^ newbrkf'bwer authority william j. cahill, jr. chief ...bases, revision 1, january 10, 1995,...

0123 Main Street9White Plains, New York 100

914-681-6840914-287-3309 (FAX)

^ NewbrkF'bwer40 Authority

William J. Cahill, Jr.

Chief Nuclear Officer

March 3, 1995IPN-95-028

U. S. Nuclear Regulatory CommissionAttn: Document Control DeskMail Station P1 -137Washington, D. C. 20555

SUBJECT:

REFERENCES:

Indian Point 3 Nuclear Power PlantDocket No. 50-286Response to NRC Request for Additional Information RegardingProposed Emergency Action Levels (TAC M89887)

1. NRC letter, N. F. Conicella to W. J. Cahill, Jr. dated December16, 1994 regarding the same subject (DSR 288538).

2. NYPA letter, W. A. Josiger to USNRC (IPN-94-030/JPN-94087) dated July 12, 1994 regarding upgraded EmergencyAction Levels.

Dear Sir:

The Authority's response to the NRC staff's recent RAI (Request for AdditionalInformation, Reference 1) regarding upgraded Emergency Action Levels (EALs) for theIndian Point 3 Nuclear Power Plant is Attachment 1.

Also attached are four associated documents which have been revised to reflect theAuthority's response to the NRC staff's questions. Attachment I1is Revision 1 of theEALs. Attachment Ill is the EAL Technical Bases Report. Attachment IV is the FissionProduct Barrier Evaluation, and Attachment V is the Indian Point 3 specific EAL guideline(PEG). These documents supersede and replace those included with Reference 2.

The Authority plans to implement these upgraded EALs after the restart of IndianPoint 3, but not before May 31, 1995.

9503070258 950303PDR ADOCK 05000286F PDR

\V

-2

No commitments are being made by the Authority in this submittal. If you haveany questions, please contact Ms. Charlene D. Faison.

Very truly yours,

Wiliam JCaill, Jr.Chief Nuclear Officer.Nuclear Generation

List of Attachments:

1. Indian Point 3 Emergency Action Levels, Response to Request for AdditionalInformation

11. Indian Point 3 Emergency Action Levels, Revision 1, Based on Proposed Responseto NRC RAls, January 10, 1995

Ill. New York EAL Upgrade Project, Indian Point 3 Emergency Action Levels, TechnicalBases, Revision 1, January 10, 1995, Operations Support Services, Inc.,OSSI-92-402A-4-1P3

IV. Fission Product Barrier Evaluation, Revision 1, New York Power Authority, IndianPoint Station Unit 3, January 10, 1995, OSSI 92-402A-2-1P3

V. EAL Upgrade Project, Plant Specific EAL Guideline, PEG, Indian Point Unit 3,Revision 1, January 10, 1995

cc: Next page

cc: All with attachments.

Regional AdministratorU.S. Nuclear Regulatory Commission475 Allendale RoadKing of Prussia, PA 19406

Resident Inspector's OfficeIndian Point Unit 3U.S. Nuclear Regulatory CommissionP.O. Box 337Buchanan, NY 10511

Mr. Nicola F. Conicella, Project ManagerProject Directorate I-1Division of Reactor Projects I/IlU.S. Nuclear Regulatory CommissionMail Stop 14B32Washington, DC 20555

RBC-D W/LTR MIDl 3/3/95... .9503070258

NOTICE

THE ATTACHED FILES ARE OFFICIALRECORDS OF THE INFORMATION &RECORDS MANAGEMENT BRANCH.THEY HAVE BEEN CHARGED TO YOUFOR A LIMITED TIME PERIOD ANDMUST BE RETURNED TO THERECORDS &ARCHIVES SERVICESSECTION, T5 C3. PLEASE DO NOTSEND DOCUMENTS CHARGED OUTTHROUGH THE MAIL. REMOVAL OFANY PAGE(S) FROM DOCUMENTFOR REPRODUCTION MUST BEREFERRED TO FILE PERSONNEL.

~- NOTICE 3 qi-2̂ .

Attachment I to IPN-95-028

Indian Point 3Emergency Action Levels

Response to NRC Request for Additional Information

New York Power AuthorityDocket No. 50-286

Indian Point 3Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION

GENERAL RAIs

Remponse to General RAI #1 (page 1)

As stated in the RAI, ICs are a subset of power plant conditions which represent a potential or actualradiological emergency. EALs are "a pre-deterined, site-specific, observable threshold for a plant ICthat places the plant in a given emergency class." When a site-specific, observable threshold (EAL) isreached, entry into its associated emergency class is required irrespective of the IC from which the EAL isderived. As stated in the RAI, ICs provide criteria that may be relevant to emergency classification basedon the users "judgment." Therefore, it follows that use of judgment may be required for those conditionsin which no "pre-determined, site-specific, observable threshold" can be defined.

Since ICs lack "site-specific, observable thresholds" for emergency classification, for those postulatedconditions in which no site specific observable threshold exists, the users judgment must be based on thegeneric definition of the associated emergency classification.

EAL Category 9.0 "Other" defines EALs in each emergency class which are based upon the user'sjudgment. Category 9.0 is used when the plant condition does not meet any of the EAL thresholds ofCategory 1.0 throughi Category 8.0 but it is determined that the plant condition meets either theemergency class definition criteria or the NUMARC/NESP-007 fission product barrier loss or potentialloss criteria. To address the concerns raised by the staff in this RAI. the bases document has been revisedto include each of the NUMARC/NESP-007 ICs. SDeCific reference to these ICs is now incornorated inthe iudgmnent EALs providing a mechanism for the user to determine how an EAL (or several diverseEALs) is related to the plant conditions of concern.

Response to General RAI #2 (vage 2)

Though not specifically stated, it is inferred that this RAI is in reference to EALs 5.2.3, 5.2.4 and 5.2.5.

For any actual or imminent release, dose projections performed in accordance with IP-lOOT, "Determiningthe Magnitude ofRelease", use of actual meteorology is specified. Therefore, implicit in the performanceOf any dose projection is the use of actual meteorology.

To address the staff s concern that classification based uvon these EALs be as the result of an "actual orimminent" release of gaseous radioactivity, the EALs have been revised to include the "Actual orImminent" terminology,

Response to General RAI #3 (naee 2)

[Par. 11NUMARC/NESP-007 does not require that the generic fission product barrier matrix be implemented on asite-specific basis. On September 22 -23, 1992 the Emergency Action Levels Implementation Workshopwas conducted by NUMARC. Specifically stated in presentations and in the workshop training materials(Section 3 page PF-39, page BF-30 and the PWR Fission Product Barrier Matrix Breakout Session GuideSection 7) attached, was the fact that the matrix format is not required. It only requires that compliancewith all combinations are documented. NUMARC/NESP-007 does not preclude the development ofEALsbased on an evaluation of fission product barrier loss/ potential loss conditions as part of the developmentprocess. The fission product barrier loss matrix as presented in NUMARC/NESP-007 was" chosen toclearly show the synergism among the EALs and support more accurate dynamic assessments." Further,NUMARC/NESP-007 states "The guidance presented here is not intended to be applied to plants as-is.


The EAL guidance is intended to give the logic for developing site-specific EALs using site-specific EALpresentation methods." The Fission Product Barrier Evaluation and the subsequent binning of the IP3fission product barrier based EAUs into categories was specifically performed to support the user's abilityto "dynamically assess how far present conditions are from escalating to the next higher emergency class."By defining logical event categories and subcategories in which to place these EALs, the ability to performa dynamic assessment is enhanced. The usability and correctness of the 1P3 method ofEAL presentationhas been demonstrated and documented in numerous dynamic simulator scenarios during EAL validationexercises.

The Fission Product Barrier Evaluation demonstrates that the 1P3 fission product barrier-based EALs aretechnically correct and meet the intent ofNUMARC/NESP-007. To address the stafrs concerns, thoseEALs which are derived from the Fission Product Barrier Evaluation have been annotated to indicate thefission Droduct barrier lossnotntial loss which they reoresent. In addition. the bases document has beenrevised to include the fission Droduct barrier lossgte~ntial loss indicators in a matrix format.

[Para. 21NUMARC/NESP-007 states "The presentation method shown for Fission Product Barriers was chosen toclearly show the synergism among the EALs and to support more accurate dynamic assessments." It doesnot 'state or imply that this method of presentation is necessary either to depict the synergism or to providethe ability for dynamic assessments. Rather, it is provided as a guide for the EAL writer to ensure that theselected presentation methodology properly reflects the desired synergistic quality and assessmentcapability. While NUMARC/NESP-007 does not define the term "dynamic assessment", it is assumedthat it means the ability to evaluate fission product barrier loss and potential loss indicators underevolving plant conditions. Unlike the NUMARC/NESP-007 matrix format, the 1P3 EAL presentationmethod places similar EALs into categories and subcategories that focus the user's attention to the specificEAL threshold that corresponds to the plant condition of concern. This provides a logical classificationand escalation path of related indicators and thus allows for rapid assessment of emergency conditionsassociated with fission product barrier loss. It is important to note that the 1P3 EAL categories andsubcategories are not simply representations or abbreviations of the NUMARC/NESP-007 ICs. Rather,each 1P3 category and associated subcategory is a pathway from broad indicators ofpotential emergencyevents to a set of specific threshold conditions that require emergency classification.

The EALs derived from the Fission Product Barrier Evaluation take into account the intended 'synergism'of the fission product barrier basis information which cannot be adequately addressed by theNUMARC/NESP-007 matrix format. An example would be a condition in which RCS leakcage intocontainment is in excess ofnormal makeup capacity (RCS potential loss) in conjunction with a secondaryside release with primary to secondary leakage in excess of technical specifications (Containment loss).Under a matrix format, this combination of conditions would require a Site Area Emergency (SAE)declaration because NUMARC/NESP-007 requires an SAE for the potential loss ofthe fuel clad or RCSwith the loss of another barrier. This is clearly not intended. NUMARCINESP-007 containment lossindicator #4 basis states that the Site Area Emergency associated with the containment loss indication isintended to be escalatory from RCS breaches associated with SG tube ruptures.

The Fission Product Barrier Evaluation does not rely on single indications as stated in the RAI. For themajority of the bounding conditions defined in the Fission Product Barrier Evaluation the indicatorssubsumned into other combinations of conditions consist of those indicators which are either:

" Completely bounded by another combination for the same indicator, or* Are a subset of another indicator.

In the case cited (>300 PtCi/cc DEI-13 1 in conjunction with primary system leakage > 75 gpm), thecombination was omitted in the Fission Product Barrier Evaluation because this condition would result inexceeding the 17 R/hr SAE EAL. The 17 R/hr SAE EAL is based on >300 iiCi/cc DEI-131I inconjunction with primary system leakage into containment.

Implementatiodn

*SiteSpecific

Analysis

Required

*MatrixFormatisnotRequired

But

YouMustDocument ComplianceWith

AllPossibleCombinations

*EALBasesWillBe

Valuable

-fortraining

-for futureEALrevisions

*Not A

SmallProject

*NeedSupport fromOther GroupsEspeciallyOperations

andEngineering

PR39NUMARI~

P-j--9

Implementation

*SiteSpecificAnalysisRequired

*Matrix

Formatisnot R

eq-uired;but

Youmustdocumentcompliancewith

all possiblecombinations

*EALbaseswillbevaluable

-Fortraining

-Forfuture

EALrevisions

*Notasmallproject

*Need

supportfromothergroups

especially,,OperationsandEngineering.

NUMARC

BF-30

PWR FissionProductBarriersSeptember 22, and 23, 1992

NRC liaison: William Reckley

Role: Provide NRC perspectiveImprove NRC knowledge of industr'y concern

Purpose of Breakout Session

Open forum to elicit comments and questions related to the methodology, discuss-lesson"learned from the pilot programs, and scenario examples with cross reference to EALs.[Quick Review-of PWR Fission Product Barriers Matrix* Barriers

* Mode Applicability

Use of EOP information, CSFs, EOP transitions, etc.* Layout of Matrix

Pilot Program Lessons

* Management acceptance of the concept.* Operations support and acceptance of the need for improvement and themethodology.

* Team approach [Task Force: (Care Group) Ops, EEP, HP, Training (LiaisonGroup), Maintenance, Security, Industrial Safety, Licensing].* Expect resistance to the Matrix. There are benefits but alternatives are possible.

Scenarios

Loss of Reactor Coolant ScenariosRCS Unidentified leakage of 5 gpm. T/S Shutdown ______'Leakage increases to 15 gpm -----Leakage increases to 65 gpm UE %)gpm EAL no longer exists (based on charging/highhead capacity)


TO address the staff's concerns, the EALs have been reised to add this combination as a gmeific fissionDroduct barrier EAL. This EAL has been added in light of the assumptions which are made in thederivation of the containment radiation monitor value associated with the fuel clad loss EAL as well asvariables in the bounding assumptions (i.e. differences in time after shutdown and coolant volumereleased).

[Pani.31(Subpara. 11Loss ofcontainment cooling will not result in a containment pressure (3.0 psig) sufficient to result in acontainment isolation. In addition, procedural requirements require the containment to be vented underthis condition to maintain pressure well below the isolation setpoint.

A faulted steam generator could result in a containment isolation signal. To address those conditions inwhich a valid containment isolation sional is not the result of a breach of the RCS. but as a result ofafaulted SG inside contairnent, classification would be made based on EAL 4.1.1 which has been modifiedto address Phase "A". Phase "B" or CVI isolation failures. regadless of initiating event.

[Subpara. 21NUMARC/NESP-007 state in the basis for containment barrier loss #I: "Conditions leading tocontainment RED path result from RCS barrier and/or Fuel Clad Barrier Loss. Thus, this EAL isprimarily a discriminator between Site Area Emergency and General Emergency representing a potentialloss of the third barrier." Therefore, entry into Containment RED path by itself is intended to result in aGeneral Emergency.

As stated in the 1P3 PEG, in order to reach containment RED path, a containment pressure of 47 psigmust be exceeded. This pressure is well in excess of the maximum pressure attained from the DBALOCA and is greater than the maximum pressure attained for all analyzed steam line breaks insidecontainment specified in the IP3 FSAR. Therefore, to attain such a containment pressure, the energysource must be as a result ofa severely degraded core (metal water reaction) in conjunction with RCSbreach or a severe ATWS condition in conjunction with RCS breach. Per NUMARC/NESP-007 IC SS2such an ATWS leads to imminent or potential loss of fuel clad.

Reference in this iustification to core cooling and heat sink RED path has been deleted from the FissionProduct Barrier Evaluation.


[Subpara. 31Per the IP3 EALs, core cooling RED only requires declaration of a Site Area Emergency. Justification#10 in the Fission Product Barrier Evaluation referenced in this RAI was in error and should have read"... and warrants declaration of a Site Area Emergency." The Fission Product Barrier Evaluation has beenrevised to correct this error and to reference the vRoer Justifications.

[Subpara. 41Per the 1P3 EALs, core cooling RED and functional restoration procedures not effective within 15 minutesis the threshold for a General Emergency. Justification #11 referenced in this RAI has been revised andthe Fission Product Barrier Evaluation has been revised to reflect the yrovr references.

tSubpara. 51The justification was not intended to infer that a loss of RCS subcooling can only occur from a loss ofRCS. Rather, that any core cooling ORANGE or RED path represents a loss of subcooling resulting froma loss ofRCS. Justification #12 has been reworded to reflect the following basis.

ORANGE path core cooling is entered when either CET> 700OF or RVLIS water level - top of fuel (REDpath ifboth conditions exist or CETs> 1200 IF). The RCS pressure corresponding to 700 IF isapproximately 3100 psig. This pressure is more than 600 psig greater than the pressurizer safety valve liftpressure and 365 psig greater than the RCS safety limit. If the RCS is intact under this condition, RCSbarrier loss is imminent. RCS inventory is never intentionally reduced to the top of fuel (39%RVLIS)under hot conditions or power operations. A reduction in RCS volume of this magnitude indicates asignificant breach of the RCS barrier since no intentional valving configuration would result in such adecrease. Any condition which results in an inventory loss of this magnitude must be attributed to anRCS breach caused by a RCS line break or unisolated primary system discharging in excess of makeupcapacity. It would be extremely poor judgment to assume that a loss of the RCS barrier has not occurredunder either of these conditions. It should be noted that vessel water level below the top of fuel isconsidered a RCS barrier loss in the BWR fission product EALs. There is no difference in themechanisms which could cause vessel level to dop below the top of fuel between BWRs and PWRs.Important to this basis is, for the purpose of emergency declaration, the potential release of fissionproducts to the environment. In the case where the fuel clad is actually or potentially breached, theassumption that the fission products would be contained, even in the absence of other RCS loss indicatorsnot immediately apparent with vessel level below the top of fuel is inappropriate. Figure 4.16 of NUREG1228 "Source Term Estimation During Response to Severe Nuclear Power Plant Accidents" shows howeach of the critical safety functions is related to fission product barrier maintenance as regards preventingradioactivity releases. Core heat removal (core cooling) along with RCS pressure control and RCS heatremoval (heat sink) are shown to be directly related to RCS boundary maintenance.

It should also be noted that NUTMARC/NESP-007 considers RED path heat sink a potential loss of RCS,yet the conditions requiring entry into tis path are based on insufficient SG level and feedwater flow.These conditions are not direct threats to RCS barrier integrity but may lead to RCS pressure conditionswhich in turn may lead to RCS barrier breach. NUMARC/NESP-007 provides no technical basis tosupport how a RED path heat sink represents a potential loss ofRCS boundary. it would appear that theRCS inventory loss conditions requiring entry into core cooling ORANGE or RED path are much moredirectly indicative of actual or potential RCS breach than is entry into RED path heat sink.

ISubpara. 6]The Fission Product Barrier Evaluation and EALs associated with the combinations referenced have berevised to include the speified combinations: Coolant activity > 300 uCi/cc 1-131 eauivalent incombination with vriMar sMste leaka> 75 gin. RCS subcooling < SI inititon setvint due to RCS

leakaf. RED oath Inteitrty or> 0.06 uCi/cc on R-l1 and R-12 due to RCS leakae

[Pama 41

OBJECTV

000-a000000000

~000go000

0-*

wo

-a001

o00

e0/

.

IPLA

DESIGN

GOALS

REACTIVTY

Figure

4.16

I Relationshipamongcriticalsafetyfunctions,maintainingfission

productbarriers,and

preventingarelease

Source:

NUREG-1210

CRITICAL JSAF

FUNCTIONS


It is still.appropriate to define, where possible, distinct EALs which are indicative of multiple barrierloss/potential loss. This minimize the time to classify while assuring multiple conditions are readilyevaluated and properly classified. Based on exhaustive operator interviews, the use of a fission productbarrier matrix format has been determined to be overly burdensome and confusing for the user resulting inmissed or inorrect classifications. Tins concern has been expressed by other licensees who haveattempted to implement NUMARC/NESP.007 fission product barrier EALs with only a matrix format.

Because of the complexity of the NUJMARC/NESP-007 fission product barrier loss/potential lossdefinition of the Site Area Emergency, some licensees have attempted to deviate from NUMARC andsimplif the fission product barrier losstpotential loss definition by removing the intended reducedweighting of the containment The reduced weighting of the containment at the SAE classification is asignificant part ofthe basis in the intended synergism between barrier loss indicators. The 1P3 FissionProduct Barrier Evaluation maintains this intended synergism of NUMARC while eliminating theinherent complexity. The 1P3 EAL format has been validated by operating crews utilizing scenarios in theplant-specific simulator to test each EAL. The results of this validation have been documented andfeedback incorporated into the EALs to futher ensure their usability.

Indian Point 3 Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION

Remnu to General RAI #4 (Date 5)

NUMARC/NESP-007 Section 3.9 states:

"Plantemergencyoperating proceduires(EOPs)are designed to maintainand/orrestorea set ofCSFswhich are listed in the orderofpriorilyofrestorationeffortsduring accidentconditions.ff.

7Tere are diverse and reduindantplantsystems to support each CSF. By monitoring theCSFsinsteadof the individualsystem component status, the impactofmultiple events isinherently addressed,e.g. the number ofoperablecomponents available to maintain thefunction.

The EOPscontaindetailedinstructionsregardingthe monitoringof thesefunctions andprovidesa schemefor classifying the significanceof the challenge to the functions. InprovidingEALs basedon these schemes, the emergencyclassificationcanflow from theEOPassessmentratherthan being basedon aseparateEAL assessment. This isdesirableas it reducesambiguity andreducesthe time necessaryto classify the event."

As stated by NUMARC, each CSF is supported by diverse and redundant plant systems. The entryconditions for CSFSTs are also supported by diverse and redundant instrumentation. Containment REDpath is not a single indicator but a defined, measurable and operationally significant condition which isknown to be indicative of multiple fission product barrier losses. The 1P3 EAL scheme does not relysolely on this condition to determine when a general emergency due to the loss of fission product barriersmust be declared. Nor does it preclude the declaration of a general emergency based on other fissionproduct barrier loss EALs which may or may not manifest themselves under a given condition. The EP3EAL scheme does require classification of aGeneral Emergency because, in and of itself, this conditionrepresents a loss of the fuel clad, RCS barriers and a potential loss of containment barrier.

Res~onse to General RAI #5 (pat 5)

Refer to Response to General RAI #3 [Para. 31 [Subpara. 5]

Indian Point 3Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITONAL INFORMATION

SPECIFC RAY,

Resnonue to Specific RAI #1 (nauc 5)A.EAL # 5.1.1has been revised to reference Rerformance of an assessment of the release. The EAL has alsobeen revised to include criteria reauiring declaration if the assessment is not accomulished within 60mnues-B.R-19 is a liquid effluent process monitor. This release path only applies to NUMARC/NESP-007 ICAUl.1 and AAIl. The upper range ofR-14 is1E6 cpm. The value associated with the Alert criteria isin excess of this instruments range and is therefore indicated as N/A. Subsequent to the original NRCsubmittal, IP3 replaced Monitor R-024 with a new Backup Hi-Range Vent Monitor. The Bottom range ofthis monitor is 140 Ci/sec. This value is greater than the SAE trigger threshold is therefore indicated asN/A. A value of 360 Ci/ sec as obtained from this monitor will indicate a GE.

Steam dump and main steam safety valve monitors are not specified since release from these paths aredependent upon system flow rate which in turn is dependent upon the number ofvalves open and the RCSpressure over the duration of the release. Due to the wide range of release rates possible for a givenmonitor reading, no single trigger value would be appropriate. Releases from these paths are classifiedunder the subcategory 5.2 EALs.

The IP3 PEG has been revised to lDrovide these iustifications.

Response to Specific RAY #2 (DaLpe 6)0

EAL # 5.1.2 has been revised to reference verformance ofan assessment ofthe release. The EAL has alsobeen revised to include criteria reauiring declaration if the assessment is not accomplished within 15minutes.

Response to Specific RAI #3 (Daee 7)

As stated in the basis for IC AA2 in the 'P3 PEG: "There is no indication that water level in the spent fuelpool or refueling cavity has dropped to the level of the fuel other than by visual observation. Since AA2.2addresses visual observation of fuel uncovery, EAL AA2.3 is unnecessary. Since there is no levelindicating system in the fuel transfer canal, visual observation of loss of water level would also berequired, EAL AA2.4 is unnecessary." Therefore, EAL 2.4.3 addresses the concerns of these exampleEALs.


Response to Snecific RAI #4 (name 7)0

The conditional "and" criteria was added to be consistent with the IC from which this EAL was derived aswell as with the technical bases. As stated in NUMARCINESP-007: "It is this impaired ability to operatethe plant that results in the actual or potential degradation of the level of safety of the plant The cause ormagnitude of the increase in radiation levels is not a concern of this IC." The NUMARC AA3 IC states"...radiation levels within the fakcility that impedes operation of systems required to maintain..." Thereforethe intent if the IC is not to declare simply upon the existence of such as radiation level, rather, to declareifaccess is impeded. If access to the area is not required then access is not impeded. The 1P3 PEG hasbeen revised to reflect the EAL and bases wordiniz.

Response to Specific RAI #5 (pate 8)

EAL N5.1.3 has been revised to reference uerfor-mance ofan assessment of the release. The EAL has alsobeen revised to include criteria reauirina declaration if the assessment is not accomplished within 15minutes.

The source terms utilized to determine the values in Table 5.1 are those utilized in the 1P3 dose projectionprocedure IP-lO0l, "Determining the Magnitude ofRelease". The IP-1001 dose assessment methodologyuses dose conversion factors derived from WASH-1400 inventories and RG 1.4 design base fractions.Annual average (ODCM) meteorology was applied in determining the effluent monitor values.

Response to Specific RAI #6 (name 9)

EAL # 5.1.4 has been revised to reference wurformance ofan assessment of the release. The EAL has alsobeen revised to include criteria rouiring declaration if the assessment is not accomplished within 15minutes.

Table 5.2 has been revised to quantift doses in rem. The term "TEDERate has been changed to"External ENxoosr Rate. The term "CDE Thyroid Rate has been changed to Thyroid Exposure Rate(for 1 hr. of inhalation)".

Resnonse to Specific RAI #7 (nate 10)

Refer to Response to General RAI #3 [Para. 31 [Subpara. 51 for justification of use of ORANGE or REDpath core cooling as a RCS loss indicator. Use of this CSF as a RCS loss indicator is not a conservatism,but rather one of multiple indications of potential Fuel Clad and RCS barrier loss available to the user.While this CSF indicator by itself requires declaration of a Site Area Emergency, it is not inconsistentwith NUMARC. For example, NUMARC/NESP-007 specifies RED path Heat Sink as both a potentialloss of fuel clad and RCS barriers. Even though NUMARC/NESP-007 does not provide a basis for howRED path heat sink relates to RCS barrier potential loss, none the less, a Site Area Emergency is requiredbased on this singular CSF.

Resnoms to Specific RAI #7 (name 11 -this RAI is also identified as #7)

In the case cited (>300 ILCi/cc DEI-13 1 in conjunction with primary system leakage > 75 gpm), thecombination was originally omitted in the Fission Product Barrier Evaluation because this conditionwould result in exceeding the 17 R/hr SAE EAL (refer to response to general RAI #3, para. 3, subpara 3).The 17 R/hr SAE EAL was based on >300 ;&Ci/cc DEI-13 1 in conjunction with primary system leakage


into containment However, this EM. has been added in light of the assumptions which are made in thederivation ofthe containment radiation monitor value rssociated with the fuel clad loss EM. as well asvariables in the bounding assumptions (i.e. differences in time after shutdown and coolant volumereleased).

The Fission Product Barrier Evaluation and EALs associated with the combinations referenced have beenrevised to include the s&Oecqid combinations: Coolant activity > 300 uCi/cc; 1-131 eouivalent incombination with DriMa sMtm leakage > 75 gnm. RCS subcooling < SI initiation setooint due to RCSleakaae. RED Rath Intgritv. or> 0.06 uCi/cc on R-l1 and R-12 due to RCS leakage.

Regarding the combination ofa primary to secondary leak in excess of the RCS barrier loss threshold (75gpm) with unisolable release of secondary side to atmosphere and failed fuel (300 lLCi/cc: DEI-13 1), thiscondition would be classified as a General Emergency as cited in the RAI.

EAL 4.2.2 states:

"Unisolated faulted (outside VC) ruptured steam generatorANDAny indicators of fuel clad damage, Table 4.2"

The technical bases of this EM. states:

"This EM. is intended to address the full spectrum of Steam Generator (SG) tube rupture eventsin conjunction with a loss of containment due to a significant secondary line break with actual orpotential loss of the fuel clad integrity. This EM. addresses ruptured SG(s) with an unisolablesecondary line break corresponding to the loss of 2 of 3 fission product barriers (RCS barrier andcontainment barrier) with the actual or potential loss of the third (fuiel cladding). This allows thedirect release of radioactive fission and activation products to the environment. Resultant offsitedose rates are a function of many variables. Examples include: coolant activity, actual leak rate,SG carry over, iodine partitioning&and meteorology.

The indications utilized should be consistent with the diagnostic activities of the emergencyoperating procedures (EOPs), if available. This should include indication of reduction in primarycoolant inventory, increased secondary radiation levels, and an uncontrolled or completedepressurization of the ruptured SG. Secondary radiation increases should be observed viaradiation monitoring of condenser air ejector discharge, SG blowdown, main steam, and/or SGsampling system. Determination of the "uncontrolled" depressurization of the ruptured SGshould be based on indication that the pressure decrease in the ruptured steam generator is not afunction of operator action. This should prevent declaration based on a depressurization thatresults from an EOP induced cooldown of the RCS that does not involve the prolonged release ofcontaminated secondary coolant from the affected SG to the environment. This EM.encompasses steam breaks, feed breaks, and stuck open safety or relief valves.

Table 4.2 presents fuiel clad loss and potential loss indicators:

" ORANGE path in F-O.2, Core Cooling: Refer to EM. #1.1.1 basis* RED path in F-0.3, Heat Sink: Refer to EM. #1.2.1 basis" Coolant activity>300 1LCi/cc of1-13 1: Refer to EM. #2.1.2 basis" Containment rad monitor reading > 17 R/hr: Refer to EM. #2.2.2 basis

This condition represents a loss ofboth RCS and primary containment with the loss or potentialloss offuiel cladding and thus warrants declaration of a General Emergency."


Also, EAL 4.1.6 states:

"Either:Any Phase "A*or Phase "B" or containment ventilation isolation valve(s) not closed whenrequired following confirmned LOCAORInability to isolate any primary system discharging outside containment

ANDRadiological release to the environment exists as a resultANDAny indicators of fuel clad damage, Table 4.2"


The technical bases of thisEALste:

"This EAL indicates loss ofboth RCS and containment with loss or potential loss of the fuelcladding and therefore warrants declaration of a General Emergency.

Failure of Phase "A7 or Phase "B" or CVI v'alves to isolate is intended to address incompletecontainment isolation that allows direct release to the environment. It represents a loss of boththe RCS and containment barrier.

The criterion "Inability to isolate any primary system discharging outside containment" addressesany breach of the RCS and containment which is not protected by the Phase "A", Phase "B" orCVI systems or which results from an interfacing system LOCA (not addressed by NUMARC).No leakage threshold is specified since leaks outside containment, particularly under dynamicconditions, are difficult to quantify and may manifest themselves with diverse symptoms.Symptoms of a primary system discharging outside containment may be indicated via massbalance, decreasing RCS inventory without corresponding containment response, or areatemperatures and radiation levels outside containment. It is for this reason that Senior WatchSupervisor/Emergency Director judgment is intended to be used in evaluating this criteria.However, it is intended that the magnitude of the leak associated with this EAL be consistentwith the RCS barrier loss threshold of 75 gpm or greater.

Table 4.2 presents fuel clad loss and potential loss indicators:

" ORANGE path in F-0.2, Core Cooling: Refer to EAL #1.1.1 basis" RED path in F-0.3, Heat Sink: Refer to EAL #1.2.1 basis" Coolant activity > 300 jiCi/cc of 1-13 1: Refer to EAL #2.1.2 basis* Containment rad monitor reading > 17 R/hr: Refer to EAL #2.2.2 basis"

The condition described in the RAI would be classifiable under either of these EALs.

Resnouiue to Specific RA! NO (pan 13)

Phase "A", Phase "B" and Containment Ventilation Isolation (CVI) valves are those valves associatedwith the Phase "A", Phase "B" and CVI isolation logic. Phase "A", Phase "B" and CVI are protectivesubsystems of the Containment Isolation System (CIS) designed to close containment isolation valves inthose systems which either come into direct contact with primary pressure or the containment atmosphereand penetrate the containment barrier. These valves are designed to close under conditions which areindicative of a LOCA (any automatic SI signal - Phase A &CVI or requiring containment spray - Phase B&CVI). Failure of one or more of these valves to close following a confirmed LOCA does not by itselfprovide a pathway outside containment As long as one valve in the line is closed, or if both valves fail toclose but no downstream pathway exists, classification under this EAL would not be required. Thecriterion "AND Radiological pathway to the environment exists" provides this discriminator. There is nointerface between the Phase "A", Phase "B" and CVI systems but each is comprised of diverse systemswhich provide the containment isolation function under LOCA conditions. The determination of theexistence ofa LOCA is consistent with the diagnostic activities specified in E-0 'Reactor Trip or SafetyInjection'.

The criterion "Inability to isolate any primary system discharging outside containment" addresses anybreach of the RCS and containment which is not protected by the Phase "A", Phase "B" or CVI systems orwhich results from an interfacing system LOCA (not addressed by NUMARC). No leakage threshold isspecified since leaks outside containment, particularly under dynamic conditions, are difficult to quantifyand may manifest themselves with diverse symptoms. Symptoms ofa primary system discharging outside

-12-

Indin Point 3Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION

containment may be indicated via mass balance, decreasing RCS inventory without correspondingcontainment response, or area temperatures and radiation levels outside containment. It is for this reasonthat Senior Watch Supervisor/Emergency Director Judgment is intended to be used in evaluating thiscriteia. However, it is intended that the magnitude of the leak associated with this EAL be consistentwith the RCS barrier loss threshold of 75 gpm or greater.

The technical bases for EALs 4.1.3 and 4.1.6 have been revised to add the clarification that it is intendedthat the magnfitude of the leak associated with this EAL be consistent with the RCS barrier loss thresholdof 75 anm or maer.

Resnonse to Specific RAI #9 (pate 13)

As described in Response to General RAI #3 [Pana. 31 [Subpara. 5], RCS inventory is never intentionallyreduced to the top of fuel (39% RVLIS) under hot conditions or power operations. A reduction in RCSvolum of this magnitude indicates a significant breach of the RCS barrier since no intentional valvingconfiguration would result in such a decrease. Any condition which results in an inventory loss of thismagnitude must be attributed to a RCS breach caused by a RCS line break or unisolated primary systemdischarging in excess of makeup capacity. It would be extremely poor judgment to assume that a loss ofthe RCS barrier has not occurred under this condition. Important to this basis is, for the purpose ofemergency declaration, the potential release of fission products to the environment. in the case where thefuel clad is actually or potentially breached, the assumption that the fission products would be contained,even in the absence ofother RCS loss indicators, with vessel level below the top of fuel is inappropriate.As stated above, it requires a significant RCS inventory loss to attain this level. Therefore, consideringvessel level below the top of fuiel a loss ofRCS is not conservative, but appropriate.

It should also be noted that vessel water level below the top of fuel is considered a RCS barrier loss in theBWR fission product barrier EALs. There is no difference in the mechanisms which could cause vessellevel to drop below the top of fuel between BWRs and PWRs.

There is also a conflict within NUMARC/NESP-007 regarding vessel water level. As stated in the RAI,NUMARC/NESP-007 would only require declaration of an Alert due to vessel level below the top of fuelbased on fission product barrier loss. The fission product barrier loss EALs only apply under poweroperations and hot condition. Yet system malfunction IC SS5 requires declaration of a Site AreaEmergency for vessel level resulting in core uncovery when in cold shutdown or refueling modes. Thiswould mean that without other RCS loss indicators, if the vessel level dropped'to below the fuel under hotconditions, the emergency would have to be upgraded to a Site Area Emergency if the plant achieved coldconditions.

Response to Specific RAI #10 (Daze 14)

Refer to Response to General RAI #3 [Pama 31 [Subpara. 21. It would be inappropriate not to declare aGeneral Emergency based on a valid indication ofcontainment pressure in excess of 47 psig resultingfrom a loss of reactor coolant, regardless of the availability of other futel clad and RCS barrier loss EALs.It is understood that ifother applicable fuiel clad and RCS barrier loss indicators are available, they wouldserve to confirm their respective barrier losses. But NUMARC/NESP.007 does not require confirmationby multiple barrier loss indicators for a single barrier. That is, any one valid barrier loss indicator issufficient to consider that barrier lost The basis supporting declaration of a General Emergency uponentry into RED path containment is that it is indicative of loss of both fuel clad and RCS with potentialloss of containment.

The only source of significant hydrogen concentration in containment is severe fuel damage resultingfrom metal-water reaction and subsequent discharge into the containment atmosphere. A containment

-13-


hydrogen concentration of 4% corresponds to a range of 30% - 40%/ metal-water reaction (Attachment 9to EP- 1028 "Core Damage Assessment") and is well into the possible uncoolable core geometry region(Figure B-10 NUREGIBR-O 150, Vol. 1,Rev. 2). Failure to declare a General Emergency, based on avalid indication, under these conditions is inappropriate.

-14-


Reseu to Specific RAI #11 (Dupe 15)

The actuation setpoint for the Phase "B" isolation is 23 psig. This pressure is significantly high toindicate a significant loss OfCoolant accident for containment pressure increases resulting from a loss ofcoolant accident. EAL 4.1.4 has been revised to specify a confirmed Rase "B" isolation signal as a resultof a loss -ofreactor coolant to discriminate from a severe faulting of SGs inside containment.

Table 4.1 identifies fuel clad loss indicators for use in combination w~ith the RCS loss and the containment91MI indicator ("Conffimed phase "B- isolation signal due to LOCA with less than minimum

containment cooling safeguards equipment operating"). Table 4.2 includes fuel clad loss and potentialloss indicators for use in combination with RCS loss and containment loss indicators. RED Rg1L corcoolinz has been added to the fuel clad loss inicator list consistent with the fission Droduct barrierMatrix The term "fuel clad damage indicators " was used to represent both fuel clad loss and potentialloss indictors. The term 'fuel clad loss indicators" was used to represent fuel clad loss indicators only.

Rawaose to Snecific lAX #12 (Daee 16)

Refer to Response to General RAI #3 [Para. 31 [Subpara. 51 for justification of use ofRED path corecooling as a Fuel Clad and RCS loss indicator.

NUMARC/NESP-007 Section 3.9 states:

"Plantemergency operatingprocedures (EO~s)aredesignedto maintainand/orrestorea set ofCSFswhich are listedin the orderofpriorityofrestorationeffortsduringaccidentconditions.".

There are diverseand redundantplant systems to supporteach CSF By monitoringtheCSFs insteadofthe individualsystem component status, the impact ofmultiple events isinherentlyaddressed,e.g. the number of operablecomponents available to maintainthefunction.

The EO~scontain detailedinstructionsregardingthe monitoringofthesefunctions andprovidesa scheme for classifyingthe significanceofthe challenge to thefunctions. InprovidingEA4 s basedon these schemes, the emergency classiicationcanflow from theEOP assessment ratherthan being basedon a separateFAL assessment. This isdesirableas itreducesambiguity andreduces the time necessary to classify the event."

As stated by NUMARC, each CSF is supported by diverse and redundant plant systems. The entryconditions for CSFSTs are also supported by diverse and redundant instrumentation. Core Cooling REDpath is not a single indicator but a defined, measurable and operationally significant condition which isknown to be indicative of multiple fission product barrier losses. The 1P3 EAL scheme does not relysolely on this condition to determine when a General Emergency due to the loss of fission product barriersmust be declared. Nor does it preclude the declaration ofa General Emergency based on other fissionproduct barrier loss EALs which may or may not manifest themselves under a given condition. The IP3EAL scheme does require classification of a General Emergency because, in and of itself, this conditionrepresents a los of the fuel clad RCS barriers and a potential loss of containment barrier.

Resnonse to Smeific AI #13 (Date17)

The conditions defined by this EAL were identified as other site specific indications ofcontainmentbarrier failure that unambiguously indicate loss or potential loss of containment barrier.

Indiain Point 3Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION

Both doors open on VC airlock is a clear breach of the containment barrier. While these doors arenormally interlocked to Preclude this condition, an interlock failure is possible. Since lP3 Tech. Spec.allows this condition for up to 4 hrs., the 4 hr. criteria was specified. This is consistent with theNUMARC/NESP-007 philosophy not to declare events within the Tech. Spec. allowed envelope.

Inability to close containment pressure relief or purge valves which results in a radiological release path tothe environment for >4 hrs. was also identified as a clear breach of containment barrier. Thecontainent pressure relief and purge valves may be periodically opened under routine plant operationsand therefore a condition in which these valves cannot be closed, even though no automatic isolatingevent exists (LOCA) is possible. Since 1P3 Tech. Spec. allows this condition for up to 4 hrs., the 4 hr.criteria was specified. This is consistent with the NUMARC/NESP-007 philosophy not to declare eventswithin the Tech. Spec. allowed envelope.

Response to Specific RAI #14 (oaae 18)

EAL 2.2.1 has been revised to indicate >0.06 u~i/cc on R-1lI and R-12 due to RCS leakage. Reference tocoolant activity has been deleted. The technical bases has been revised to sumurt this changfe.

This EAL is included under the "Reactor Fuel" category and "Containment Radiation" sub category sincethe indication is based on containment radiation monitor readings. These readings are most closelyassociated with the reactor fuiel. The 1P3 EAL presentation method places similar EALs into categoriesand subcategories that focus the user's attention to the specific EAL threshold that corresponds to theplant condition of concern. This provides a logical classification and escalation path of related indicatorsand thus allows for rapid assessment of emergency conditions associated with fission product barrier loss.It is important to note that the IP3 EAL categories and subcategories are not simply representations orabbreviations of the NUMARC/NESP-007 ICs. Rather, each IP3 category and associated subcategory is apathway from broad indicators of potential emergency events to a set of specific threshold conditions thatrequire emergency classification. Those EALs which are derived from the Fission Product BarrierEvaluation have been annotae to indicate the fission nroduct barrier loss/notential loss which te

Response to Smciic RAI #15 (naff 18)A.This EAL is included under the "Reactor Fuel" category and "Containment Radiation" sub category sincethe indication is based on containment radiation monitor readings. These readings are most closelyassociated with the reactor fuel. The 1P3 EAL presentation method places similar EALs into categoriesand subcategories that focus the user's attention to the specific EAL threshold that corresponds to theplant condition of concern. This provides a logical classification and escalation path of related indicatorsand thus allows for rapid assessment of emergency conditions associated with fission product barrier loss.It is important to note that the 1P3 EAL categories and subcategories are not simply representations orabbreviations of the NUMARC/NESP-007 ICs. Rather, each 1P3 category and associated subcategory is apathway from broad indicators ofpotential emergency events to a set of specific threshold conditions thatrequire emergency classification. Those EALs which are derived from the Fission Product BarrierEvaluation have been annotatedt to indicate the fission product barrier losvtential loss which thgy

I.NUMARC/NESP-007 does not specify that multiple fission product barrier loss indicators must be presentto consder that barrier lost. The logic term used between each fission product barrier losstpotential lossindicator in Table 4 ofNUMARC/NESP-007 is "OR". This means that any one indicator is sufficient to

Indian Point 3 Emergency Action LevelsRESPONSE TO REQUESTS FOR ADDITIONAL INFORMATI0N

consider the barrier lost or potentially losL Furthermore, NUMARC/NESP-007 does not state that theSam indicator should not be used to indicate the loss cf more than one fission product barrier.

NUMARC/NESP-007 also states in part:

"5. Slgn~fcant Rwiioacdve Inventor in Containment"

"The (site-specifc)readingisa value which indicatessigrnficantfuel damage well in excess ofthe E4LS associatedwith both loss ofFuelCladandloss ofRCS barriers.As statedin Section3.8, amajor releaseofradioactivityrequiringoffsie protective actionsfromn core damage is notpossible unless a majorfailureoffuel claddingallows radioactivematerialto be releasedfromthe core into the reactorcoolant. Regardlessofwhether containmentis challenged, this amountofactivityin containment, if released, couldhave such severe consequences that it is prudentto&tW this asapotentialssofcontainment, such thata GeneralEmergency declarationisl-rtL

It is also important to note that it is not expected that emergency classification would be based oncontainment radiation alone. Provided that other indicators are available, classification would beconfirmed by those redundant indicators. But, in the event of a severe accident, many of the otherindicators of multiple fission product barrier loss may not be available. Therefore, it would be appropriateto rely on this single indicator since it is indicative ofmultiple fission product barrier loss/potential loss.

Refer to the attached site specific analysis used to determine the containment radiation monitor setpoints.

Resueue to Specific RAI 016 (ne 19)A.EAL 8.4.1 has been revised to state an "Earthguake felt in~lant based 1=on a consensus of Control RoomOnerator on dut AND.'

B.NUMARCJNESP-007 quotes the followving paragraph from the referenced EPRI guidance defining a "feltearthquake" as:

"An earthquake of sufficient intensity such that: (a) the inventory ground motion is fet at thenuclear plant site and recognized as anearthquake based on a consensus of Control Room operators on duty at the time, and (b) forplants with operable seismic intuettothe seismic switches of the plant are activated. Formost plants with seismic instrumentation , the seismic switches are set at an acceleration of about0.01 g."

The referenced EPRI guidance clearly states that the "felt" earthquake requires both conditions of theearthquake being feclt and activation of seismic switches..

Response to Specific RAI 017 (name 20)

EA. 8.4.3 has been deleted. The exmple EM. from which it was derived. HUI-3 and its generic basesDrovides no 12ecific imidance for declaration beyond that which the IC provides. Therefore this EML hasbeen subsuimed into the "Other" catezorv EA. 9.1.1. The section 8.4 EALs have been renumbered

-17-

Reson to Specific RAI #18 (naze 21)


EAL 8.2.1 has begn risad to ame -Confirmed fire in ar contiguous to any plant area. Table 8.2 not...".

Remnonue to Snecific RAI #19 (page 21)A., B.EAL 8.1.1 has been risd to include an security event which rEuresents a ootential degrdation in thelevel of aft of the Wlagt

EAL B.1.2 has been revised to include any security event which rEuresents an actual substantialdegradatiOn of the level of saft of the RuLat

EM. 8. 1.3 has been revised to include an security event which Mrerescnts actual or likely failures of plantsystems needed to DRotc the public.

C.EML 8.1.1 has been revised to state "-.but outside vlant vital areas, Table 8.2".

Response to Snecific RAI #20 (pan 22)

Toxic or fammable gases do not in themselves pose any threat to the safe operation of the plant but maypreclude access to areas necessary for safe operation of the plant. Therefore the concern of this EM. areconcentrations which are either life threatening or preclude access to areas needed for safe plantoperation. No specific thresholds have been defined since specific thresholds are dependent upon the typeof toxic or flammable gas involved as well as the amount and type ofpersonal protective equipmentavailable to those individuals requiring access. Therefore, the determination as to whether concentrationsare sufficient to be ife threatening or preclude access to areas required for safe operation is left to thejudgment of the user. Where specific criteria are available to the user it is expected that criteria would beconsidered in this evaluation.

Response to Specific RAI #21 (Dare 23)

EM. 7.2.3 has been revised to sncif entr into ONOP-FP-IA. "Safe Shutdown From Outside the ControlRoom" which urovides xruidance for control room evacuation.

-18-


Response to Sneclfl RA! #22 (pn 23)

Revisd EAL 7.2.5 SWat MPARI Control =Wno be established E ONOP-FP- IA. 'Safe Shutdwn From

Beasne to Specific RA! 023 (page 24)

The statement 'At least (site-specific) emergency generators are supplying power to emergency buses"serves no purpose. This HAL is concerned only with the loss of off-site AC power capability. If one of theemergency diesels is not supplying its emergency bus under hot conditions then an Alert would bedeclared based on EAL 6.1.3 (SM). NUMARC provides no criteria for the condition in which offsite ACPower capability Is lost and one emergency diesel generator is not supplying its emergency bus under coldconditions. If neither emergency diesels are supplying their emergency busses, either an Alert would bedeclared based on EAL 6.1.2 or a SAE based on EAL 6.1.4, depending on plant operating mode.

Remponse to Specific RA! #24 (nampe 24)

EALs 7.3.1 and 7.3.3 have been revised to add the words "saftssem annunciators or indications..."

Besnonse to Specific RAI #25 (Dame 25)

The term "unplanned" is not necessary. There would never be a plantned loss off all onsite or offsitecommunications capability. For a planned outage ofcommunicat ions equipment, alternatecommunications systems would always be established.

Response to Specific RA! #26 (Daae 26)

Both DC buses would never be de-energized for any planned activity unless the reactor was defueled.

Response to Specific RA! #27 (naye 27)A.EAL 6.1.2 mode Wfuicabilily has been revised to include the defuiel mode.

B.The statement *Failure of (site-specific) emergency generators to supply power to emergency buses" servesno purpose. This EAL is concerned only with the loss of all AC emergency bus power capability for> 15minutes under hot conditions. By definition, if the emergency busses are de energized for> 15 nun.,neither the AC transformers nor emergency diesel generators were successful in supplying power toemergency busses.

Besnonse to Sneific RA! #28 (Daze 28)

EAL 1. 1.1 and it's associated technical bases have been revised to be consistent with theNUMARCNESP-M7 criteria,

-19-

Responhe to Specific RA! #29 (Dawe 29)


The 1P3 Technical Specifications do not specify required functions to maintain cold shutdown. EAL 7.2.4is derived from IC SA3 which states: "Inability to Maintain Plant in Cold Shutdown." The anticipatorycriteria is provided in the use of the term "cannot be maintained." The definition section of the TechnicalBases Document defines the term as follows: "The value of the identified parameter(s) is not able to bekept above /below specified limits. This determination includes making an evaluation that considers bothcurrent and future system performance in relation to the current value and trend of the parameter(s).Neither implies that the parameter must actually exceed the limit before the action is taken nor that theaction must be taken before the limit is reached." NUMARC/NESP-007 "Questions and Answers"published in June 1993 defines the term 'function' as : "The action which a system subsystem orcomponent is designed to perform." The evaluation of both current and future system performance(function) is inherent in this definition of "cannot be maintained."

Response to Snecafic RAI #30 (Daze 30)

EAL 7.3.3 has been revised to include the term "significant transient".

Response to Specific RAI #31 (page 31)

The proper EAL reference isEAL 6.1.3. This was identified as a typographical error. EAL 6.1.3 hasbeen revised to state "Available safeguard bus AC power reduced to only one of the followiniz....

Response to Specific RA! #32 (namp 32)

The concern of NUMARC IC SS1I and this EAL is the loss of ability to provide AC power to the0safeguards busses and their vital loads. A condition can exist where the supply transformers and/oremergency diesel generators are available but a fault on the bus precludes powering vital loads. Thereforeit is more appropriate and inclusive to define the EAL by the inability to power the safeguards buses ratherthan the loss of the power sources.

Response to Sneific RA! #33 (naze 32)

EAL 1.1.2 and it's associated technical bases have been revised to be consistent with theMUMARC/NESP-007 criteria,

Response to Specific RAI.#34 (Dam 33)

1P3 Technical Specifications Section 1.2 defines hot shutdown as: Reactivity within the limits of Figure3. 10-1 andTavg> 200 OFand< 353 VF. As stated in theRA, EAL 1.1.2 addresses loss ofreactivitycontrol. The NUMARC/NESP-007 basis for SS4 also states that the EAL is intended addresses loss offunictions, including ultimate heat sink. No reference to core cooling is made. However, EAL 1.2.1 andEAL 3.1.3 provide for the declaration of a Site Area Emergency under conditions which loss of functionsthreaten core cooling~ It is also important to differentiate between function and operability of componentsor equipment which support a function. NUMARC1NESP-007 "Questions and Answers" published inJune 1993 defines 'function' as: "The action which a system, subsystem or component is designed toperform. Safety functions, as applied to PWRs are reactivity control, RCS inventory control andsecondary heat removal." NUMARC/NESP-007 Section 3.9 states "There are diverse and redundant plantsystems to support each CSF. By monitoring the CSFs instead of the individual system component status,the impact of multiple events is inherently addressed, e.g., the number ofoperable components available


to maintain the function."' Since it would be impossible to define all possible losses of system componentoperability une which loss of funrction may occur, consistent with Section 3.9 ofNUMARC1NESP-007,the loss of function is defined by CSF status. For secondary heat removal, that CSF is RED path heatsink The Techni-cal base document has been revised to reflect that EALs 1.1.2. 1.2.1 and 3.1.3 alsoserve to su~nrt IC SS4.

Respnse to Specific RAI #35 (Date 33)

The EAL does not imply that the reactor vessel head can be removed while in hot condition. Since thisconfiguration would never occur under hot conditions, that portion of the EAL based on visual observationwould not apply or be evaluated.

As stated in the RAI, oneof theNUMARC ICs from which EAL 3.1.3 is derived is NUMARC IC SS5:"Loss ofWater Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel."' Thereare numerous conditions which can lead to a loss ofRCS inventory to the extent resulting in coreuncovery while in cold shutdown or refuel modes. The one addressed in the generic bases for PWRs is"sequences such as prolonged boiling following loss of decay heat removal." Loss of inventory can alsooccur as a result of drain down events. The concern of this IC and EAL is uncovery of the fuel, regardlessof the cause. Therefore the criteria regarding loss of decay heat removal serves no function. The EALwording "RVLIS cannot be maintained.." provides for the anticipatory criteria.

The mode applicability was expanded to include the inability to maintain RVLIS above top of fuelconsistent with use ofRVLIS level as a fuel clad barrier potential loss and RCS barrier loss indicator.Refer to Response to Specific RAT #10.

The RAI makes reference to local high power densities which can "uncover" fuel and cause fuel damagewithout loss of RCS inventory. While this may be true, this EAL makes no reference to local fueluncover)'. Rather, this EAL addresses loss of inventory indicated by RVLIS. Local uncover would not beobservable by RVLIS. Refer to Response to Specific RAI #10 for justification for use of RVLIS indicationas aloss ofRCS.

Response to Specific RAI #36 (nate 34)A.The wording "is not likely" has been added to EAL 6.1.5 regarding restoration of power.

The wordinz has been revised to reflect the wording: "Actual or immrinent entry into ORANGE or REDvath on F-0.2 Core Coolimg"

B.The concern of NUMARC IC SGT and this EAL is the loss of ability to provide AC power to thesafeguards buses and their vital loads. A condition can exist where the supply transformers and/oremergency diesl generators are available but a fault on the bus precludes powering vital loads. Thereforeit is more appropriate and inclusive to define the EAL by the inability to power the safeguards buses ratherthan the loss of the power sources

Response to Specific RAI #37 (Dae 36)

EAL 1.1.3 has been revised to include the core cooline OR heat removal loixic by inclusion ofRED oathcore cooling in combination with RED path Subcriticality. EAL 1.3.2 has been subsumed into theSubcrit sub-category sinc this is the common condition in combination with either core cooling o

-21-


Resnoeme to Snecifi RI #38 (pane 36)

Table 7.3 has been revised to inclde Panel "FBF".

-22-

Attachment 11to IPN-95-028

Indian Point 3 Emergency Action Levels, Revision 1,Based on Proposed Response to NRC RAls,

January 10, 1995

New York Power AuthorityDocket No. 50-286

ISincethisdocumentwasprepared,theSrity

haschangedtheJobTitle"Shift

Supervisor"to"ShiftManager."

ThedutiesandresponsibilitiesoftheShift

ManagerarethesameasthosepreviouslyassignedtotheShiftSupervisor.

This

documentwillberevised,priortoimplementation,

toreflectthischange.

IndianPoint 3EmergencyActionLevels

Rev.1

BasedonProposedResponsestoNRCRAIs

Category1.0

CSFSTStatus

Category2.0

ReactorFuel

Category3.0

ReactorCoolant System

Category4.0

Containment

Category5.0

Radioactivity

Release

Category6.0

ElectricalFailures

Category7.0

Equipment Failures

Category8.0

Hazards

Category9.0.

Other

2/16/95

1.0

CSFST

Status

1.1

Subcriticality

1.1.1

Alert

(SA2]

Any

failure

ofan

automatic

trip

signal

to

reduce

power

range<

5%

AND

Manual

trip

issuccessful

Power

Operations,

Hot

Shutdown

1.1.2

site

Area

Emergency

[SS2

]

RED

path

inF-0.1

SUBCRITICALITY

AND

Emergency

boration

isrequired

Power

Operations,

Hot

Shutdown

1.1.3

General

Emergency

[SG2]

RED

path

inF-0.1,

SUBCRITICALITY

AND

Actual

orimminent

entry

into

either:

RED

path

inF-0.2,

CORE

COOLING

OR

RED

path

inF-0.3,

HEAT

SINK

Power

Operations

oWF1.0

Status

1.0

CSFST

Status

1.2

core

Cooling

1.2.1

siteArea

Emergency

[fpl

/fl,

rlj

CSS4]

ORANGE

or

RED

path

inF-0.2,

CORE

COOLING

Power

Operations,

Hot

Shutdown

1.2.2

General

Emergency

[fl1

,ri,

cpl]

RED

path

inF-0.2,

CORE

COOLING

AND

Functional

restoration

procedures

not

effective

within

15min..

PowerOperations,

Hot

Shutdown

1.0

CSFST

Status

4o0

Categolffr.O0

CSFST

Status

1.3

Heat

Sink

1.3.1

site

Area

Emergency

[fpl

,rpl]

[554]

RED

path

inF-0.3,

HEAT

SINK

AND

Heat

sink

isrequired

Power

Operations,

Hot

Shutdown

1.0

CSFST

Status

1.4

Integrity

1.4.1

Alert

[rp1]

RED

path

on

F-0.4,

INTEGRITY

Power

Operations,

HotShutdown

1.0

CSFST

Status

1.5

Containment

1.5.1

General

Emergency

[fl,

ri,

cpl]

RED

path

F-0.5,

CONTAINMENT

resulting

from

loss

ofcoolant

0

,NW

categolr.0

CSFST

Status

Power

Operations,

Hot

Shutdown

2.0

Reactor

Fuel

2.1

Coolant

Activity

2.1.1

Unusual

Event

[SU4]

Coolant

sample

activity:

*1.

0ACi/cc

dose

equivalent

1-131

OR

*100/(E

Bar)

ACi/cc

for

all

noble

gases

with

half-lives

>10

min.

All

2.1.2

Alert

[fl]

Coolant

activity

>300

ACi/cc

1-131

equivalent

Power

operation,

hot

shutdown

2.1.3

Site

Area

Emergency

(fl1

.rpl/rlj

Coolant

activity

>300

pCi/cc

1-131

equivalent

and

any

ofthe

following:

*RED

path

on

F-0.4,

INTEGRITY

oPrimary

system

leakage

>7S

gpm.

*RCS

subcooling

<SI

initiation

setpoint

*>

0.06

p&Ci/cc

on

R-11

and

R-12

due

to

RCS

leakage

Power

operation,

hot

shutdown

IgoIV2..0

:tor

Fuel

2.2

Containment

Radiation

2.2.1

Alert

[rlj

>0.06

p&Ci/cc

on

R-11

and

R-12

due

to

RCS

leakage

Power

operation,

hot

shutdown

2.2.2

Site

Area

Emergency

(fl,

rl]

Containment

radiation

monitor

R-25

or

R-26

>17

R/hr

Power

operation,

hot

shutdown

2.2.3

General

Emergency

(fl1,

rl,

cpl]

Containment

radiation

monitor

R-25

or

R-26

>68

R/hr

Power

operation,

hot

shutdown

2.0

Reactor

Fuel

2.0

Reactor

Fuel

CategoIT'2.o

Reactor

Fuel

2.3

Refueling

Accidents

or

other

.Radiation

Monitor.

2.3.1

Unusual

Event

[AU2]

Spent

fuel

pool

(reactor

cavity

during

refueling)

water

level

cannot

be

restored

and

maintained

above

the

spent

fuel

pool

lowwater

level

alarm

setpoint

All

2.3.2

Alert

[AA2

]

Confirmed

sustained

alarm

on

any

ofthe

following

radiation

monitors

resulting

from

anuncontrolled

fuel

handling

process:

"R-2/R-7

Vapor

Containment

Area

Monitors

"R-5

Fuel

Storage

Building

Area

Monitor

oR-25/26

Vapor

Containment

High

Radiation

Area

Monitors

"R-12

Containment

Gas

Activity

All

2.3.3

Alert

[AA2]

Report

ofvisual

observation

ofirradiated

fuel

uncovered

All

0

CategodT'3.0

Reactor

Coolant

System

3.0

Reactor

Coolant

System

3.1

RCS

Leakage

3.1.1

Unusual

Event

[SU5]

Unidentified

or

pressure

boundary

leakage

>10gpm

OR

Identified-leakage

>25

gpm

Power

operation,

hot

shutdown

3.1.2

Alert

[rpl]

3.*0

Reactor

Coolant

System

3.2

Primary

to

Secondary

Leakage

3.2.1

Unusual

Event

[ci]

Unisolable

release

ofsecondary

side

to

atmosphere

from

the

affected

steam

generator(s)

with

primary

to

secondary

leakage

>0.3

gpm

inany

steam

generator

Power

operation,

hot

shutdown

3.2.2

Primary

system

leakage

>75

gpm

Power

operation,

hot

shutdown

3.1.3

[SS5]

site

Area

Emergency

[fpl

,ri]

Bite

Area

Emergency

[rpl,

cl]

Unisolable

release

ofsecondary

side

to

atmosphere

from

the

affected

steam

generator(s)

with

primary

to

secondary

leakage

>75

gpm

Power

operation,

hot

shutdown

RVLIS

cannot

be

maintained

>39%

with

no

RCPs

running

OR

With

the

reactor

vessel

head

removed,

it

isreported

thatwater

level

inthe

reactor

vessel

isdropping

inan

uncontrolled

manner

and

core

uncovery

is

likely

Power

operation,

hot

shutdown,

cold

shutdown,

refuel

3.2.3

site

Area

Emergency

[fl,cl]

Unisolable

release

ofsecondary

side

to

atmosphere

from

the

affected

steam

generator(s)

with

primary

to

secondary

leakage

>0.3

gpm,in

any

steam

generator

AND

Coolant

activity

>300

pCi/cc

of1-131

Power

operation,

hot

shutdown

40

NW

CtglV.

Reactor

Coolant

System

3.0

Reactor

Coolant

System

3.3

RCB

Subcooling

3.3.1

Alert

[rl]

RCS

subcooling

<SI

initiation

setpoint

Power

operation,

hot

shutdown

Containment

4.0

Containment

4.1

Containment

Integrity

Status

4.1.1

Unusual

Event

[ci)

4.*1

Containment

Integrity

Status

4.*1.3

Both

doors

open

on

aVC

airlock

for

>4

hrs. OR

Inability

to

close

containment

pressure

relief

or

purge

valves

which

results

ina

radiological

release

pathway

to

the

environment

for>4

hrs

OR

Any

Phase

"A"l

or

Phase

"B"

orcontainment

ventilation

isolation

valve(s)

not

closed

when

required

which

results

ina

radiological

release

pathway

to

the

environment

Power

operation,

hot

shutdown

4.1.2

Site

Area

Emergency

[ri,

ci]

Rapid

uncontrolled

decrease

incontainment

pressure

following

initial

increase

OR

Loss

ofprimary

coolant

inside

containment

with

containment

pressure

orsump

level

response

not

consistent

with

LOCA

conditions

site

Area

Emergency

[ri,

cl)

Either:

Any

Phase

"A"l

or

Phase

"B"

orCVI

valve(s)

not

closed

when

required

following

confirmed

LOCA

OR

Inability

to

isolate

any

primary

system

discharging

outside

containment

AND

Radiological

release

to

the

environment

exists

asa

result

Power

operation,

hot

shutdown

4.1.4

cpl]

General

Emergency

[f1,

rl,

Confirmed

phase

"B"

isolation

signal

following

confirmed

LOCA

with

less

than

minimum

containment

cooling

safeguards

equipment

operating,

Table

4.3

AND

Any

indicators

offuel

clad

loss

,Table

4.1

Power

operation,

hot

shutdown

Power

operation,

hot

shutdown

S0

4.0

containment

CategIW4.0

Containment

4.0

Containment

4.1

Containment

Integrity

Status

4.1.5

General

Emergency

[fpl/f1,

rl,cl]

Either:

Rapid

uncontrolled

decrease

in

containment

pressure

following

initial

increase

OR

Loss

ofprimary

coolant

inside

containment

with

containment

pressure

orsump

level

response

not

consistent

with

LOCA

conditions

AND

Any

indicators

offuel

clad

damage,,

Table

4.2

Power

operation,

hot

shutdown

4.1.6

General

Emergency

[fpl/f1,

rl,cl]

Either:

Any

Phase

"All

or

Phase

"B"

or

CVI

valve(s)

not

closed

when

required

following

confirmed

LOCA

OR

Inability

to

isolate

any

primary

system

discharging

outside

containment

AND

Radiological

release

to

the

environment

exists

asa

result

AND

Any

indicators

offuel

clad

damage,

Table

4.2

Power

operation,

hot

shutdown

4.0

Containment

4.2.

SG

Tube

Rupture

v/

Secondary

Release

4.2.1

site

Area

Emergency

[rl,

cl]

Unisolable

secondary

side

line

break

with

SG

tube

rupture

as

identified

inE-3

"Steam

Generator

Tube

Rupture"

Power

operation,

hot

shutdown

4.2.2

General

Emergency

[fpl/f1,rl

cl]

Unisolable

secondary

side

line

break

with

SG

tube

rupture

asidentified

inE-3

"Steam

Generator

Tube

Rupture"

AND

Any

indicators

offuel

clad

damage,

Table

4.2

Power

operation,

hot

shutdown

V

I

0

WCa

tego

f7W%

.0Containment

4.0

Containment

4.3

Combustible

Gas

Concentrations

4.3.1

General

Emergency

(fl,rl,cplJ

34%

hydrogen

concentration

incontainment

Power

operation,

hot

shutdown

1.Coolant

activity

>30

0/h

Ci/c

cof

I131

2.Containment

radiationmonitor

R-25/R

26reading

>17

R/hr

3.RED

path

inF-0.2,

CORE

COOLING

Table

4.2

Fuel

Clad

Damage

Indicators

1.ORANGE

or

RED

path

inF-0.2,

CORE

COOLING

2.RED

path

inF-0.3,

HEAT

SINK

AND

Heat

sink

isrequired

3.Coolant

activity>

300

ACi/cc

ofI

131 4.Containment

radiation

monitor

R-25/R

26

reading

>17

R/hr

Table

4.1

Fuel

Clad

Loss

Indicators

Table

4.3

Minimum

Containment

Cooling

Safeguards

Equipment

Fan

Cooler

Units

Operating

Spray

Pumps

Required

<3

23

15

0

5.0

Radioactivity

Release

/Are

aRadiation

5.1

EffluentMonitors

5.1.1

Categol.

0Radioactivity

Release

Unusual

Event

[AUl]

5.0

Radioactivity

Release/

Area

Radiation

5.1

Effluent

Monitor.

5.1.4

Avalid

reading

on

anymonitors

Table

5.1

column

"HNUE"I

for>

60min.

unless

sample

analysis

can

confirm

release

rates

<2x

technical

specifications

within

this

time

period.

All

5.1.2

Alert

General

Emergency

[AGi

]

Avalid

reading

on

any

monitors

Table

5.1

column

"IGE

"Ifo

r>

15min.

unless

dose

assessment

can

confirm

releases

are

below

Table

5.2

column

"IGE

"Iwithin

this

time

period.

All.

[AA1)

Avalid

reading

on

anymonitorsTable

5.1

column

"Alert"

for

>15

min.

unless

dose

assessment

can

confirm

releases

are

below

Table

5.2

column

"Alert"within

this

time

period.

All

5.1.3

site

Area

Emergency

[AS1J

Avalid

reading

on

any

monitors

Table

5.1

column

"SAE"

for>

15min.

unless

dose

assessment

can

confirm

releases

are

below

Table

5.2

column

"SAE"

within

this

time

period.

All

5.0

Radioactivity

Release/Area

Radiation

CategoW.

0Radioactivity

Release

5.2

Dose

Proj

ecti

ons/

Environmental

Measurements/

Rele

arse

Rate.

S.2.1

5.2

Dose

Projections/

Environmental

Measurements/

Relearse

Rate.

5.2.4

Unusual

Event

[AUl

]

Confirmed

sample

analyses

for

gaseous

or

liquid

release

rates

>2x

technical

specifications

limits

for

>60

min.

All

5.2.2

Alert

[AA1]

Confirmed

sample

analyses

for

gaseous

or

liquid

release

rates

>200

xtechnical

specifications

limits

for

>15

min.

All 5.2.3

Alert

[AA1]

site

Area

Emergency

[AS1

]

Dose

projections

orfield

surveys

resulting

from

actual

or

imminent

release

which

indicate

doses/

dose

rates

>Table

5.2

column

"SAE"

at

the

site

boundary

or

beyond.

All 5.2.5

General

Emergency

[AGi]

Dose

projections

orfield

surveys

resulting

from

actual

or

imminent

release

which

indicate

doses/

dose

rates

>Table

5.2column

"IGE

"Iat

the

site

boundary

or

beyond.

All

Dose

projections

or

field

surveys

resulting

from

actual

or

imminent

release

which

indicate

doses/

dose

rates

>Table

5.2

column

"Alert"

atthe

site

boundary

or

beyond.

All

5.0

Radioactivity

Release/Area

Radiation

5.0

Radioactivity

Release/Area

Radiation

WCategoP.

0Radioactivity

Release

5.3

Area

Radiation

Levels

S.3.1

Unus

ual

Even

t[AU2]

Any

sustained

direct

ARM

readings

>10

0x

alarm

or

offscale

hiresulting

froman

uncontrolled

process

All

5.3.2

Alert

[AA3]

Sustained

area

radiation

levels

>15

mR/hr

ineither:

Control

Room

OR

Central

Alarm

Station

and

Secondary

Alarm

Station

All

S.3.3

Alert

[AA3

]

Sustained

abnormal

area

radiation

levels

>8

R/hrwithin

any

areas,

Table

5.3

AND

Access

isrequired

for

safe

operation

or

shutdown

All

Catego~p.

Radioactivity

Release

Monitor

NUE

Alert

SAE

GE

R-27

7.2E4

gCi/se:

3.60

Ci/sec

36.0

Ci/sec

360

Ci/sec

B/UHRVntMon

N/A

N/A

N/A

N/A

R-14

150,000

cpm

N/A

N/A

N/A

R-19

9.50

MCi/cc

475

MCi/cc

N/A

N/A

Table

5.2

Dose

Projection

/Env

.Measurement

Classification

Thresholds

Alert

SAE

GE

TEDE

10mRem.

100

mRem

1000

mRem

CDE

Thyroid

N/A

500

niRem

5000

niRe

m

External

exposure

rate

10mRem/hr

100

mRem/hr

1000mRem/hr

Thyroid

exposure

rate

N/A

500

mRem/hr

5000

mnRe

m/hr

(for

1hr.

ofinhalaton)

_______

Table

5.3

Plant

Areas w0

Tabl

5.

Efluen

Moito

ClasifcatonTresold

0

CategoqI5.0

Radioactivity

Release

oAuxiliary

Feedpump

BuildingU

oP.A.B.

oFuel

Storage

Building

oControl

Building

"ServiceWater

Pumps

oRefueling

Water

Tank

o.

Diesel

Fuel

Tanks.

"Vital

Area

Access

to

Containment

"AppendixR

Diesel

Generator

"Backup

Service

Water

Categoq.

0Electrical

Failures

6.0

Electrical

Failures

6.1

Loss

of

AC

Paver

Sources

6.1.1

Unusual

Event

[SUl

)

Loss

ofpower

to

all

Station

Transformers

5,

2,

3,6

for>

15min.

from

all

ofthe

following

offsite

sources:

oUnit

Auxiliarytransformer

0Station

Auxiliary

transformer

013W92

and

13W93

feeders

All

6.1.2

Alert

[SAl

]

Loss

ofall

safeguard

bus

AC

power>

15

min.

cold

shutdown,

Refueling,

Defueled

6.1.3

Alert

[SA5

]

Available

safeguard

bus

ACpower

reduced

to

only

one

ofthe

following

for>

15

min.: "

480V

EDG

31

o480V

EDG

32

"480V

EDG

33

oAppendixR

Diesel

oUnit

Auxiliary

transformer

oStation

Auxiliary

transformer

"13

W92

and

13W9

3feeders

Power

operation,

hot

shutdown

6.0

Electrical

Failures

6.1

Loss

ofAC

Power

Sources

6.1.4

Site

Area

Emergency

[551]

Loss

ofall

safeguard

bus

AC

power

>15

min.

Power

operation,

hot

shutdown

6.1.5

General

Emergency

(SG1

]

Loss

ofall

safeguard

bus

AC

power

AND

either:

Power

restoration

to

any

emergency

bus

isnot

likely

in24

hrs.

OR

Actual

orimminent

entry

into

ORANGE

orRED

path

on

F-0.

2,"CORE

COOLING"

Power

operation,

hot

shutdown

6.0

Electrical

Failures

6.2

Loss

ofDC

Power

Sources

IWCategoVIM.0

Electrical

Failures

6.2.1

Unusual

Event

[SU7]

<105

vdc

bus

voltage

indications

for>

15

min.

on

the

switchable

voltmeter

for

all

of

the

following

panels:

o31

o32

o33

'o34

Cold

Shutdown,

Refueling

6.2.2

Site

Area

Emergency

[SS3]

.<105

vdo

bus

voltage

indications

for

>15

min.

on

the

switchable

voltmeter

for

all

of

the

following

panels:

031

032

033

034

Power

operation,

hot

shutdown

*64

7.0

Equipment

Failures

701

Technical

Specification\Requirements

7.1.1

Unusual

Event

[SU2

J

Plant

isnot

brought

to

required

operating

mode

within

Technical

Specifications

LCO

Action

Statement

Time.

Power

operation,

hot

shutdown

7.0

Equipment

Failures

egoq.

0MW

ent

Failures

7.2

System

Failures

or

Control

Room

Evacuation

7.2.1.

Unus

ual

Event

[HUl

l

Report

ofmain

turbine

failure

requiring

turbine

trip

resulting

in:

Damage

to

turbine

generator

seals

OR

Casing

penetration

PowerOperations

7.2.2

Alert

[HAl]

Turbine

failure

generated

missiles

which

causes

or

potentially

causes

any

required

safety

related

system

orstructure

to

become

inoperable

Power

Operations,

Hot

Shutdown

7.2.3

Alert

(HA5

]

Entr

yin

toONOP-FP-lA,

"Safe

Shut

down

From

Outside

the

Control

Room"

All

7.0

Equipment

Failures

7.2

System

Failures

or

Control

Room

Evacuation

0

Categof7v7.0

Equipment

Failures

7.2.4

Alert

[SA3]

Reactor

coolant

temperature

cannot

be

maintained

<200

OF

Cold

Shutdown,

Refueling

7.2.S

Unplanned

loss

ofmost

safety

system

annunciators

or

indications

on

Control

Room

Panels,

Table

7.3

for

>15

min.

AND

Increased

surveillance

isrequired

for

safe

plant

operation

Poweroperation,

hot

shutdown

sitsArea

Emergency

[HS2]

Control

Room

evacuation

AND

Plant

control

cannot

beestablished

per

ONOP-FP-1A,

"Safe.Shutdown

FromOutside

the

Control

Room"

in2

15min.

All

7.3.2

Unusual

Event

(SU6]

Loss

ofall

communications

capability

affecting

the

ability

to

either:

Perform

routine

operations

(phones,

sound

powered

phone

systems,

page

party

system,

and

radios/walkie

talkies)

OR

Notify

offsite

agencies

orpersonnel

(ENS,

Bell

line

s,FAX

transmissions,

and

dedicated

phone

systems)

All

7.0

EquipmentFailures

7.0

EquipmentFailures

7.3

LossofIndications/Alarms

CommunicationCapability

7.3.1

7.3

LossofIndications/Alarms

CommunicationCapability

7.3.3

Unusual

Event

[SU3]

Alert

[SA4

1

Unplanned

loss

of

most

safety

system

annunciators

or

indications

onControl

Room

Panels,

Table

7.3

for

>15

min.

wcategoIM.

Equipment

Failures

AND

Increased

surveillance

isrequired

for

safe

plant

operation

AND

either:

Asignificant

plant

transient

in

progress

OR

CFM4

Sand

QSPDS

are

unavailable

Power

operation,

hot

shutdown

7.3.4

site

Area

Emergency

CSS6

]

Loss

ofmost

safety

system

annunciators

or

indications

on

Control

Room

Panels,

Table

7.3 AND

Loss

of

CFMS,

QSPDS

and

other

control

room

indicators

needed

to

monitor

critical

safety

function

status

AND

Asignificant

plant

transient

inprogress

Power

operation,

hot

shutdown

Table

7.3

Vital

Control

Room

Panels

SAF

SBF-1

SBF-2

CD

EF

GH

J-K

LH

N

0FAF

FBF

FCF

FCF

--

--

74j

catego"

.Hazards

8.0

Hazards

8.2.

Security

Threat.

8.1

securityThreat.

8.1.3

Unusual

Event

[HU4

]

Bomb

device

orother

indication

of

attempted

sabotage

discovered

within

plant

Protected

Area

but

outside

Plant

Vital

Areas,

Table

8.2.

OR

Any

security

event

which

represents

apotential

degradation

inthe

level

of

safety

ofthe

plant.

All 8.1.2

Alert

[HA4]

Intrusion

into

plant

ProtectedArea

by

an

adversary.

OR

Any

security

event

which

represents

an

actual

substantial

degradation

ofthe

level

ofsafety

ofthe

plant.

site

Area

Emergency

[HSl]

Intrusion

into

aplant

security

vital

area

by

an

adversary.

OR

Any

security

event

which

represents

Actual

or

likely

failures

ofplant

systems

needed

toprotect

the

public.

All

8.1.4

General

Emergency

[HG1]

Security

event

which

results

in:

Loss

ofplant

control

from

the

Control

Room

AND

Loss

ofremote

shutdown

capability

All

8.0

Hazards

8.0

Hazards

8.2

Fire

or

Explosion

w

8.1.1

All

Catego.

Hazards

Unusual

Event

[HU2]

Confirmed

fire

inor

contiguous

to

any

plant

area,

Table

8.2

not

extinguished

in

215

min.

ofControl

Room

notification:

Vehicle

crash

into

or

projectile

which

impacts

plant

structures

or

systems

within

Protected

Area

boundary

All

8.3.2

Unusual

Event

[HU3]

8.2.2

Unusual

Event

[Hil

l]

Report

by

plant

personnel

of

an

explosion

within

Protected

Area

boundaryresulting

in

visible

damage

tonon-vital

permanent

structures

orequipment.

All

8.2.3

Alert

(HA2]

Fire

orexplosion

inany

plant

area,

Table

8.2,

which

causes

orpotentially

causes

any

required

safety

related

system

or

structure

to

become

inoperable

All

Report

or

detection

oftoxic

orflammable

gases

that

could

enter

or

have

entered

within

the

Protected

Area

boundary

in

amounts

that

could

affect

the

health

of

plant

personnel

orsafe

plant

operation

OR

Report

by

local,

county

or

state

officials,

orUnit

2,

for

potential

evacuation

ofsite

personnel

based

on,

offsite

event

All 8.3.3

Alert

(HAl

]

Vehicle

crash

or

projectile

impact

which

causes

orpotentially

causes

any

required

safety

related

system

or

structure

to

become

inoperable,

Table

8.2

All 8.0

Hazards

8.3

Man-Made

Events

8.3.1

Unusual

Event

[Hill]

8.3

Man-MadeEvents

8.3.4

Alert

846

8.2.1

All

8.0

Hazards

[HA3]

Hazards

Report

or

detection

oftoxic

or

flammable

gases

within

aplant

area,

Table

8.2,

in

concentrations

that

will

be

life

threatening

to

plant

personnel

or

preclude

access

to

equipment

(evenwhen

using

personal

protective

equipment)

needed

for

safe

plant

operation

All

Earthquake

felt

inpiant

based

upon

aconsensus

ofControl

Roomoperators

on

duty

AND

either

Kinemetrics

StrongMotion

Accelographs

inthe

VC

produce

an

alarm

inthe

Control

Room

OR

At

least

one

amber

Peak

Shock

Annunciator

isli

t

All

8.4.2

Unusual

Event

[HUll

Report

by

plant

personnel

oftornado

within

plant

Protected

Area

boundary

All

8.4.3

Unusual

Event

[HUl)

River

level

314.51

(OMSL)

OR

Intake

structure

level<-4.5'

(OMSL)

All

8.0

Hazards

8.0

Hazards

8.4

Natural

Events

8.4.1

Unusual

Event

[HUl]

8.4

Natural

Events

18.4.4

Alert

[HAl

]

800

catego

Hazards

Earthquake

felt

inplant

based

upon

aconsensus

of

Control

Room

operators

on

duty

AND

Kinemetrics

Strong

Motion

Accelographs

in

the

VC

produce

an

alarm

inthe

Control

Room

AND

Amber

and

red

Peak

ShockAnnunciators

indicate

seismic

activity

All

8.4.5

Alert

[HAl]

Sustained

winds

>90

mph

OR

Tornado

strikes

aplant

vital

area,

Table

8.2

All

8.4.6

Alert

[HAl]

Assessment

by

the

Control

Room

personnel

that

anatural

event

has

occurredwhich

causes

orpotentially

causes

any

required

safety

related

system

orstructureto

become

inoperable,

Table

8.2

River

leve

l3

I5'

(OMSL)

OR

Intake

structure

level

resulting

ina

loss

of

service

water

flow

All

All8.0

Hazards

8.4

NaturalEvents

8.4.7

Alert

[HAl]

.Q-

Table

8.2

Plant

Areas

"Auxiliary

Feedpump

Building

oP.A.B.

oCAS/SAS

oFuel

Storage

Building

"Control

Building

oControl

Room

"ServiceWater

Pumps

"Refueling

Water

Tank

oEDGRooms

oDiesel

Fuel

Tanks

oVital

Area

Access

to

Containment

o-Appendix

RDiesel

Generator

oBackup

Service

Water

Categof7D.0

other

9.0

Other

9.1.01

Unusual

Event

Any

event,

asdetermined

by

the

Shift

Supervisor

or

Emergency

Director,

that

could

lead

to

or

has

led

to

apotential

degradation

ofthe

level

ofsafety

of

the

plant.

All

9.0Other

9.1.4

Alert

Any

event,

asdetermined

by

the

Shift

Supervisor

or

Emergency

Director,

that

could

lead

or

has

led

toa

loss

or

potential

loss

ofeither

fuel

clad

or

RCS

barrier,

AttachmentA.

Power

operation,

hot

shutdown

UnusualEvent

Any

event,

asdetermined

by

the

Shift

Supervisor

orEmergency

Director,

that

could

lead

to

or

has

led

to

aloss

or

potential

loss

of

containment,

Attachment

A.

Power

operation,

hot

shutdown

9.1.3

Alert

Any

event,

asdetermined

by

the

Shift

Supervisor

or

Emergency

Director,

that

could

cause

or

has

caused

actual

substantial

degradation

ofthe

level

of

safety

ofthe

plant.

All

9.1.5

Bite

Area

Emergency

As

determined

by

the

Shift

Supervisor

or

Emergency

Director,

events

are

inprogress

which

indicate

actual

or

likely

failures

of

plant

systems

needed

to

protect

the

public.

Any

releases

are

not

expected

to

result

inexposures

which

exceed

EPA

PAGs.

All

9.1.6

Site

Area

Emergency

Any

event,

asdetermined

by

the

Shift

Supervisor

or

Emergency

Director,

that

could

lead

or

has

led

to

either:

Loss

or

potential

loss

ofboth

fuel

clad

and

RCS

barrier,

Attachment

A.

OR

Loss

orpotential

loss

ofeither

fuel

clad

orRCS

barrier

inconjunction

with

aloss

ofcontainment,

Attachment

A.

9.1.2

Categaryw

Other

Power,operation,

hot

shutdown

9.0

Other

9.1.7

General

Emergency

As

determined

by

the

Shift

Supervisor

or

Emergency

Director,

events

are

inprogress

which

indicate

actual

or

imminent

core

damage

and

the

potential

for

alarge

release

ofradioactive

material

inexcess

of

EPA

PAGs

outside

the

site

boundary.

All

9.1.8

General

Emergency

Any

event,

asdetermined

by

the

Shift

Supervisor

orEmergency

Director,

that

could

lead

or

has

led

toa

loss

ofany

two

fission

product

barriers

and

loss

or

potential

loss

ofthe

third,

Attachment

A.

Power

operation,

hot

shutdown

'9.0

^ newbrkf'bwer authority william j. cahill, jr. chief ...bases, revision 1, january 10, 1995,...

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