 87 APPENDIX A - ACRONYMS AND ABBREVIATIONS ........................................................A-1 APPENDIX B - DEFINITIONS ******************************************************************************************* 8-1
 
EMERGENCY ACTION LEVELS 1  REGULATORY BACKGROUND 1.1  OPERATING REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities.
Several of these regulations govern various aspects of an emergency classification scheme. The relevant sections for this document are:
* 10 CFR &sect; 50.47(a)(l)(i)
* 10 CFR &sect; 50.47(b)(4)
* 10 CFR &sect; 50.54(q)
* 10 CFR &sect; 50.72(a)
* 10 CFR &sect; 50, Appendix E, IV.B, Assessment Actions
* 10 CFR &sect; 50, Appendix E, IV.C, Activation of Emergency Organization These regulations are supplemented by regulatory guidance documents. Documents of particular relevance to NEI 99-01 include:
NUREG-0654/FEMA-REP-l, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support ofNuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]
NUREG-1022, Event Reporting Guidelines JO CFR &sect; 50. 72 and&sect; 50. 73 Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors This list is not all-inclusive. It is strongly recommended that scheme developers consult with licensing and regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions also may be directed to the NEI Emergency Preparedness staff.
1.2  INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)
Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR &sect; 50 and the guidance in NUREG 0654/FEMA-REP-l. The initiating conditions germane to a 10 CFR &sect; 72.32 emergency plan (as described in NUREG-1567) are contained within the classification scheme for a 10 CFR &sect; 50.47 emergency plan.
The generic ICs and EALs for an ISFSI are presented in Section 5, ISFSI ICs/EALs. IC E-HUl covers credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that process and/or repackage spent fuel. Additionally, appropriate aspects oflC HUI and IC HAl will also be included to address a HOSTILE ACTION directed against an ISFSI.
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The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.
NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.
1.3 NRC    ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2011, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety functions of core cooling and containment integrity, ultimately leading to core damage in three reactors. Although the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to prevent fuel damage from the loss of cooling.
Following a review of the Fukushima Daiichi accident, the NRC concluded that measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). One such measure was that each spent fuel pool be provided with reliable level instrumentation to significantly enhance the ability of key decision-makers to effectively allocate resources following a beyond design basis event. To this end, the NRC issued Order EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.
NRC Order EA-12-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." All licensees must therefore provide:
* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool.
* A display in an area accessible following a severe event.
* Independent electrical power to each instrument channel and an alternate remote power connection capability.
NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modifj; Licenses with Regard to Reliable Spent Fuel Pool Instrumentation", provides guidance for complying with NRC Order EA-12-051.
NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These 2
 
EALs are included within existing IC RA2, and new !Cs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.
It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use.
The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). Licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the plan's effectiveness. Based on this determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.
1.4 ORGANIZATION AND PRESENTATION OF INFORMATION The scheme's information is organized by Recognition Category in the following order.
R - Abnormal Radiation Levels I Radiological Effluent C - Cold Shutdown I Refueling System Malfunction E - Independent Spent Fuel Storage Installation (ISFSI)
F - Fission Product Barrier H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunction 1.5 IC AND EAL MODE APPLICABILITY The following table shows Recognition Categories applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.
MODE APPLICABILITY MATRIX Category Mode                R        c      E        F        H        s Power Operations          x                x        x        x        x Startup              x                x        x        x        x Hot Standby            x                x        x        x        x Hot Shutdown            x                x        x        x        x Cold Shutdown            x        x      x                  x Refueling            x        x      x                  x Defueled              x        x      x                  x 3
 
Vogtle Units 1and2 Technical Specifications Table 1.1-1 provides the following operating mode definitions:
Reactivity      % Rated Condition      Thermal      Average RCS Mode            Title Power(a)    Temperature (&deg;F)
(Kerr) 1    Power Operation        2: 0.99          >5              NA 2    Startup                2: 0.99          :::; 5            NA 3    Hot Standby            <0.99            NA              2: 350 4    Hot Shutdown(b)        < 0.99          NA        350 > Tavg > 200 5    Cold Shutdown(b)      <0.99            NA              :s 200 6    Refueling(c)            NA            NA                NA (a)    Excluding decay heat.
(b)    All reactor vessel head closure bolts fully tensioned.
(c)    One or more reactor vessel head closure bolts less than fully tensioned.
In addition to these identified modes, "Defueled" is also applicable to the Vogtle EAL scheme, consistent with NEI 99-01 guidance. Defueled is a 'No Mode' condition where all of the fuel has been removed from the reactor vessel (i.e., full core offload during refueling or extended outages).
These modes are used throughout the Vogtle EALs with no modifications from NEI 99-01.
When a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.
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2  GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 2.1  GENERAL CONSIDERATIONS For any 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), the associated operating mode applicability, notes and the informing basis information. In the recognition category F matrices, EALs are referred to as fission product barrier thresholds; the thresholds serve the same function as an EAL.
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.
All emergency classification assessments will be based on valid indications, reports or conditions. A valid indication, report, or condition, has been verified using appropriate means, leaving no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.
Indications will be validated in a manner that supports timely emergency declaration.
For ICs and EALs that have a stipulated time duration, the emergency director will not wait until the applicable time has elapsed, but will 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 will be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.
A planned work activity resulting in an expected event or condition that 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 such cases, the controls associated with the planning, preparation and execution of the work will ensure compliance with the operating license is maintained, provided the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 &sect; CFR 50.72.
Some EALs are assessed based on the results of analyses necessary to ascertain whether a specific EAL threshold has been exceeded. The EAL and/or the associated basis discussion will identify the necessary analysis. 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.
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Although the EALs have been developed to address a full spectrum of possible events and conditions that may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-0 I scheme provides the emergency director with the ability to classify events and conditions based on judgment using EALs 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 into the fission product barrier tables; judgment may be used to determine the status of a fission product barrier.
2.2 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. An EAL(s) evaluation must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the IC is met and the associated ECL is declared in accordance with plant procedures.
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.
2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS In the event of multiple emergencies or conditions, the user will identify all EALs met or exceeded. The highest applicable ECL identified during this review is declared. For example:
If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared.
There is no "additive" effect from multiple EALs meeting the same ECL. For example:
If two Alert EALs are met, an Alert will be declared.
Related guidance for classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification ofNRC Guidance for Emergency Notifications During Quickly Changing Events.
2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time an event or condition occurred, and prior to any plant or operator response, determines whether 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.
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2.5 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 will be made as though 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.
2.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification. Termination criteria contained in procedure NMP-EP-110, Emergency Classification and Initial Actions shall be completed for an event to be terminated. At termination, on an event specific basis, the site enter either normal operating conditions or a recovery condition with a recovery organization established for turnover from the ERO. Guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02.
2.7 CLASSIFICATION OF SHORT-LIVED EVENTS Event-based I Cs 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 end 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. For example an earthquake, or failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip.
2.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs and EALs in this document employ time-based criteria that require 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, some transient conditions may cause an EAL to be met for a brief period of time. The following guidance should be applied to the classification of these conditions.
EAL momentarily met during expected plant response - When 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 applicable EAL is not considered met and the associated emergency declaration is not required. This example presents an illustration:
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An A TWS occurs and the auxiliary feedwater system fails to automatically start.
Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification will be based on the ATWS only.
It is important to stress that the 15-minute emergency classification assessment period 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 where 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.
2.9 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. Personnel could discover an event or condition existed that met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. It may be the event or condition was not recognized at the time, or there was an error in the emergency classification process.
In these cases, no emergency declaration is warranted, but the guidance in NUREG-1022 is applicable. Specifically, the event will be reported to the NRC in accordance with 10 CFR &sect; 50.72 within one hour of the discovery of the undeclared event or condition. The licensee will also notify appropriate state and local agencies in accordance with the agreed upon arrangements.
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3  ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY              EMERGENCY RGl Release of          RSl Release of            RAl Release of          RUl Release of gaseous radioactivity    gaseous radioactivity      gaseous or liquid      gaseous or liquid resulting in offsite    resulting in offsite      radioactivity resulting radioactivity greater dose greater than 1,000  dose greater than 100      in offsite dose greater than 2 times the mrem TEDE or 5,000      mrem TEDE or 500          than 10 mrem TEDE      ODCM limits for 60 mrem thyroid CDE.        mrem thyroid COE.          or 50 mrem thyroid      minutes or longer.
Op. Modes: All          Op. Modes: All            COE.                    Op. Modes: All Op. Modes: All RG2 Spent fuel pool      RS2 Spent fuel pool        RA2 Significant        RU2 UNPLANNED level cannot be          level at 195 foot level. lowering of water level loss of water level restored to at least 195 Op. Modes: All            above, or damage to,    above irradiated fuel.
foot level for 60                                  irradiated fuel.        Op. Modes: All minutes or longer.                                  Op. Modes: All Op. Modes: All RA3 Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All 9
 
RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #1 will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(1)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Plant Vent RE-12444E                          50 &#xb5;Ci/cc Turbine Building Vent (SJAE) RE-12839E        2.1 x I 0 3 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 1,000 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 5,000 mrem for one hour of inhalation.
Basis:
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 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 monitor reading threshold values are 10
 
determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used to determine the monitor reading threshold values in ICs RS 1 and RA 1. This protocol will maintain intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at the EPA PAO of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAO 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.
                                            /
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RG2 ECL: General Emergency Initiating Condition: Spent fuel pool level cannot be restored to at least 195 foot level for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
(I)      Spent fuel pool level cannot be restored to at least 195 foot level for 60 minutes or longer.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. The spent fuel pool level instrument is located outside the control room but in close proximity. 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.
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RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #I will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(I)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Plant Vent RE-12444E                          5.0 &#xb5;Ci/cc Turbine Building Vent (SJAE) RE-12839E        2.1 x 102 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the site boundary.
(3)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10 percent 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 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 monitor reading threshold values are 13
 
determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in I Cs RG 1 and RA 1. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at 10 percent of the EPA PAO of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1 :5 ratio of the EPA PAO 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 uses IC RO 1.
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RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at 195 foot level.
Operating Mode Applicability: All Emergency Action Levels:
(1)    Lowering of spent fuel pool level to 195 foot level.
Basis:
This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. The spent fuel pool level instrument is located outside the control room but in close proximity. This condition stems from major failures of plant functions needed to protect the public that 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 uses IC RG 1 or RG2.
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RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than I 0 mrem TEDE or 50 mrem thyroid CDE.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3 or 4)
Notes:
* The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.
* 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.
* The pre-calculated effluent monitor values presented in EAL #I will be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.
(I)    Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer:
Plant Vent RE-12444E                          0.50 &#xb5;Ci/cc Turbine Building Vent (SJAE) RE-12839D        2.1x10 1 &#xb5;Ci/cc (2)    Dose assessment using actual meteorology indicates doses greater than I 0 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary.
(3)    Analysis of a liquid effluent sample indicates a concentration or release rate that wo.uld result in doses greater than I 0 mrem TEDE or 50 mrem thyroid CDE at or beyond the site boundary for one hour of exposure.
(4)    Field survey results indicate EITHER of the following at or beyond the site boundary:
* Closed window dose rates greater than I 0 mR/hr expected to continue for 60 minutes or longer.
* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.
Basis:
This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to I percent 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 plant safety as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).
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Radiological effluent EALs are 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 monitor reading threshold values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term (noble gases, particulates, and halogens) used is the same as those used to determine the monitor reading threshold values in ICs RGI and RSI. This protocol maintains intervals between the threshold values for the three classifications. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and the threshold values are based on a site boundary (or beyond) dose of 10 mR/hour whole body or 50 mR/hour thyroid, whichever is more limiting.
The TEDE dose is set at I percent of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the I :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 uses IC RS I.
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RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
(1)    Uncovery of irradiated fuel in the REFUELING PATHWAY.
(2)    Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by a HIGH Alarm on ANY of the following radiation monitors:
Fuel Handling Building RE-008 CNMT BLDG Low Range** RE-002/003                **Mode 6 only during fuel movement Fuel Handling BLDG EFFL. ARE-2532 A/B Fuel Handling BLDG EFFL. ARE-2533 AIB (3)    Lowering of spent fuel pool level to 204 feet (Level 2).
Basis:
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
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 plant safety. The spent fuel pool level instrument is located outside the control room but in close proximity.
This IC 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 IC E-HUl.
Escalation of the emergency is based on either Recognition Category R or C I Cs.
EAL#l This EAL escalates from RU2. 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), 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 will be based on the totality of available indications, reports and observations.
18
 
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 the fuel is actually uncovered. To the degree possible, readings will be considered in combination with other available indications of inventory loss.
A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance with Recognition Category C during the Cold Shutdown and Refueling modes.
EAL#2 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 will be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).
EAL#3 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 is 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 uses ICs RSI or RS2.
19
 
RA3 ECL: Alert Initiating Condition: Radiation levels that impede access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels: (I or 2)
Note: 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.
(I)    Dose rate greater than 15 mR/hr in ANY of the following areas:
* Control Room (RE-001)
* Central Alarm Station (Survey Only)
(2)    An UNPLANNED event results in radiation levels that prohibit or impede access to any Table HI plant rooms or areas:
Table Hl Applicable Building                  Room Number Mode 1CB-226, 1CB-A45, 3
2CB-223, 2CB-A22 1CB-A77, 1CB-B6 l, 1CB-B76, 1CB-B79 3
2CB-A79, 2CB-B01 Control Building    2CB-B04, 2CB-B 18 1CB-226, 1CB-A45 1CB-B84, 2CB-B85                            4 2CB-223, 2CB-A22 1CB-A48, 1CB-A50 4
2CB-Al5, 2CB-Al6 AFW Pump Operation and standby AFW Pump House                                                1, 2, 3 Readiness 1AB-A28, 2AB-A72 1, 2, 3 A-level demin vessel valve galleries 1AB-A24, 2AB-A77                            3 1AB-A08, 2AB-Al01                          3 Auxiliary Building    1AB-C85, 1AB-C89 4
2AB-C38, 2AB-C44 1AB-Bl5 MEZZ 1AB-Bl9 MEZZ 4
2AB-Bl 17 MEZZ 2AB-B 119 MEZZ Basis:
UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution 20
 
or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.
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 plant safety. The emergency director should consider the cause of the increased radiation levels and determine if another IC may be applicable.
For EAL #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 will 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.
* 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 I 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.
Escalation of the emergency classification level uses Recognition Category R, C or FI Cs.
21
 
RU1 ECL: Notification of Unusual Event Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)
Notes:
* The emergency director will declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.
* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.
* 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.
(1)      Reading on ANY of the following effluent radiation monitor greater than 2 times the ODCM limits for 60 minutes or longer:
SG Blowdown Effluent Line (RE-0021)                2 x release permit setpoint Turbine Building Drain Effluent Line (RE-0848)    2 x release permit setpoint Plant Vent (RE-12442C)                            2 x release permit setpoint Plant Vent (RE- l 2444C)                          2 x release permit setpoint Turbine Building Vent, SJAE (RE-12839C)
No Confirmed Primary-Secondary Leakage        1.6 x 10-3 &#xb5;Ci/cc Confirmed Primary-Secondary Leakage      2 x release permit setpoint (2)      Reading on ANY of the following effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.
Liquid Radwaste Effluent Line (RE-0018)          2 x release permit setpoint Waste Gas Process Effluent Line (ARE-0014)      2 x release permit setpoint (3)      Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the ODCM limits for 60 minutes or longer.
Basis:
This IC addresses a potential decrease in the level of plant safety 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.
22
 
Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment indicates degradation in these features and/or controls.
Radiological effluent EALs are 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 will not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.
EAL #1 - This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.
EAL #2 - This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL will typically be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).
EAL #3 - 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 uses IC RA I.
23
 
RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:
Personnel report of low water level LSHL-0625 SFP low level Alarm (ALB05 E02)
AND
: b.      UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.
RE-0008 in the spent fuel pool building RE-0002, -0003, -0004 in containment
* RE-0011 at the seal table*
RE-0005, -0006 in containment *
                  *Not applicable in Modes 1-4 Basis:
UNPLANNED: A parameter change or an event that is not I) 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: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition can be a precursor to a more serious event and indicates a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of plant safety.
A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications 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.
24
 
The effects of planned evolutions will 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 uses IC RA2.
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4  COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS GENERAL                SITE AREA ALERT          UNUSUAL EVENT EMERGENCY              EMERGENCY CGl Loss ofRPV          CSl Loss ofRPV        CAl Loss ofRPV        CUl UNPLANNED inventory affecting      inventory affecting    inventory.            loss of RPV inventory fuel clad integrity with core decay heat        Op. Modes: Cold        for 15 minutes or containment              removal capability. Shutdown, Refueling    longer.
challenged.              Op. Modes: Cold                              Op. Modes: Cold Op. Modes: Cold          Shutdown, Refueling                          Shutdown, Refueling Shutdown, Refueling CA2 Loss of all        CU2 Loss of all but offsite and all onsite one AC power source AC power to            to emergency buses for emergency buses for    15 minutes or longer.
15 minutes or longer. Op. Modes: Cold Op. Modes: Cold        Shutdown, Refueling, Shutdown, Refueling,  Defueled Defueled CA3 Inability to      CU3 UNPLANNED maintain the plant in  increase in RCS cold shutdown.        temperature.
Op. Modes: Cold        Op. Modes: Cold Shutdown, Refueling    Shutdown, Refueling CU4 Loss of Vital DC power for 15 minutes or longer.
Op. Modes: Cold Shutdown, Refueling CU5 Loss of all onsite or offsite communications capabilities.
Op. Modes: Cold Shutdown, Refueling, Defueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Cold Shutdown, Refueling 26
 
CG1 ECL: General Emergency Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the General Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(1)    a.      RPV level less than 181 '-10" [TOAF] (63% on RVLIS full range) for 30 minutes or longer.
AND
: b.      ANY indication from the Containment Challenge Table Cl.
(2)    a.      RPV level cannot be monitored for 30 minutes or longer.
AND
: b.      Core uncovery is indicated by ANY of the following:
RE-005 OR 006                                                  j 2: 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery AND
: c.      ANY indication from the Containment Challenge Table Cl.
Containment Challenge Table Cl CONTAINMENT CLOSURE NOT established*
Explosive mixture inside containment - greater than OR equal to 6% H2 UNPLANNED increase in containment pressure
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 30 minute time limit, then declaration of a General Emergency is not required.
Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling 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.
27
 
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 RCS/reactor vessel 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 bum (i.e.,.at the lower deflagration limit).
A hydrogen bum 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 containment is challenged.
In EAL I .a, RVLIS is used to determine when reactor water level is less thari TOAF. RVLIS indication is only available during Mode 5 up to the point of reactor head disassembly prior to Mode 6 entry. Once RVLIS becomes unavailable classification ofIC CGI is accomplished in accordance with EAL2.
In EAL 2.b, 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 actions to terminate leakage, recover inventory control or 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 indicate leakage from the RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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.
28
 
CS1 ECL: Site Area Emergency Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2 or 3)
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.
(I)    a.      CONTAINMENT CLOSURE not established.
AND
: b.      RPV level less than 185 '-4" [6" below Bottom ID of loop] (72% on Full Range RVLIS).
(2)    a.      CONTAINMENT CLOSURE established.
AND
: b.      RPV level less than 181 '-10" [TOAF] (63% on RVLIS full range).
(3)    a.      RPV level cannot be monitored for 30 minutes or longer.
AND
: b.      Core uncovery is indicated by ANY of the following:
RE-005 OR 006                                      I~ 40 REM/hr Erratic Source Range monitor indication UNPLANNED increase in Containment Sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tanks (WHT) levels of sufficient magnitude to indicate core uncovery Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling 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.
This IC addresses a significant and prolonged loss of RPV 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 to protect the public and warrant a Site Area Emergency declaration.
Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause 29
 
reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.
In EALs l .a and 2.a the specified levels represent reactor vessel levels that are lower than the monitoring capability of RCS level instrumentation and therefore must be monitored using RVLIS.
This level will only be observable in Mode 5 with RVLIS in operation. In Mode 6 or when RVLIS is not in operation the IC should be evaluated suing EAL 3.
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 EALs l .b and 2.b reflects that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment.
In EAL 3.a, 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 actions to terminate leakage, recover inventory control or 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 indicate leakage from the RPV.
These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation ofShutdown 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 uses IC CG 1 or RG 1.
30
 
CA1 ECL: Alert Initiating Condition: Loss of RPV inventory.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      Loss ofRPV inventory as indicated by level less than elevation 185'-IO" (73% on Full Range RVLIS).
(2)      a.      RPV level cannot be monitored for 15 minutes or longer AND
: b.      UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT) or Waste Holdup Tank (WHT) levels due to a loss of RPV inventory.
Basis:
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.
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 EAL #I, a lowering of water level below the bottom ID of the RCS Loop setpoint (187' 6")
indicates that operator actions have not been successful in restoring and maintaining RPV water level.
The 187' 6" level specified in the EAL is the minimum RCS level for RHR operation as outlined in the procedure for mid-loop operations. Below this level, loss of RHR pump net positive suction head (NPSH) may occur resulting in a loss of decay heat removal capability. 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.
Although related, EAL #1 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 Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.
For EAL #2, 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 indicate leakage from the RPV.
31
 
The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the RPV inventory level continues to lower, then escalation to Site Area Emergency uses IC CS 1.
32
 
CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note:    The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC Power (Table Sl) to BOTH 1(2)AA02 AND 1(2)BA03 for 15 minutes or longer.
Table Sl Unit 1                                    Unit2 Unit Auxiliary Transformer 1NXAA          Unit Auxiliary Transformer 2NXAA Unit Auxiliary Transformer 1NXAB          Unit Auxiliary Transformer 2NXAB Reserve Auxiliary Transformer 1NXRA        Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer 1NXRB        Reserve Auxiliary Transformer 2NXRB Diesel Generator 1A                        Diesel Generator 2A Diesel Generator IB                        Diesel Generator 2B Standby Auxiliary Transformer ANXRA        Standby Auxiliary Transformer ANXRA Basis:
This IC addresses a total loss of AC power (see Table Sl above) 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.
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 emergency 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. When in these modes, this condition represents an actual or potential substantial degradation of the level of plant safety.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses IC CSl or RSI.
33
 
CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (1 or 2)
Note: The emergency director will declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(I)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F for greater than the duration specified in the following table.
Table C2: RCS Heat-up Duration Thresholds RCS Status              Containment Closure Status            Heat-up Duration Not Intact                    Not Established                  0 minutes (or at reduced inventory)              Established                    20 minutes*
Intact Not applicable                  60 minutes*
(but not at reduced inventory)
* If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.
(2)    UNPLANNED RCS pressure increase greater than 10 psig. (This EAL does not apply during water-solid plant conditions.)
Basis:
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling 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.
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 plant safety.
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 the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory, and CONTAINMENT CLOSURE is not established. In this case, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.
34
 
The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operation in PWRs). The 20-minute criterion was included to allow time for operator action to address the temperature increase.
Finally, 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 will allow sufficient time to address the temperature increase without a substantial degradation in plant safety.
EAL #2 provides a pressure-based indication of RCS heat-up.
Escalation of the emergency classification level uses IC CS 1 or RS 1.
35
 
CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
(1)      a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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 may 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.
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.
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.
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 36
 
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 plant safety.
The first threshold for EAL l .b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL 1.b 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 all 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 uses IC CS 1 or RS 1.
37
 
CU1 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of RPV inventory for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for 15 minutes or longer.
(2)    a.      RPV level cannot be monitored.
AND
: b.      UNPLANNED increase in Containment sump, Reactor Coolant Drain Tank (RCDT),
or Waste Holdup Tank (WHT) levels.
Basis:
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.
This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of plant safety.
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.
EAL #1 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.
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.
EAL #2 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 38
 
and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they indicate leakage from the RPV.
Continued loss of RCS inventory may result in escalation to the Alert emergency classification level using either IC CAI or CA3.
39
 
CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or
* longer.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      a.      AC power capability to BOTH I (2)AA02 AND I (2)BA03 is reduced to a single power source (Table SI) for 15 minutes or longer.
AND
: b.      Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                                  Unit2 Unit Auxiliary Transformer INXAA        Unit Auxiliary Transformer 2NXAA Unit Auxiliary Transformer INXAB        Unit Auxiliary Transformer 2NXAB Reserve Auxiliary Transformer INXRA Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer INXRB      Reserve Auxiliary Transformer 2NXRB Diesel Generator IA                    Diesel Generator 2A Diesel Generator I B                    Diesel Generator 2B Standby Auxiliary Transformer ANXRA Standby Auxiliary Transformer ANXRA Basis:
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.
This IC describes a significant degradation of offsite and onsite AC power sources (see Table S 1 above) where 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. When in these modes, this condition is considered to be a potential degradation of the level of plant safety.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Examples of this condition include:
40
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g.,
an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes is the 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.
41
 
CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels: (I or 2)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    UNPLANNED increase in RCS temperature to greater than 200 &deg;F.
(2)    Loss of ALL RCS temperature and RPV level indication for 15 minutes or longer.
Basis:
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.
This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of plant safety. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the emergency director will 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.
EAL # 1 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, where 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 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.
EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators are 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 is the threshold to exclude transient or momentary losses of indication.
Escalation to Alert uses IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.
42
 
CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.
Operating Mode Applicability: Cold Shutdown, Refueling Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105 VDC on required 125 VDC buses 1(2)AD1, 1(2)BD1, 1(2)CD1, 1(2)DD1for15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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.
This condition is considered to be a potential degradation of the level of plant safety.
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 Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable),
then a loss of vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of vital DC power to Train A would not warrant an emergency classification.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Depending upon the event, escalation of the emergency classification level uses IC CAI or CA3, or an IC in Recognition Category R.
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CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Cold Shutdown, Refueling, Defueled Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will be assessed only when extraordinary means are being used 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.).
EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Georgia and South Carolina; Burke County, Georgia; Aiken County, South Carolina: Barnwell and Allendale, South Carolina; and the Savannah River Site in South Carolina.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
44
 
5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS UNUSUAL EVENT E-HUl Damage to a loaded cask CONFINEMENT BOUNDARY.
Op. Modes: All 45
 
E-HU1 ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than ANY of the values listed in Table El.
Table El Location of Dose Rate                            Total Dose Rate (Neutron + Gamma mR/hr)
HI-TRAC 125 Side - Mid-height                                  950 Top                                          200 HI-STORM 100 Side - 60 inches below mid-height                            170 Side - Mid-height                                  180 Side - 60 inches above mid-height                          110 Center of lid*                                    50 Middle of top lid**                                  60 Top (outlet) duct                                130 Bottom (inlet) duct                                360
* The  center of the  top lid represents a 6 in. radius.
          ** The middle of the top lid represents an approximately 4 in. wide cylindrical "strip" located about mid-distance of the lid.
Basis:
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
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 radiation reading values listed in the table represent 2 times the site-specific cask specific technical specification allowable radiation level on the designated surface of the spent fuel cask. The technical specification multiple of"2 times" is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of 46
 
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, determining if the "on-contact" dose rate limit is exceeded may be based on measurement of a dose rate at some distance from the cask.
Security-related events for ISFSis are covered under ICs HUI and HAI.
47
 
6 FISSION PRODUCT BARRIER ICS/EALS                          LOSS POTENTIAL LOSS LOSS                          LOSS POTENTIAL LOSS FUEL CLAD                                        CONTAINMENT Recognition Category "F" Initiating Condition Matrix GENERAL EMERGENCY Loss of any two barriers and Loss or Potential Loss of the third barrier.                                              /Loss of at least 2
                                                                                                                      -  YES-    FGI -Loss of ANY Two Barriers AND Loss or Barriers?                      Potential Loss of Third Barrier FGl Op. Modes: Power Operation, Hot Standby, Startuv, Hot Shutdown
                                                                                                        '----------NO--~
SITE AREA EMERGENCY Loss or Potential Loss of any two barriers. LOSS POTENTIAL LOSS                                  LOSS POTENTIAL LOSS FUEL CLAD                                        CONTAINMENT FSl Op. Modes: Power Operation, Hot Standby, Startup, Hot Shutdown ALERT                                                                                                  FS1 - Loss or Potential Loss of ANY Two Barriers Any Loss or any Potential Loss of either the Fuel Clad or RCS barrier.
FAl POTENTIAL          POTENTIAL LOSS                LOSS Op. Modes: Power Operation, Hot Standby,                  LOSS              LOSS FUEL CLAD              RCS Startup, Hot Shutdown FA I - ANY Loss or ANY Potential Loss of EITHER Fuel Clad ill! RCS 48
 
Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGl GENERAL EMERGENCY                        FSl SITE AREA EMERGENCY                                FAlALERT Loss of any two barriers and Loss or          Loss or Potential Loss of any two barriers. Any Loss or any Potential Loss of either Potential Loss of the third barrier.                                                      the Fuel Clad or RCS barrier.
Fuel Clad Barrier*
* RCS Barrier                                Containment Barrier LOSS              POTENTIAL LOSS                  LOSS              POTENTIAL LOSS              LOSS              POTENTIAL LOSS
: 1. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage                      1. RCS or SG Tube Leakage                !
Not Applicable          A. CORE COOLING        A. An automatic or        A. Operation of a    A. A leaking or        Not Applicable  '
CSF-ORANGE            manual ECCS                standby charging      RUPTURED SG is entry conditions      actuation is required      pump is required      FAULTED outside met.                  by EITHER of the          by EITHER of the      of containment.
following:                following:
: 1. UNISOLABLE            1. UNISOLABLE RCS leakage                RCS leakage OR                        OR
: 2. SG tube                2. SG tube RUPTURE.                    leakage.
OR B. RCS INTEGRITY CSF - RED entry conditions met 49
 
Fuel Clad Barrier                                RCS Barrier                                  Containment Barrier LOSS            POTENTIAL LOSS                  LOSS          POTENTIAL LOSS                LOSS              POTENTIAL LOSS
: 2. Inadequate Heat Removal                    2. Inadequate Heat Removal                    2. Inadequate Heat Removal A. CORE COOLING A. CORE COOLING                Not Applicable        A. HEAT SINK CSF -      Not Applicable          A. CORE COOLING CSF - RED entry        CSF-ORANGE                                    RED entry                                        CSF - RED entry conditions met          entry conditions                              conditions met.                                  conditions met for met                                                                                            15 minutes or longer OR                                        Note: Heat Sink CSF B. HEAT SINK CSF -                            should not be RED entry                                  considered RED if total conditions met                            available feedwater flow is less than 535 Note: Heat Sink CSF                            gpm due to operator should not be                                  action.
considered RED if total available feedwater flow is less than 535 gpm due to operator action.
50
 
Fuel Clad Barrier                          RCS Barrier                            Containment Barrier LOSS              POTENTIAL LOSS          LOSS              POTENTIAL LOSS        LOSS              POTENTIAL LOSS
: 3. RCS Activity I Containment Radiation  3. RCS Activity I Containment Radiation  3. RCS Activity I Containment Radiation A. Containment          Not Applicable  A. Containment            Not Applicable  Not Applicable          A. Containment radiation monitor                        radiation monitor                                                  radiation monitor RE-005 OR 006                            RE-005 OR 006                                                      RE-005 OR 006
    ~ 2.6E+5 mR/hr.                        ~ 8.7 E+2 mR/hr.                                                  ~ l.3E+7 mR/hr.
OR B. Indications that reactor coolant activity is greater than 300 &#xb5;Ci/gm dose equivalent I-131.
51
 
Fuel Clad Barrier                                      RCS Barrier                                        Containment Barrier LOSS            POTENTIAL LOSS                    LOSS              POTENTIAL LOSS                      LOSS                POTENTIAL LOSS
: 4. Containment Integrity or Bypass                4. Containment Integrity or Bypass                  4. Containment Integrity or Bypass Not Applicable          Not Applicable            Not Applicable              Not Applicable            A. Containment isolation    A. CONTAINMENT CSF is required                RED entry conditions AND                          met.
EITHER of the              OR following:              B. CTMT hydrogen
: 1. Containment              concentration greater integrity has been    than 6%
lost based on          OR Emergency Director
: c. 1. Containment pressure greater than judgment.
21.5 psig.
OR AND
: 2. UNISOLABLE
                                                '                                                                                        2. Less than Four pathway from the CTMT fan coolers containment to and one full train the environment of CTMT Spray exists.
are operating per OR                                design for 15 B. Indications of RCS                minutes or longer.
leakage outside of containment.
: 5. Other Indications                              5. Other Indications                                  5. Other Indications Not applicable          Not applicable            Not applicable              Not applicable            Not applicable              Not applicable
: 6. Emergency Director Judgment                    6. Emergency Director Judgment                        6. Emergency Director Judgment A. ANY condition in A. ANY condition in            A. ANY condition in the A. ANY condition in          A. ANY condition in        A. ANY condition in the the opinion of the      the opinion of the        opinion of the              the opinion of the      the opinion of the          opinion of the emergency                emergency director        emergency director          emergency director        emergency director          emergency director director that            that indicates            that indicates loss of      that indicates          that indicates loss of      that indicates indicates loss of        potential loss of the    the RCS Barrier.            potential loss of the    the containment              potential loss of the the fuel clad            fuel clad barrier.                                    RCS Barrier.              barrier.                    containment barrier.
barrier.
52
 
Basis Information For Fission Product Barrier EALs FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.
: 1. RCS or SG Tube Leakage There is no Loss threshold associated with RCS or SG Tube Leakage.
Potential Loss 1.A This condition indicates a reduction in reactor vessel water level sufficient to allow the onset of heat-induced cladding damage.
: 2. Inadequate Heat Removal Loss 2.A This condition indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.
Potential Loss 2.A This condition indicates temperatures within the core are sufficient to allow the onset of heat-induced cladding damage.
Potential Loss 2.B This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss ofthe Fuel Clad Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to RCS Barrier Potential Loss threshold 2.A; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity I Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300&#xb5;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 percent to 5 percent 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.
53
 
The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold 3 .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.
Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2 percent to 5 percent 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.
It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the fuel clad barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
54
 
RCS BARRIER THRESHOLDS:
The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief valves, and other connections up to and including the primary isolation valves.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold is based on an UNISOLABLE RCS leak of sufficient size to require an automatic or manual actuation of the Emergency Core Cooling System (ECCS). This condition clearly represents a loss of the RCS Barrier.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require a safety injection is considered to be RUPTURED. If a RUPTURED steam generator is also FA UL TED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met.
Potential Loss I .A This threshold is based on an UNISOLABLE RCS leak that results in the inability to maintain pressurizer level within specified limits by operation of a normally used charging (makeup) pump, but an ECCS actuation has not occurred. The threshold is met when an operating procedure, or operating crew supervision, directs that a standby charging (makeup) pump be placed in service to restore and maintain pressurizer level.
This threshold is applicable to unidentified and pressure boundary leakage, as well as identified leakage. It is also applicable to UNISOLABLE RCS leakage through an interfacing system. The mass loss may be into any location - inside containment, to the secondary-side (i.e., steam generator tube leakage) or outside of containment If a leaking steam generator is also FA UL TED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold I .A will also be met.
Potential Loss I .B This condition indicates an extreme challenge to the integrity of the RCS pressure boundary due to pressurized thermal shock - a transient that causes rapid RCS cooldown while the RCS is in Mode 3 or higher (i.e., hot and pressurized).
: 2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss 2.A 55
 
This condition indicates an extreme challenge to the ability to remove RCS heat using the steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RCS Barrier. In accordance with EOPs, there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical to Fuel Clad Barrier Potential Loss threshold 2.B; both will be met. This condition warrants a Site Area Emergency declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and increase RCS pressure to the point where mass will be lost from the system.
: 3. RCS Activity I Containment Radiation Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 3.A since it indicates a loss of the RCS Barrier only.
There is no Potential Loss threshold associated with RCS Activity I Containment Radiation.
: 4. Containment Integrity or Bypass Not Applicable (included for numbering consistency)
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the RCS Barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
56
 
CONTAINMENT BARRIER THRESHOLDS:
The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1.      RCS or SG Tube Leakage Loss I.A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FA UL TED outside of containment. The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RCS Barrier Potential Loss l.A and Loss l.A, respectively. This condition represents a bypass of the containment barrier.
FAULTED is a defined term within the NEI 99-01 methodology. This determination is not necessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if the pressure in a steam generator is decreasing uncontrollably [part of the FAULTED definition] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lower bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values).
This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant, or to drive an auxiliary (emergency) feed water pump. These types of conditions will result in a significant and sustained release of radioactive steam to the environment (similar to a FAUL TED condition). The inability to isolate the steam flow without an adverse effect on plant cooldown meets the intent of a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period of time following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Steam releases associated with the unexpected operation of a valve (e.g., a stuck-open safety valve) do meet this threshold.
57
 
Following an SG tube leak or rupture, there may be minor radiological releases through a secondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing, etc.). These types of releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
The emergency classification levels resulting from primary-to-secondary leakage, with or without a steam release from the FAULTED SG, are summarized below.
Affected SG is FAULTED Outside of Containment?
P-to-S Leak Rate                        Yes                        No Less than or equal to 25 gpm            No classification          No classification Greater than 25 gpm                Unusual Event per SU4      Unusual Event per SU4 Requires operation of a Site Area Emergency standby charging (makeup)                                          Alert per FA 1 per FSI pump (RCS Barrier Potential Loss)
Requires an automatic or            Site Area Emergency Alert per FAl manual ECCS (SI) actuation                  per FSI (RCS Barrier Loss)
There is no Potential Loss threshold associated with RCS or SG Tube Leakage.
: 2. Inadequate Heat Removal There is no Loss threshold associated with Inadequate Heat Removal.
Potential Loss 2.A This condition represents an IMMINENT core melt sequence that, if not corrected, could lead to vessel failure and an increased potential for containment failure. For this condition to occur, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to core melting and a subsequent challenge of the Containment Barrier.
The restoration procedure is considered "effective" if core exit thermocouple readings are decreasing or if reactor vessel level is increasing. Whether the procedure(s) will be effective should be apparent within 15 minutes. The emergency director should escalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it 58
 
is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
: 3. RCS Activity I Containment Radiation There is no Loss threshold associated with RCS Activity I Containment Radiation.
Potential Loss 3.A The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that 20 percent 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 percent 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 containment that would then escalate the emergency classification level to a General Emergency.
: 4. Containment Integrity or Bypass Loss 4.A These thresholds address a situation where containment isolation is required and one of two conditions exists as discussed below. There may be accident and release conditions that simultaneously meet both thresholds 4.A.1 and 4.A.2.
4.A. l - Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (sometimes referred to as design leakage). Following the release of RCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure. Recognizing the inherent difficulties in determining a containment leak rate during accident conditions, it is expected that the emergency director will assess this threshold using judgment, and with due consideration given to current plant conditions, and available operational and radiological data (e.g., containment pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).
Two simplified examples are provided in the middle piping run of Figure 6-F-l. One is leakage from a penetration and the other is leakage from an in-service system valve.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure.
Another example is a loss or potential loss of the RCS barrier, and the simultaneous occurrence of two FA UL TED locations on a steam generator where one fault is located inside containment (e.g., on a steam or feedwater line) and the other outside of 59
 
containment. In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R ICs.
4.A.2 - Conditions are such that there is an UNISOLABLE pathway for the migration of radioactive material from the containment atmosphere to the environment. As used here, the term "environment" includes the atmosphere of a room or area, outside the containment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,
through discharge of a ventilation system or atmospheric leakage). Depending upon a variety of factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
See the simplified example in the top piping run of Figure 6-F-l. The inboard and outboard isolation valves remained open after a containment isolation was required (i.e.,
containment isolation was not successful). There is now an UNISOLABLE pathway from the containment to the environment.
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.
Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
A simplified example is shown in the bottom piping run of Figure 6-F-l. Leakage in an RCP seal cooler is allowing radioactive material to enter the Auxiliary Building. The radioactivity would be detected by the Process Monitor. Ifthere is no leakage from the closed water cooling system to the Auxiliary Building, then no threshold has been met. If the pump or system piping developed a leak that allowed steam/water to enter the Auxiliary Building, then threshold 4.B would be met. Depending upon radiation monitor locations and sensitivities, this leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A. l to be met as well.
Following the leakage of RCS mass into containment and a rise in containment pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components. Minor releases may also occur if a containment isolation valve(s) fails to close but the containment atmosphere escapes to a closed system. These releases do not constitute a loss or potential loss of containment but should be evaluated using the Recognition Category R !Cs.
The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold I .A.
60
 
Loss 4.B Containment sump, temperature, pressure and/or radiation levels will increase if reactor coolant mass is leaking into the containment. If these parameters have not increased, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). Increases in sump, temperature, pressure, flow and/or radiation level readings outside of the containment may indicate that the RCS mass is being lost outside of containment.
Unexpected elevated readings and alarms on radiation monitors with detectors outside containment will be corroborated with other available indications to confirm that the source is a loss of RCS mass outside of containment. If the fuel clad barrier has not been lost, radiation monitor readings outside of containment may not increase significantly.
However, other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc. should be sufficient to determine if RCS mass is being lost outside of the containment.
In the simplified example in the middle piping run of Figure 6-F-1, a leak has occurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building. Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of the four monitors depicted in the figure and cause threshold 4.A.l to be met as well.
To ensure proper escalation of the emergency classification, the RCS leakage outside of containment must be related to the mass loss that is causing the RCS Loss and/or Potential Loss threshold l .A to be met.
Potential Loss 4.A If containment pressure exceeds the design pressure, there exists a potential to lose the Containment Barrier. To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RCS and Fuel Clad barriers would already be lost. This threshold is a discriminator between a Site Area Emergency and General Emergency since there is now a potential to lose the third barrier.
Potential Loss 4.B 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 potential loss of the Containment Barrier.
Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. This threshold represents a potential loss of containment because containment heat removal/depressurization 61
 
systems (e.g., containment sprays, ice condenser fans, etc., but not including containment venting strategies) are either lost or performing in a degraded manner.
: 5. Other Indications Not Applicable (included for numbering consistency)
: 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is lost.
Potential Loss 6.A This threshold addresses any other factors used by the emergency director in determining whether the Containment Barrier is potentially lost. The emergency director should also consider whether to declare the barrier potentially lost in the event that barrier status cannot be monitored.
62
 
Figure 6-F-1: PWR Containment Integrity or Bypass Examples
                                                              ----------1                  *::: 4.A.2 - Airborne:-:*:->:-
I          I
: Effluent I'                        release from Inside Containment I
                                                            ' Monitor*                          .. J?~t~\Y~>'. .
Auxiliary Building
                                                            ~---------
I            I I
I  Process  I I
I            I I
I Monitor    I I
                                                    ~---*--~              Closed Cooling Water System Cooling 63
 
7    HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS GENERAL                SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                EMERGENCY HGl HOSTILE              HSl HOSTILE              HAl HOSTILE            HUl Confirmed ACTION resulting in      ACTION within the        ACTION within the      SECURITY loss of physical        PROTECTED AREA.          OWNER                  CONDITION or control of the facility. Op. Modes: All            CONTROLLED              threat.
Op. Modes: All                                    AREA or airborne        Op. Modes: All attack threat within 30 minutes.
Op. Modes: All HU2 Seismic event greater than OBE levels.
Op. Modes: All HU3 Hazardous event.
Op. Modes: All HU4 FIRE potentially degrading the level of safety of the plant.
Op. Modes: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Op. Modes: All HS6 Inability to          HA6 Control Room control a key safety      evacuation resulting in function from outside    transfer of plant the Control Room.        control to alternate Op. Modes: All            locations.
Op. Modes: All HG7 Other                HS7 Other                HA7 Other              HU7 Other conditions exist which  conditions exist which    conditions exist which  conditions exist in the judgment of the  in the judgment of the    in the judgment of the  which in the emergency director      emergency director        emergency director      judgment of the warrant declaration of  warrant declaration of    warrant declaration of  emergency director a General Emergency. a Site Area              an Alert.              warrant declaration of Op. Modes: All          Emergency.                Op. Modes: All          a (NO)UE.
Op. Modes: All                                    Op. Modes: All 64
 
HG1 ECL: General Emergency Initiating Condition: HOSTILE ACTION resulting in loss of physical control of the facility.
Operating Mode Applicability: All Emergency Action Levels:
(1)      a.      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by security shift supervision.
AND
: b.      EITHER of the following has occurred:
I.      ANY of the following safety functions cannot be controlled or maintained.
* Reactivity control
* Core cooling
* RCS heat removal OR
: 2.      Damage to spent fuel has occurred or is IMMINENT.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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 I) 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.
65
 
Timely and accurate communications between security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program}.
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 Security Plan.
66
 
HG7 ECL: General Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a General Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(I)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a General Emergency.
67
 
HS1 ECL: Site Area Emergency Initiating Condition: HOSTILE ACTION within the PROTECTED AREA.
Operating Mode Applicability: All Emergency Action Levels:
(1)      A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by security shift supervision.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA (PA). This event will require rapid response and assistance due to the possibility for damage to plant equipment.
Timely and accurate communications between security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program}.
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 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 CPR&sect; 73.71or10 CPR&sect; 50.72.
Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate security-sensitive information. This includes information that may be 68
 
                                                                                        --~~--
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 Security Plan.
Escalation of the emergency classification level uses IC HG I.
69
 
HS6 ECL: Site Area Emergency Initiating Condition: Inability to control a key safety function from outside the Control Room.
Operating Mode Applicability: All Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      An event has resulted in plant control being transferred from the control room to the remote shutdown panels due to a control room evacuation.
AND
: b.      Control of ANY of the following key safety functions is not reestablished within 15 minutes.
* Reactivity control
* Core cooling
* RCS heat removal Basis:
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 "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 as to whether the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level uses IC FG I or CG I.
70
 
HS7 ECL: Site Area Emergency Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a Site Area Emergency.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a Site Area Emergency.
                                                            /
71
 
HA1 ECL: Alert Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.
Operating Mode Applicability: All Emergency Action Levels: (I or 2)
(1)      A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by security shift supervision.
(2)      A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control of VEGP security.
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 (PA), or the need to prepare the plant and staff for a potential aircraft impact.
Timely and accurate communications between security shift supervision 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 [and Independent Spent Fuel Storage Installation Security Program].
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, 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.
72
 
Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72.
EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA (OCA).
EAL #2 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 station procedures.
The NRC Headquarters Operations Officer (HOO) will communicate to the licensee ifthe 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 (OCA) 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 Security Plan.
Escalation of the emergency classification level uses IC HSI.
73
 
HAS ECL: Alert Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.
Operating Mode Applicability: All Emergency Action Levels:
Note: 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.
(1)    a.      Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H 1 plant rooms or areas:
AND
: b.      Entry into the room or area is prohibited or impeded.
Table Hl Applicable Building                  Room Number Mode ICB-226, ICB-A45, 3
2CB-223, 2CB-A22 ICB-A77, ICB-B61, ICB-B76, ICB-B79 3
2CB-A79, 2CB-BOI Control Building    2CB-B04, 2CB-B 18 I CB-226, 1CB-A45 1CB-B84, 2CB-B85                          4 2CB-223, 2CB-A22 I CB-A48, 1CB-ASO 4
2CB-Al5, 2CB-Al6 AFW Pump Operation and standby AFW Pump House                                                I, 2, 3 Readiness 1AB-A28, 2AB-A72 I, 2, 3 A-level demin vessel valve galleries IAB-A24, 2AB-A77                          3 IAB-A08, 2AB-Al01                          3 Auxiliary Building    IAB-C85, 1AB-C89 4
2AB-C38, 2AB-C44 1AB-Bl5 MEZZ 1AB-Bl9 MEZZ 4
2AB-B 117 MEZZ 2AB-Bl 19 MEZZ Basis:
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 74
 
cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of plant safety.
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 director's 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 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 4.
* The gas release is a planned activity that includes compensatory measures to address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).
* The action that 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.
An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels.
Most commonly, asphyxiants work by displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19 percent, 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.
Escalation of the emergency classification level uses Recognition Category R, C or F !Cs.
75
 
HA6 ECL: Alert Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations.
Operating Mode Applicability: All Emergency Action Levels:
(1)    An event has resulted in plant control being transferred from the control room to the remote shutdown panels due to a control room evacuation.
Basis:
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 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 uses IC HS6.
76
 
HA7 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of an Alert.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Other conditions exist which, in the judgment of the emergency director, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.
Basis:
HOSTILE ACTION: An act toward a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include      '
I violent acts between individuals in the owner controlled area (OCA)).                                I 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.
77
 
HU1 ECL: Notification of Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.
Operating Mode Applicability: All Emergency Action Levels: (I or 2 or 3)
(I)      A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by security shift supervision.
(2)      Notification of a credible security threat directed at VEGP.
(3)      A validated notification from the NRC providing information of an aircraft threat.
Basis:
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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and represent a potential degradation in the level of plant safety. Security events that do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HAI, HSI and HGI.
Timely and accurate communications between security shift supervision 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 OROs.
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 [and Independent Spent Fuel Storage Installation Security Program].
EAL #I references security shift supervision 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 I 0 CFR &sect; 2.39 information.
EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is 78
 
assessed in accordance with station procedures.
EAL #3 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 station procedures.
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 Security Plan.
Escalation of the emergency classification level uses IC HAL 79
 
HU2 ECL: Notification of Unusual Event Initiating Condition: Seismic event greater than OBE levels.
Operating Mode Applicability: All Emergency Action Levels:
(1)      Seismic event greater than Operating Basis Earthquake (OBE) as indicated by the Seismic Monitoring System (Red OBE Exceedance Indicator).
Basis:
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 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 plant safety.
If a seismic event occurs that exceeds the OBE, an audible alarm will be sounded and the OBE Exceedance Indicator on the Seismic Monitoring System Panel will change from Green to Red.
Event verification with external sources should not be necessary during or following an OBE.
Earthquakes of this magnitude should readily be felt by on-site personnel and recognized as a seismic event (e.g., typical lateral accelerations are 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 or check of internet news sources); 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 uses IC CA6 or SA9.
80
 
HU3 ECL: Notification of Unusual Event Initiating Condition: Hazardous event.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4 or 5)
Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.
(1)    A tornado strike within the PROTECTED AREA.
(2)    Internal room or area flooding of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode.
(3)    Movement of personnel within the PROTECTED AREA (PA) is impeded due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release).
(4)    A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.
(5)    Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours.
Basis:
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
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.
This' IC addresses hazardous events that are considered to represent a potential degradation of the level of plant safety.
EAL #1 addresses a tornado striking (touching down) within the PROTECTED AREA (PA).
EAL #2 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.
EAL #3 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA (PA).
                                              , 81
 
EAL #4 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, or dam failure, or an on-site train derailment blocking the access road.
This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.
EAL #5 addresses the phenomena of the hurricane based on the severe weather mitigation procedure.
Escalation of the emergency classification level is based on I Cs in Recognition Categories A, F, sore.
82
 
HU4 ECL: Notification of Unusual Event Initiating Condition: FIRE potentially degrading the level of safety of the plant.
Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)
Note: The emergency director will declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.
(1)    a.      A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:
* 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
: b.      The FIRE is located within ANY of the Table H2 plant rooms or areas.
(2)    a.      Receipt of a single fire alarm (i.e., no other indications of a FIRE).
AND
: b.      The FIRE is located within ANY of the Table H2 plant rooms or areas.
AND
: c.      The existence of a FIRE is not verified within 30-minutes of alarm receipt.
(3)    A FIRE within the plant PROTECTED AREA (PA) not extinguished within 60-minutes of the initial report, alarm or indication.
(4)    A FIRE within the plant PROTECTED AREA (PA) that requires firefighting support by an offsite fire response agency to extinguish.
Table H2 Containment Building NSCW Cooling Towers Diesel Generator Building Auxiliary Building Fuel Handling Building Control Building Diesel Fuel Oil Storage Tank Pumphouse Auxiliary Feedwater Pumphouse 83
 
Basis:
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 (PA): The area that encompasses all controlled areas within the security protected area fence.
This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of plant safety.
EAL#l 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 include 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, 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.
EAL#2 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 EAL #I is immediately applicable, and the emergency must be declared ifthe 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.
EAL#3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA (PA) not extinguished within 60-minutes may also potentially degrade the level of plant safety.
EAL#4 84
 
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 not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.
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."
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.
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 EAL #2, 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 uses IC CA6 or SA9.
85
 
HU7 ECL: Notification of Unusual Event Initiating Condition: Other conditions exist which in the judgment of the emergency director warrant declaration of a (NO)UE.
Operating Mode Applicability: All Emergency Action Levels:
( 1)    Other conditions exist which in the judgment of the emergency director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
Basis:
This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist that are believed by the emergency director to fall under the emergency classification level description for a NOUE.
86
 
8    SYSTEM MALFUNCTION ICS/EALS I
GENERAL                    SITE AREA ALERT                UNUSUAL EVENT EMERGENCY                    EMERGENCY SG 1 Prolonged loss of      SSl    Loss of all offsite  SAl Loss of all but one    SUl Loss of all offsite all offsite and all onsite  and all onsite AC power        AC power source to        AC power capability to AC power to emergency        to emergency buses for        emergency buses for 15    emergency buses for 15 buses.                      15 minutes or longer.          minutes or longer.        minutes or longer.
Op. Modes: Power            Op. Modes: Power              Op. Modes: Power            Op. Modes: Power Operation, Startup, Hot      Operation, Startup, Hot        Operation, Startup, Hot    Operation, Startup, Hot Standby, Hot Shutdown        Standby, Hot Shutdown        Standby, Hot Shutdown      Standby, Hot Shutdown SA2 UNPLANNED              SU2 UNPLANNED loss of Control Room        loss of Control Room indications for 15          indications for 15 minutes or longer with a    minutes or longer.
significant transient in    Op. Modes: Power progress.                  Operation, Startup, Hot Op. Modes: Power          Standby, Hot Shutdown Operation, Startup, Hot Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SSS Inability to              SAS Automatic or          SUS Automatic or shutdown the reactor          manual trip fails to      manual trip fails to causing a challenge to        shutdown the reactor, and shutdown the reactor.
core cooling or RCS heat      subsequent manual          Op. Modes: Power removal.                      actions taken at the      .Operation Op. Modes: Power              reactor control consoles Operation                      are not successful in shutting down the reactor.
Op. Modes: Power Operation 87
 
GENERAL                  SITE AREA ALERT            UNUSUAL EVENT EMERGENCY                  EMERGENCY SU6    Loss of all onsite or offsite communications capabilities.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7    Failure to isolate containment or loss of containment pressure control.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SGS    Loss of all AC    SSS    Loss of all Vital and Vital DC power        DC power for 15 minutes sources for 15 minutes or or longer.
longer.                  Op. Modes: Power Op. Modes: Power          Operation, Startup, Hot Operation, Startup, Hot  Standby, Hot Shutdown Standby, Hot Shutdown SA9    Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 88
 
SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 4 hours has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND 1(2)BA03.
AND
: b.      EITHER of the following:
* Restoration of at least one AC emergency bus in less than 4 hours is not likely.
* CORE COOLING CSF - RED conditions met.
Basis:
This IC addresses a prolonged loss of all power sources to AC emergency 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 will require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. 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 AC emergency 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 emergency bus will be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success will 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.
89
 
SGS ECL: General Emergency Initiating Condition: Loss of all AC and vital DC power sources for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND I (2)BA03 for 15 minutes or longer.
AND
: b.      Indicated voltage is less than 105 VDC on ALL 125 VDC busses 1(2)AD1, 1(2)BD1, 1(2)CD1, 1(2)DD1 for 15 minutes or longer.
Basis:
This IC addresses a concurrent and prolonged loss of both AC and vital DC power. 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 loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes is the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.
90
 
SS1 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite and ALL onsite AC power to BOTH 1(2)AA02 AND 1(2)BA03 for 15 minutes or longer.
Basis:
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 is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses ICs RG 1, FG 1 or SG 1.
91
 
SSS ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.
Operating Mode Applicability: Power Operation Emergency Action Levels:
Note:      Heat Sink CSP should not be considered RED if total available feedwater flow is less than 535 gpm due to operator action.
(1)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      All manual actions to shutdown the reactor have been unsuccessful.
AND
: c.      EITHER of the following conditions exist:
* Core Cooling CSP - RED conditions met
* Heat Sink CSP - RED conditions met Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 P ICs/EALs. This is appropriate because the Recognition Category P ICs/EALs do not address the additional threat posed by a failure to shutdown 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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level uses IC RG 1 or PG 1.
92
 
SSS ECL: Site Area Emergency Initiating Condition: Loss of all vital DC power for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Indicated voltage is less than 105 VDC on ALL 125 VDC busses 1(2)AD1, 1(2)BDI, 1(2)CD1, 1(2)DDI for 15 minutes or longer.
Basis:
This IC addresses a loss of vital DC power that 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.
Fifteen minutes is the threshold to exclude transient or momentary power losses.
Escalation of the emergency classification level uses I Cs RG 1, FG I or SGS.
93
 
SA1 ECL: Alert Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      AC power capability to BOTH 1(2)AA02 AND 1(2)BA03 is reduced to a single power source (Table S 1) for 15 minutes or longer.
AND
: b.      Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.
Table Sl Unit 1                                    Unit2 Unit Auxiliary Transformer lNXAA          Unit Auxiliary Transformer 2NXAA Unit Auxiliary Transformer 1NXAB          Unit Auxiliary Transformer 2NXAB Reserve Auxiliary Transformer 1NXRA Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer INXRB      Reserve Auxiliary Transformer 2NXRB Diesel Generator I A                      Diesel Generator 2A Diesel Generator 1B                      Diesel Generator 2B Standby Auxiliary Transformer ANXRA Standby Auxiliary Transfonner ANXRA Basis:
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.
This IC describes a significant degradation of offsite and onsite AC power sources (see Table S 1 above) where 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 SUI.
An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.
94
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.
Fifteen minutes is the threshold to exclude transient or momentary losses of power.
Escalation of the emergency classification level uses IC SS 1.
95
 
SA2 ECL: Alert Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Level in at least one steam generator Steam Generator Main or Auxiliary Feed Water Flow AND
: b.      ANY of the following transient events in progress.
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical load rejection greater than 25% full electrical load
* Reactor trip
* ECCS actuation Basis:
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.
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 product barrier challenge is reduced. It represents a potential substantial degradation in the level of plant safety.
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.
Various instrumentation is also used to determine RCS Level - RVLIS, pressurizer level, digital or recorders. A loss of all control room sources for this parameter would also apply.
96
 
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 is 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, core cooling 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, ifthe value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses I Cs FS 1 or IC RS 1.
97
 
SAS ECL: Alert Initiating Condition: Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Operating Mode Applicability: Power Operation Note: A manual 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.
Emergency Action Levels:
(I)      a.      An automatic or manual trip did not shutdown the reactor.
AND
: b.      Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip 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 plant safety. 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 consoles is any operator action, or set of actions, that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).
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 consoles (e.g., locally opening breakers). 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".
The plant response to the failure of an automatic or manual reactor trip 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 shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SSS. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SSS or FS I, an Alert declaration is appropriate for this event.
98
 
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.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
99
 
SA9 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
(1)      a.      The occurrence of ANY of the following hazardous events:
* Seismic event (earthquake)
* Internal or external flooding event
* High winds or tornado strike
* FIRE
* EXPLOSION
* Other events with similar hazard characteristics as determined by the Shift Manager AND
: b.      EITHER of the following:
* Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.
* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.
Basis:
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.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
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.
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.
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 100
 
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 plant safety.
The first threshold for EAL 1.b 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 will be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.
The second threshold for EAL l .b 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 all 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 uses IC FS 1 or RS 1.
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SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)    Loss of ALL offsite AC power capability (Table S2) to BOTH 1(2)AA02 AND 1(2)BA03 for 15 minutes or longer.
Table S2 Unit 1                                  Unit2 Reserve Auxiliary Transformer 1NXRA Reserve Auxiliary Transformer 2NXRA Reserve Auxiliary Transformer 1NXRB      Reserve Auxiliary Transformer 2NXRB Standby Auxiliary Transformer ANXRA Standby Auxiliary Transformer ANXRA Basis:
This IC addresses a prolonged loss of offsite power. The loss of offsite power sources (see Table S2 above) renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of plant safety.
For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.
Fifteen minutes is the threshold to exclude transient or momentary losses of offsite power.
Escalation of the emergency classification level uses IC SA 1.
102
 
SU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer.
Operating Mode Applicability:          Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels:
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(1)      a.      An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.
Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Level in at least one steam generator Steam Generator Main or Auxiliary Feed Water Flow Basis:
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.
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 plant safety.
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.
Various instrumentation is also used to determine RCS Level - RVLIS, pressurizer level, digital or recorders. A loss of all control room sources for this parameter would also apply.
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 is reported if it significantly impairs the capability to perform emergency assessments, particularly those 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, core cooling 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 103
 
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 reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.
Fifteen minutes is the threshold to exclude transient or momentary losses of indication.
Escalation of the emergency classification level uses IC SA2.
104
 
SU3 ECL: Notification of Unusual Event Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)
(1)    RE-48000 reading greater than 5.0 &#xb5;Ci/cc.
(2)    RCS coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits as indicated by ANY of the following:
Dose Equivalent I-131 greater than I &#xb5;Ci/gm for greater than 48 hours Dose Equivalent 1-131 greaterthan Technical Specification figure 3.4.16-1 limits RCS specific activity greater than I OOrE &#xb5;Ci/gm gross radioactivity Basis:
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 plant safety.
Escalation of the emergency classification level uses ICs FA 1 or the Recognition Category R I Cs.
105
 
SU4 ECL: Notification of Unusual Event Initiating Condition: RCS leakage for 15 minutes or longer.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2 or 3)
Note: The emergency director will declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.
(I)      RCS unidentified or pressure boundary leakage greater than 10 gpm for 15 minutes or longer.
(2)      RCS identified leakage greater than 25 gpm for 15 minutes or longer.
(3)      Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.
Basis:
This IC addresses RCS leakage that could 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 plant safety.
EAL #1 and EAL #2 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). EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL 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). EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.
The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. An emergency classification is required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.
Escalation of the emergency classification level uses ICs of Recognition Category R or F.
106
 
SUS ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor.
Operating Mode Applicability: Power Operation Emergency Action Levels: (1 or 2)
Note: A manual 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.
(1)      a.      An automatic trip did not shutdown the reactor.
AND
: b.      A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
(2)      a.      A manual trip did not shutdown the reactor.
AND
: b.      EITHER of the following:
* A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.
* A subsequent automatic trip is successful in shutting down the reactor.
Basis:
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip 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 plant safety.
Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). 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 plant's decay heat removal systems.
If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another Iocation(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.
107
 
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 trip).
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".
The plant response to the failure of an automatic or manual reactor trip 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 SA5. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC SA5 or FA 1, an Unusual Event declaration is appropriate for this event.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor trip 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 trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and will be evaluated.
* If the signal does not cause a plant transient and the trip 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.
108
 
SU6 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.
Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)
(1)    Loss of ALL of the following onsite communication methods:
In plant telephones Public address system Plant radio systems (2)    Loss of ALL of the following ORO communications methods:
ENN (Emergency Notification Network)
Commercial phones (3)    Loss of ALL of the following NRC communications methods:
ENS on Federal Telecommunications System (FTS)
Commercial phones Basis:
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 will be assessed only when extraordinary means are being used 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).
EAL # 1 addresses a total loss of the communications methods used in support of routine plant operations.
EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the states of Georgia and South Carolina; Burke County, Georgia; Aiken County, South Carolina; Barnwell and Allendale, South Carolina; and the Savannah River Site in South Carolina.
EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
109
 
SU7 ECL: Notification of Unusual Event Initiating Condition: Failure to isolate containment or loss of containment pressure control.
Operating Mode Applicability: Power Operatio'n, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2)
(I)    a.      Failure of containment to isolate when required by an actuation signal.
AND
: b.      ALL required penetrations are not closed within 15 minutes of the actuation signal.
(2)    a.      Containment pressure greater than 21.5 psig.
AND
: b.      Less than 4 CTMT fan coolers and one full train of CTMT spray is operating per design for 15 minutes or longer.
Basis:
This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety.
For EAL #1, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure); a failure resulting from testing or maintenance does not warrant classification. The determination of containment and penetration status - isolated or not isolated - will be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible. The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.
This event will escalate to a Site Area Emergency in accordance with IC FS I if there is a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.
110
 
APPENDIX A - ACRONYMS AND ABBREVIATIONS AC ...................................................................................................................... Alternating Current AOP ................................................................................................. Abnormal Operating Procedure A TWS ................................................................................... Anticipated Transient Without Scram BLDG ................................................................................................................................... Building CDE ...................................................................................................... Committed Dose Equivalent CFR ...................................................................................................... Code of Federal Regulations CTMT/CNMT ............................................................................................................... Containment CSF ............................................................................................................. Critical Safety Function CSFST ...................................................................................... Critical Safety Function Status Tree DBA .............................................................................................................. Design Basis Accident DC .............................................................................................................................. Direct Current EAL ........................................................................................................... Emergency Action Level ECCS ............................................................................................ Emergency Core Cooling System ECL ................................................................................................ Emergency Classification Level EFFL .................................................................................................................................... Effluent ENN ............................................................................................. Emergency Notification Network ENS ................................................................................................ Emergency Notification System EOF .................................................................................................. Emergency Operations Facility EOP ............................................................................................... Emergency Operating Procedure EPA ............................................................................................. Environmental Protection Agency FAA ............................................................................................... Federal A via ti on Administration FBI .................................................................................................. Federal Bureau of Investigation FEMA ............................................................................. Federal Emergency Management Agency FTS ......................................................................................... Federal Telecommunications System GA ......................................................................................................................................... Georgia GE ...................................................................................................................... General Emergency HOO .................................................................................. Headquarters Operations Officer (NRC)
IC ........................................................................................................................ Initiating Condition ID ............................................................................................................................. Inside Diameter ISFSI ........................................................................... Independent Spent Fuel Storage Installation Keff .................................................................................... Effective Neutron Multiplication Factor mR, mRem, mrem, mREM ............................................................ milli-Roentgen Equivalent Man NEI ............................................................................................................. Nuclear Energy Institute NPP .................................................................................................................. Nuclear Power Plant NRC .............................................................................................. Nuclear Regulatory Commission NORAD ................................................................. Nmih American Aerospace Defense Command (NO)UE .......................................................................................... (Notification Of) Unusual Event OBE ....................................................................................................... Operating Basis Earthquake OCA ............................................................................................................. Owner Controlled Area ODCM ........................................................................................... Offsite Dose Calculation Manual ORO ................................................................................................ Off-site Response Organization PA .............................................................................................................................. Protected Area PAG ....................................................................................................... Protective Action Guideline PWR ........................................................................................................ Pressurized Water Reactor PSIG ................................................................................................. Pounds per Square Inch Gauge R ......................................................................................................................................... Roentgen RCDT ................................................................................................... Reactor Coolant Drain Tank A-1
 
RCS ............................................................................................................. Reactor Coolant System Rem, rem, REM ...................................................................................... Roentgen Equivalent Man RPS ................................................................................................... ;,..... Reactor Protection System RPV ............................................................................................................. Reactor Pressure Vessel RVLIS ...................................................................... Reactor Vessel Level Instrumentation System SAE ................................................................................................................. Site Area Emergency SC .............................................................................................................................. South Carolina SCBA ......... .... ........... . ...... ....... ...... .......... ... .... .... ...... ... ...... ..... Self-Contained Breathing Apparatus SG ........................................................................................................................... Steam Generator SI .............................................................................................................................. Safety lajection SJAE ............................................................................................................... Steam Jet Air Ejector SNC ....................................................................................................... Southern Nuclear Company SPDS ............................................................................................ Safety Parameter Display System TEDE ............................................................................................. Total Effective Dose Equivalent TOAF .................................................................................................................. Top of Active Fuel VDC .................................................................................................................. Volts Direct Current VEGP ............................................................................................ Vogtle Electric Generating Plant VOiP ................................................................................................... Voice Over Internet Protocol WHT ................................................................................................................. Waste Holdup Tank A-2
 
APPENDIX B - DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.
General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.
Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.
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.
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.
Notification of Unusual Event (NOUE): Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.
The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.
Emergency Action Level (EAL): 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 (ECL): 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 effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in descending order of severity, are:
General Emergency (GE)
Site Area Emergency (SAE)
Alert Notification of Unusual Event (NOUE)
Fission Product Barrier Threshold: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.
Initiating Condition (IC): 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.
B-1
 
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.
CONFINEMENT BOUNDARY: The barrier(s) between areas containing radioactive substances and the environment.
CONTAINMENT CLOSURE: Per Operating Procedure 14210-1/2, Containment Building Penetrations Verification - Refueling.
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 may require a post-event inspection to determine if the attributes of an explosion are present.
FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.
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.
* 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 a nuclear power plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (OCA)).
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.
OWNER CONTROLLED AREA (OCA): The site property owned by or otherwise under the control of VEGP security.
B-2
 
PROJECTILE: An object directed toward an NPP that could cause concern for its continued operability, reliability, or personnel safety.
PROTECTED AREA (PA): The area that encompasses all controlled areas within the security protected area fence.
REFUELING PATHWAY: This includes the reactor refuel cavity the fuel transfer canal, and the spent fuel pool, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel.
RUPTURE(D): The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.
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.
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.
UNISOLABLE: 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.
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.
B-3
 
Southern Nuclear Operating Company Joseph M. Farley Nuclear Plant Units 1 and 2; Edwin I. Hatch Nuclear Plant Units 1 and 2; Vogtle Electric Generating Plant Units 1 and 2; License Amendment Request for Changes to Emergency Action Level Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant Responses to Requests for Additional Information ENCLOSURE 5 VOGTLE ELECTRIC GENERATING PLANT UNITS 1AND2 EAL VERIFICATION AND VALIDATION REVISIONS
 
Vogtle EAL V&V Documentation Index V&V#                      Descriotion                        IC      Threshold RCS Modes/Temperatures -TS Reference (Table I. I-I)  EAL Basis Section 1.5 I RCS Cold Shutdown Temperature reference (200 &deg;F)      CA3            (I)
CU3            (I)
Radiation Monitor Calculations:                        RGI            (!)
RE-12839E                                              RS!            (!)
RE-12444E                                              RA!            (!)
CalcX6CNA!5 Radiation Monitor Calculations: RE-0018
* RU!          (!) (2)
RE-0021 RE-0848 ARE-0014 2                                  RE-12839 RE-!2442C RE-12444C CalcX6CNAl5 Radiation Monitor Calculation: RE-005                  CG!          (2)b RE-006                CS!          (3)b FCB L3A      FCB L3A RCSB L3A    RCSB L3A CalcX6CNAJ5                                          CB PL3A      CBPL3A ODCM/TS Reference to Site Boundary                    RGI          (2) (3) 3                                                        RSI          (2) (3)
RA!        (2) (3) (4)
SFP Level 3 & 2 Indications                            RG2            (!)
4                                                        RS2            (!)
RA2            (3)
Rad Monitor Ref(ARP):          RE-008                RA2            (2)
RE-002/003 5                                ARE-2532 A/B ARE-2533 A/B Annunciator Response Procedure (Control Room)          RA3            (!)
6
 
==Reference:==
RE-001 Annunciator Response Procedure (SFP Level)            RU2            (l)a 7 Reference Annunciator Display: ALB05 E02 (SFP Lo Level)
Radiation Monitor Information:          RE-008        RU2            (l)b RE-002 RE-003 8                                        RE-004 RE-005 RE-006 RE-0011 RVLIS Indications/Displays (RVLIS)                    CG!            (!)a CS!        (I)b (2)b 9
CA!            (!)
CalcX6CNAJ5 Containment Sump/Reactor Coolant Drain Tank            CG!            (2)b (RCDT)/Waste Holdup Tank (WHT) references (FSAR)      CS!          (3)b IO CAI            (2)b CU!            (2)b 1011412016 Page I of2
 
Vogtl e EA L V&V Doc um entation Ind ex V&V#                        Descriotioo                      IC      Threshold                  Commented [JRB1] : Renumbered V&V#'s due to Hi Concentration (?: 6%) - calculation/reference        CO i      ( l )b (2)c              deletion o f previous V 12 - Contaimnent Pressure
 
==Reference:==
 
II Calc X6CNA l5                                        CB PL4B    C B PL4B                  ~ 13 psig with Containment Closure es tab li sh ed,~ 52 psig Emergency Buses Drawing                                CA2            ( I)                with TS Containment Integrity Intact Cale X6CNA15.
Deletion was im plemented in response to RA J I I CU2          ( l )a SO I          ( l )a 12                                                          S08          (l )a SS I          ( I)
SA i          ( l )a SU I          ( I)
DC System Information                                  CU4            ( I) 13                                                          S08          ( l )b SS8            ( I)
ISFS I TS/Dose Reading Calculatio n                    E-1-I U I      ( I) 14 Calc X6CNA /5 CSFST Procedu re                                    FCB PLI A  FCB PLI A RCS B PLIB      RCSB PLI B FC B L2A    FC B L2A FCB PL2A    FCB PL2A FC B PL2B  FCB PL2B 15                                                      RCSB PL2A      RCSB PL2A CB PL2A      CB PL2A C B PL4A    CB PL4A C B PL4C. I  C B PL4C. I SO I          (I )b SSS          ( l )c 16  Seismic Monitorin g Svstem Pane l                      HU2            ( 1)              [ Commented [JRB2] : RAJ 23 revision Rad Monitor Ca lculati on: RE-4 8000                    SU3            ( I) 17 Cale X6CNA 1-1 V7 (p f! 9)
RCS Sample Ac tivity                                    SU3            (2) 18 TS 3.4. 16 RCS Operationa l Leakage                                SU4        ( I) (2) 19 TS 3.4. 13 Contai nment Spray Initiation Setpoint (2 1.5 psig)    SU7          (2)a 20 TS Table 3.3.2-1                                                                          ( Commented [JRB3] : RAI 28 revision 10/ 14/20 16 Page 2 of 2
 
V12 Page 1 of 1 t                            *                                                                                                                                                                        :;:;;;;;---
                                                                                                                                                                                                                                                                                                                                          -**-***u,... *--*.
                                                                                                                                      "::!Ji::-
                                                                                                                            ""l ..... I                                                                                                                                                                                            _........__-
u .. * * - - * * * -* * * * ..
                                                                                                                        ....~
                                                                                                                                                                                                                                                                                                                  '* ::..*.-tt..P:a:::-:-:.-:...n..~~- ..
I
                                                                                                                        ...~
:t-                                                                                                                                                                                                                                  ~
                            ,,";::;+
0
:.:.::S"::"                                                                                                                                                                                                                                                        '7
                                                                                                                                                                                                                                                          ***=r.**. . . . . V ......                                                                      -0
                                                                                                                                                                                                                                                ~~=.:.      *l t"r. rr.::!~?*            **~:::--:-..=**
l"'.-..-
              =****
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                                      , f,,.. .., ,,J) ,
I              I                    I
                                                                                      , I ,,.,
I Me"  iI
                                                                                                                      ~~~      ... "
I r-                                r*~:re        ... **
I '
I J...j I I I
I    I LI .. . i I '=***                          I
                                                                                                                                                                                                                                                                                                                                                  ~          -    .,,., ...*..
L .........f f .*p* ...... f L--J t                                                                      l:.:-::...~f                        t....
                                                                                                                                                                                                                &#xa5;
                                                                                                                                                                                                                          .. f....~ ..... *-.t~ t) ..... Ii                            M  *-
::.::.;: ~*
  ------------------------------------------------------------------------- ----------------------------------                                                              ~It.NT      VOO TL
        ~~...,,7-
                ..                                  f 1*-            '"::"                    f-..I                                              ~f.        ' "'
                                                                                                                                            ~
            "''"                                    *-*-*..                                    *-*,_....                                                                      ."*ff'              "~
                                                                                                                                                                            -l[e_ ' ,~,- f"
                                                                      *-..                                                              *-                                            ~-*                :l i    *~:::*.::.*-*
l, .. _
w                                                  *--&;                            .
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  ,_..Ji~
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                                                                                                                                                                                          '        ""                              ~
y y y .,.
y y y .,.
y y y '1'
                                                                                                                                                                                                                                                                                                          .,. y y y
                                                                                                                                                                                                                                                                                                          .,. y y y 11*  y y y y y y '1 '
y y y .,.
y y y '1'
                                                                                                                                                                                                                                                                                                                                                                      '1 ' y y y
                                                                                                                                                                                                                                                                                                                                                                      ., . y y y
                                                                                                                                                                                                                                                                                                                                                                      .,. y y y
:=r                                                      (*-
t r*. :."-
                                                                                                                                                      .. . . H y y y .,.            '1 ' y  y y                  y y y "'                      .,. y y y y y y .,.            '1' y  y y                  y y y .,.                      ., . y y y I ,*-..*-* .....                                                            ._,.            t*--
J;-t*'_;** - - .-..~ ~ t"=t_,. '! *-:::... f-"
y y y .,.            '1
* y y y                    y y y '1*                      .,. y y y
                                                                                                                                                                                                                                                                                    'r''j''j''l' ) -    '1' 'r' -
                                                                                                                                                                                                                                                                                                                  'r''r'                'r' 'j''j''1'                  '1 ' 'j''j' 'r' w::*.=-~ .J=,                                      '.>::.".::"-: ~
9                                                        ...
1 *r
                      ;;:...~--
              * .. , ,_o
                                        . .. ;~-:;*
                                                                        !P..".''        ....                      *--
                                                                                                                                                        ~ t*- f .;..:                                                                                                                                  f'"""                                                    r***H*
                ...*-                    *-ti,)                  . . .*-- *-.=..              *....                '!:"~~                      *            *:;,-.:.-
                                                                                                                                                                                                                                                                          -:i'.t~'.r:::::::
0&#xa3;.WJ.....!:.:
t  T
              ~-=r-T
                                            ~-=r-T
                                                                ~=-
                                                                                  ;~
T I
___L' ouu.:.;-:.... *cc
_c**
                                                                                                                                                                    .,,.,;.~  ... ,...
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V13 Page 1of2 provided to prevent explosive concentrations of hydrogen during battery charging.
The ventilation system shall be adequate to maintain the hydrogen concentration below 2 % in accordance with IEEE 484.
* Each room shall be provided with a shower and an eye-wash basin in case personnel come in contact with battery acid.
To prevent possible hydrogen explosion, switches and receptacles shall not be located inside battery rooms.
3.2      POWER GENERATION DESIGN BASES None 3.3      MAJOR COMPONENT DESIGN BASES 3.3.1    The Class IE de 125-V de system and all equipment are classified as Seismic Category I and Safety Class IE. The system equipment shall be capable of continuous operation between 100 and I 40 V de, except vital ac bus inverters, the reactor trip switchgear control, residual heat removal (RHR) isolation valve inverters, and the turbine-driven auxiliary feedwater pump control may be allowed to operate over a de input voltage range of I 05 to 140 V de .
The 25 kVA inverters provide power to RHR isolation valves. The breakers for these valves are located in "Trip" position during normal plant operation; they do not operate during the duty cycle time of the battery, or for any LOCA/LOOP or SBO coping scenario.
3.3.2    Requirements for Batteries 3.3.2. I Battery size shall be determined in accordance with the method indicated in IEEE 485 .
3.3.2.2  Each Class IE battery shall have sufficient capacity to independently supply the required loads for a loss-of-coolant accident, loss of offsite power, or main steam line break for a duration of 2.75 hr. Each Class IE battery shall have sufficient capacity to independently supply the required loads for a station blackout for a duration of 4 hr.
3.3.2.3  initial battery capacity shall be 25 % greater than required according to the calculation method indicated in lEEE 485 to allow for aging and extend the time interval for battery replacement as required by the battery replacement criteria of lEEE 450.
3.3.2.4  Batteries shall be sized to provide their required output at 70&deg;F.
3.3.2.5  A margin of I 0 % load growth shall be initially included in the sizing of each battery.
DC-1806                                            6                                        VER13
 
V13 Page 2 of 2 3.3.3.7  A de ammeter, de voltmeter, de overvoltage relay, ac power "on" light, and ac undervoltage relay shall be provided on each charger. The relays shall alarm in the control room.
3.3.3.8  The chargers shall be suitable for parallel operation so that each subsystem ' s redundant charger can be manually put into operation in conjunction with the normal charger to recharge a discharged battery in a shorter amount of time (if allowed by the manufacturer) .
3.3.3.9  The battery charger shall prevent the charger from becoming a load on the battery due to a power feedback during loss of ac power to the chargers.
3.3.3 .10 ac breakers shall be provided to protect the charger from internal faults and to isolate the charger from the ac source.
3.3.3 .11 de breakers shall be provided to protect the battery and charger from internal charger faults and to isolate the charger from the de system.
3.3.4    Requirements for 125-V de Metal-Enclosed Switchgear 3.3.4.1  There is one de switchgear lineup for each subsystem. It shall be connected to the battery, the normal battery charger, and the redundant battery charger associated with that tra in.
The switchgear shall feed MCCs, de distribution panels, and the inverter for the vital instrumentation system (DC-1807).
3.3.4.2  The switchgear air circuit breakers shall be equipped with direct-acting, dual-magnetic, overcurrent tripping devices providing adjustable overcurrent and short-circuit protection.
3.3.4.3  The 125-V de switchgear breakers shall serve as a means for energizing and deenergizing power sources and loads connected to the 125-V de switchgear bus. The switchgear shall also provide suitable protection for the loads during overload and short-circuit conditions.
3.3.4.4  Trains C and D shall each provide 125-V de power to an associated 480-V, 3-phase inverter for RHR iso lation valves.
3.3.4.5  The vital ac bus inverters may be allowed to operate over a de input voltage range of 105 to 140 V de. The de feeder cables shall be designed to maintain a minimum of I 05 V de during the entire battery load profile.
3.3.4.6  The RHR isolation valve inverters may be allowed to o erate over a range of I 05 to 140 V de. The de feeder cables to the RHR isolation valve inverters shall be designed to maintain a minimum of 105 V de at the RHR isolation valve inverters.
3.3.5    125-V de MCCs 3.3.5.1  One MCC shall be provided for each train A, B, and C subsystem. It shall be connected to the switchgear associated with that subsystem. The MCC shall feed motor-operated DC-1806                                              8                                        VER 13}}

Latest revision as of 19:15, 9 January 2025

Southern Nuclear Operating Company Response to RAI for License Amendment Regarding Emergency Action Level Schemes to Adopt NEI 99-01 Rev 6 and to Modify Radiation Monitors at Farley Nuclear Plant, Part 1C of 1C
ML16314A512
Person / Time
Site: Hatch, Vogtle, Farley  Southern Nuclear icon.png
Issue date: 11/03/2016
From:
Southern Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML16314A191 List:
References
NL-16-1830
Download: ML16314A512 (717)
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