ML14164A303

Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 2 of 7

ML14164A303
Person / Time
Site:	South Texas
Issue date:	05/15/2014
From:	South Texas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14164A341	List: ML14164A300 ML14164A303 ML14164A305 ML14164A308 ML14164A311 ML14164A314 ML14164A318 NOC-AE-140030, Revision to Unit 1 and Unit 2 Emergency Action Levels, Evaluation of the Proposed Changes License Amendment Request (LAR) for Revision of EAL Scheme to NEI 99-01 Revision 6. Part 7 of 7
References
NOC-AE-14003087
Download: ML14164A303 (83)

Text

CU4 ECL: Netificatien of Unusual Event UNUSUAL EVENT Initiating Condition: Loss of Vital DC power for 15 minutes or longer.

Operating Mode Applicability: Cold Shutdo, n, Refueling 5, 6

[ --amplk-Emergency Action Levels:

Note: The Emergency Director should declare the Unusmal Event UNUSUAL EVENT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1)

Indicated voltage is less than (site spe.ific bu. s voltage v*aue) 105.5 VDC on required Vital DC buses 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 tnefeaseextend the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A efand C is-are 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 UnuIual Event UNUSUAL EVENT. A loss of Vital DC power to Train A and!or C would not warrant an emergency classification.

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

Depending upon the event, escalation of the emergency classification level EMERGENCY CLASSIFICATION LEVEL would be via IC CAI or CA3, or an IC in Recognition Category AR.

CU4 - EAL-I Selection Basis:

The minimum voltage for Class 1E 125 VDC battery buses was determined in calculation 13-DJ-006, Rev. 3 to be 105.5 volts. At 105.5 volts or less, 0POP05-E0-ECOO, Loss of All AC Power, directs the operators to open the battery output breakers.

REFERENCES:

1.

Calculation 13-DJ-006. Rev. 0, 125 VDC Battery Four Hour Coping Analysis

2.

OPOP05-EO-ECOO, Rev. 23, Loss of All AC Power 64 1 P a g e

DeVelOper Notes; The "site specific bus voltage value" should be based em the mainimum bus voltage necessary for adequate oppr-ation of S4AFT SY-4 3STEFM equipment. This welltag výaluc shoud ineer-omte a mar-gin of at least1 moin~tits ef oper-tion befer-e the onset of inability to operate these leads. This veltage is usually neer the mninimaum voltage selectcd when battcrfj sizing is peffcrmed.

The tyical value for an entirc battor-y scsapoimately 105 YDC. Fer-a 60 eel! strinig of batteries, the eel!

'y I

1i v

co 1

• *I 1.8 1 Volts per-de,.

rEi A Ait..k, i ~. I I I A 4

-ARA p-.

A V RAP Zia 65 1 P a g-e

CU5 ECL: Notificatien of Unusual Event UNUSUAL EVENT Initiating Condition: Loss of allALL onsite or offsite communications capabilities.

Operating Mode Applicability: Cold Shutdown, Rcfuzling5._6, Defueled F*mnple-Emergency Action Levels: (1 or 2 or 3)

(1)

Loss of ALL of the following Onsite communication methods in Table C4.

(Site Spco%_ir4Vc list ofcomniaicn ethOdS)

(-*H(2)

Loss of ALL of the following Offsite Response Organization (ORO) communication methods in Table C4.

(site specifie list of communieatiens methods),

(3)

Loss of ALL of the following NRC communication methods in Table C4.

(site speeifie list of C4mmucnicaiotns methods)

Table C4: Commu~nicaitions Methods~

EAL-1 EAL-2 EAL-3 ONSITE ORO NRC Plant PA system X

Plant Radios X

Plant telephone system X

X X

Satellite phones X

X Direct line from Control Rooms to Bay City X

X Microwave Lines to Houston X

X Security radio to Matagorda County X

Dedicated Ring-down lines X

ENS line I

X 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, etc.).

EAL # 1-addresses a total loss of the communications methods used in support of routine plant operations.

66 1 P a g e

EAL #2-addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here arc D......

t..) Matanorda County Sheriffs Office, and Texas Department of Public Safety Disaster District in Pierce.

EAL #3-addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CIJ5: EAL-1. EAL-2. and EAL-3 Selection Ba~i*:

CU5: EAL-I EAL-2 and EAL-3 Selection Basis:

Lines not included for offsite communications to ORO and NRC included links that would need relaving of information. Links were obtained from procedures OPGP05-ZV-001 1, Emergency Communications.

REFERENCES:

1.

OPGP05-ZV-00 11, Rev. 8, Emergency Communications EAL 1-1 The "site spe.ifi. list of communi.ation+s m.th.ds" shouild inludc all comm.nic-ationis mc.th.ds ts.d fcr: routtino plant cefmmunicatins (e.g., commercial er-site telephones, page part)' systems, r-adies, ete.). This listing shouild includc installd plant equipment and compenents, and noet items ewned and maintained by EAL 412 The "site sp..ific list of em. muniati.ns methods" sh..uld inleludI all coImmunicif.i÷atiens mctheds used to pcrforam intitial emer-gency notificationts to OROs as descr-ibcd in the site 59 Emergency 4nI-An. 4-AInc uting SH804 10,419 14nciuic instic i

pan equipmcnft anzl fi oponontS, A449 not AtomS 81A'ncl iant M-ainlt-Ain-d-by indiNiduals. Example mo-th-d-s -are ring dowm'dedicatcd telcphenc linces, eommcrceial telephone lines, r-adios, satcllite tclephonces and internct based comimunicationts teehnology.

in the Basis scction, inseit the site specific listing of the OR~s r-equir~ing notification ef an cmer-geney dcclar-ation ftom the Control Roomf in accordanoc with the site Emer-geney Plan, and typically within 15 mninutcs Er-A 1-9 3 The "site speeific list of communicationis methods" should ineludc all comF~muieations methods use' to pcfcrmff iniitial Omffergeny notifiationlS to thO NRC; as deScribed in-theA s-ite mrgny ln The listing shul iicluide installed plant equipment and components, and not itcmes e%%ed and maintainod by indiv'iduals. These mctheds arc t)'pieally the dedieatcd Emer-gency Notification System (ENS) tclephonce line and commerceial telephonelines-v T*T A

ý. I -

ek-4L Asslgnmmnt :uincuic..

I.....L 671 P a g e

CAI ECL: Alen-ALERT Initiating Condition: Loss of(r-eaeoe*.esse"*CS IPWRI or* RPr

[BWRI) inventory.

Operating Mode Applicability: Cold Shutdo-,.n, Refueling 5,6 Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare the A4eft-ALERT promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1)

Loss of (raeeteF re.se. RCS [PWR*] OF RPM 1BW4 R

) inventory as indicated by level less than (si9-speeifie level) 32 ft. 9 inch (+ 6 inches above hot leg centerline).

(2) a. Rea*t*r.. vzc,,R CS [PrAI -r

.RPV 1BWRI) level cannot be monitored for 15 minutes or longer AND

b. UNPLANNED inereaserise in (SITE SPECIFIC SUMP AND/OR TANK) ANYANY of the following sump or tank levels in Table C2 levels-due to a loss of (reactor vesselRCS [PWR, OFr PRPV f#WR44-inventory.

Table C2: RCS Leakage Containment Normal Sump Pressurizer Relief Tank (PRT)

Reactor Coolant Drain Tank (RCDT)

MAB Sumps I thru 4 Containment Penetration Area Sump SIS/CSS Pump Compartment Sump Basis:

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.

EAL #1-A lowering of water level below (site specific level) elevation 32'- 9" indicates that operator actions have not been successful in restoring and maintaining (reactor vesselR.CS [PWR or RPM [BWRI) water level.

The heat-up rate of the coolant will ine-easerise as the available water inventory is reduced. A continuing deerease-reduction 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 inefeaserise in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

68 1 P a g e

EAL 92-The inability to monitor (reactor vessel/RCS [.P.WRj ofr RZ [R WRj) 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 (reactor vessel/RCS.... w[or.eri :R f#-W-I4The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1 If the (reactor vessel/RCS [PRI/p] er-RP7 [BJR]) inventory level continues to lower, then escalation to Site-Area Emergeney-SITE AREA EMERGENCY would be via IC CS 1.

CAI: EAL-I Selection Basis:

The minimum RCS level at which an RHR pump can be started per OPOP02-RH-0001 is 32 feet 9 inches (+ 6 inches above hot leg centerline). If RCS inventory is reduced below this level, normal decay heat removal systems may not be available for core cooling. This threshold is not applicable to reduced inventory vacuum fill since this is a controlled evolution and not indicative of an RCS loss.

CAI: EAL-2 Selection Basis:

The tanks/sumps selected for this EAL were obtained from OPOP04-RC-0003, Excessive RCS Leakage. Since other system leaks could raise levels in various tanks and sumps, the list was limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.

Although procedure OPOP04-RC-0003 is designated for use in modes 1-4. its logic is applicable to this EAL.

REFERENCES:

1.

OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage

2.

OPOP02-RH-000.1. Rev. 63, Residual Heat Removal System Operation DeVelcpcr NctcS:

Forf EA-1 41 the "Site Specific. le'Vel" ShOUld1 be based On either:

" IIBWRI Low Lem, EGGS acttuteion setpoinU~bevel 2. This setpoint was ehosen because it is a stanidard p Ratoally Signifinant setpoint at whieh some (tý,ieally high pr~essure EGGS) injection systems w 'ould autotmatieally staft and is a value signifieantly below the lew RPMZ water-level RPS aetuatinston speeified in WC CUI.

[P2WRI The minimum alloAW0able level that supports operation of normally used decay heat r-emoval sys'tcms (e.g., Residual Heat Removal OF Shutd..o Cooling). Iffmultiple levels exist, speei6' each along with the anrpit ode or: configuration dependency en-teria-.

Fer-EAL 112 The type and range of RCS level instr~umentto ma way during an outage as the plant moeves-throeugh var-iosperin modes and refueling evoluitions, paoiticlar-ly for-a PA As appropriate to the plant design, alteffate mfeafn's of determinfing RCS_ level -areP insitalled to assure that the albility to monitor. level Awithin the range requir-ed by oper-ating proeedurfes will not be interrupted. The instrumentatio rag ncssar-y to 69 1P a g e

SuPP01t imHPIOMOntatio 01 cpcrfAtlg ffreatoeus in the GeldJ bhut~le% and Refiuoling mo~oes mnay he diltoron fte.g., naflarrwe) thanRHI tha lt rgiro Euing meaes higior tHan LWJl

?5hUt~eWf.

Eniter anfy "Site speeif~ie st iinventor,' (i.e., the lost irn tUC'1 A

A

-:1 I! I

)

1 I

J 1

nip aaORWO tmm ic;'cz toot 8CL DOif Be efeOOIC W inoerons MR

@-1 o wor af Oiczz: M.01AA A-

-I I I eontwry wattle 011cr t11e listed Sumfp orF taIIKJ.

ý H teq-- -. -.-.-

70-LP a g e

CA2 I ECL: Aen ALERT Initiating Condition: Loss of oIALL offsite and IIALL onsite AC power to emergency buses for 15 minutes or longer.

Operating Mode Applicability: Cold ShutdozW

, Refueling5, 6, Defueled I F&ape-Emergency Action Level*:

Note: The Emergency Director should declare the A4eI4t-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 to (site pecific emergency buses) MUALL three 4160V AC ESF Busses 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.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Ar:ea Emergeney SITESITE AREA EMERGENCY because of the inereasedadditional 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. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

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

Escalation of the emergency clawsifization level EMERGENCY CLASSIFICATION LEVEL would be via IC CS1 or A-I-RS1.

CA2 - EAL-1 Selection Basis:

N/A

REFERENCES:

1.

OPOP04-AE-0001, Rev. 44, First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus

2.

OPOP04-AE-0004, Rev. 15, Loss of Power to One or More 4.16 KV ESF Bus

3.

OPSP03-EA-0002, Rev. 32, ESF Power Availability

4.

Drawing OOOOOEOAAAA, Rev. 24, Single Line Diagram. Main One Line Diagram. Unit No. 1 & 2 711 Page

DevelopeF Nots I ý -

I milirirue "rrnnruTflr" mr r" I ln'1'rr ri'v'1' "~m1in "nriii r'TI'u1 ;u"~ r-'~v-~--

ffawber ef operating generaters ntecessary Bfor that SOurcc tO prc;'ide adcquate power-to ant AG emefrgcney bus.

For example, if a backuip pwer sorc is compr

~fised of two genierator-s (i.e., ýwo 50,1 eapaeity genierator-s sized to feed 1 AG emergency bus), the EAL and Basis section maust speeifý that beth generators9 for-that sour~e arc The "site specific. emer-gency buses" are the buses fed by offeite or: emergencey AG power sources that supply pewer-tc the eleetrieal distribution system that pewcr-s SAFETY SYSTEMS. T-here is tpoically 1 emer-geney bus pcr train of SAFET-Y SYSTEMS.

The EAL andior-Basis scetinmyepf use ef a nont safety related power-source pr-evided that oper-ation ei this sourve is eeontrOllcd in accordancee with abnonnmal or-emregeney perain prcdurs, or. beyonid design basis aeeidcnt r~espose guidelines (e.g., FLEX sutppo4 guidelines), Suchb powe~r. soumrce ;hetgd gefter-P1y maeet the "Alternate ae source" definition provided in 10 CFR 50.2.

At multi unit stations, the EA~s may credit compnsatry easur-es that are pr-eeedtr-alized and canb implemented w'.ithin 15 minutes. Consider capabilities suchb as power sourcc cro-ss ties, "9%4ng" generators9, other-poersorcsdescr~ibed in abnon~mal or-emergency oper-ating proeedurfes, etc. Plants that have a proceduiralized capabiliy to supply off-site AC power: to an affccted unit via a croess tie tacopin it may credit this-power-soutrce in the EAL proevided that the planned cross,1 tiesatgmeets the requir-ements of 10CR5.3 t

4-LLAsgmn AtmotflWUeS: 4.12.4.

72 1 P a m e

CA3 I ECL: Alei4 ALERT Initiating Condition: Inability to maintain the plant in cold shutdown.

Operating Mode Applicability: CcGd Shutde,., Rcfucling 5 SE-9*mple-Emergency Action Levels: (1 or 2)

Note: The Emergency Director should declare theAlei4-ALERT promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

(1)

UNPLANNED ineFeaserise in RCS temperature to greater than (site.p..ifi.

T*z*:aial Sp-^ifieatien

,e.ld shutdown t.mp.ratur..

limit) 200 0F (Tavg) for greater than the duration specified in the-fellewing table-Table C3.

Table C3: RCS Heat-up Duration Thresholds RCS Status Containment Closure Status Heat-up Duration Intact (but not at reduced inventory Not applicable 60 minutes*

W4_Rj)

Not intact (or at reduced inventory Established 20 minutes*

H.WR- )

Not Established 0 minutes

• 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 imr-eaerise greater than (site

.p..ifi pr.....u. rvadig)l0 2sig. (This EAL does not apply during water-solid plant conditions. [PW-RI)

Basis:

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

EAL #1-The RCS Heat-up Duration Thresholds table addresses an uieFeaserise in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact, or RCS inventory is reduced (e.g., mid-loop operationi*-PWR-s). The 20-minute criterion was included to allow time for operator action to address the temperature ineraserise.

The RCS Heat-up Duration Thresholds table also addresses an inefeaserise in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature inefea&erise without a substantial degradation in plant safety.

73 IPagze

Finally, in the case where there is a* *nefearerise in RCS temperature, the RCS is not intact or is at reduced inventory fPQW 14J, and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel.

EAL #2: provides a pressure-based indication of RCS heat-up.

Escalation of the emfergeney CSI or ARS!AS!RS1.

F *Velassificatien level EMERGENCY CLASSIFICATION LEVEL would be via IC CA3 - EAL-1 Selection Basis:

Table C3 was adopted from NEI 99-01, Rev. 6. This EAL addresses the concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal. A number of phenomena such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design. and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncover can occur. NRC analyses show that there are sequences that can cause core uncovery in 15 to 20 minutes, and severe core damage within an hour after decay heat removal is lost. The allowed time frames are consistent with the guidance provided by Generic Letter 88-17 and believed to be conservative given that a low pressure containment barrier to fission product release is established.

CA3 - EAL-2 Selection Basis:

An UNPLANNED RCS pressure rise greater than 10 psig provides a pressure-based indication of RCS heat-up.

The pressure change. per NE1 99-01 Rev. 6, is the lowest change in pressure that can be accurately determined using installed instrumentation, but not less than 10 psig.

REFERENCES:

1.

Technical Specifications Table 1.2 (Mode, Temperature, Power, keff Table)

Dev*,ope, Notes.:

Fer E-AL HI Enter. the "Site speeific Technlical speeification 891d shutdown temper-atuelimt"wh indicated-.

The RCS shetuld be considered intaet or noet intact in aecordancc with site speeifie criter-ia.

PForEAL #-92 The "site Specific prcssure rcading" should be the lowest change in pr-essure that emn be accuratey detcrmined using insialied ifistfumcntatien, but noet less thant 10 psig.

For-PSARs, &si WC and its m~cis-eeiated EALs addfess the cnrzrisd by GCncriff Lcttfr 88 17, AeS,9 efipe~a)

THai Remeval. A number-efphefnomena such as pressrztin ;'rcig t

am gnr-ater-U tube diraining, RC level differencees when operating at a mid loop condition, decay, heat removal system design, and level intrmnttinprObleoms cman lcadd to conaditions hrcdecay hemt romovafl is lest Rand eero uncegvffry an Occurf.

NRC analyses show that there arce seguenees that can cause corce uneever-y in 15 te 20 mninutes, and seNvcr-cor-e damage within an hour aftcr dccay heat FcmeNval is lost. The allewcd tifnc framcs arc consistcnt with thc 74 1 P a Pge

guidmee centainm 9CL Assi orzevided b;, Gznzflrip Izttcr. 8R 17 And hplicvpd $A hp xpzra ivz rn that n 1low. f;;Rssu~

I ent MRrcrt n~z proouei roicase is esiablished:

Rnmpcnt trbtz 12B 75 I P a g-e

CA6 ECL: Aleft-ALERT Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.

Operating Mode Applicability: Cold Shutde., R.f.. linig.*

5 I Ea.mpf-Emergency Action Levels:

I (1) a. The occurrence of ANY of the following hazardous events in Table C5:

Table C5: Hazardous Events Seismic event (earthquake)

" Internal or external flooding event High winds or tornado strike FIRE EXPLOSION (site sp*.ifie hazardo)Predicted or actual breach of Main Cooling Reservoir retaining dike along the North Wall

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

b. EITHER of the following:

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

OR

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

Basis:

This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant.

76 1 P a g e

EAL-#_.b. 1-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.

EAL_#-I.b.2 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 emer-CS l or ARSI AS 1 RS1.

eaev elassificatien level EMERGENCY CLASSIFICATION LEVEL would be via IC CA6: EAL-I Selection Basis:

The listed hazards are taken directly from NEI 99-01, Rev. 6. The only additional hazard was the inclusion of the Main Cooling, Reservoir since it is a credible hazard and analyzed in the STPEGS UFSAR (reference 2).

REFERENCES:

1.

STPEGS UFSAR, Rev. 13, Section 3.4.1, Flood Protection Pfe!Opff Notes+

POr kzite Specifie h"OAzr, EdcN'0I8Pcrz Should eensidefr incwiung 9thcr zignifieent, site spee~ifz haimr-Els to the bullztz-4d list P-8ntaineed ini EAL !.a (z.-.. a zih)

-I*

I I

I I.

l -n-1 11, 1-1-1 PcWefr Piont e'ortf

. I -~

~ i

.r -cp~~

.i -. 1 (Ir eac cprafl3rcunan ianeC A

.glmn aecordanee with site speeifie design cntr-ite.

ECL Assiknment Attributes: 3.1.2.B 771Page

CS1 I ECL: Site Afea Emfert SITE SITF AREA FMFR(jFNCY efley Initiating Condition: Loss of (Cfeaete vesseCS IPWR] or-RPX,BWRJ) inventory affecting core decay heat removal capability.

Operating Mode Applicability: Cold Shutd*zvn, Refueling a56 E&*omp!e-Emergency Action Levels: (1 or 2-ef4)

Note: The Emergency Director should declare the Site Area Emergene-y SITESITE AREA EMERGENCY promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.

(1) a. CONTAINMENT CLOSURE not established.

AND

b. (Reaetef-vesse4CS [P TRI or-RPV [BAR]level less than (site specific level) 33% of plenum.

f. 1'%X)T'r A

[TP I.T'r f.T f'C', T" T-less Ln t-2ý7i

-A-

ý -, -

ý -

P-Q*-A--

q R

A 41

b. (Reactor vesseWRCSJ IlPWRI or-RPV I[WRI) level less than (site specific level).

L (2) a. (3)-a._(Reaeter: -ess4RCS -PWRj or-RPV*

• [,W) level cannot be monitored for 30 minutes or longer.

AND

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

&+

ýs

ýLtzLW,-nrtnr I(ntlnfain-t I-lii1no XI 4x

.A a

t -~1,gig~n 1/Innitnre RE-8055 or RE-8099 reading greater than 9,000 mR/hr. : (9itc spesific value)

Erratic source range monitor indication [PW4 UNPLANNED iner-easerise in levels in Table C2any of the fe uncovery (site sp..ifie s.m. and/or tank0 ANY of the following sump or tank 4ov'wing s.mps or-tank leve's-of sufficient magnitude to indicate core (Ot0bdc Pr.Q sitSeQ-4ee susc&fe in;;d i A t i en Q Table C2: RCS Leakage Containment Normal Sump Pressurizer Relief Tank (PRT)

Reactor Coolant Drain Tank (RCDT)

MAB Sumps 1 thru 4 Containment Penetration Area Sump SIS/CSS Pump Compartment Sum, 78 1Page

Basis:

This IC addresses a significant and prolonged loss of (reactor vessel/RCS IPWI or-R.P IV JWo'") nventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergcny 84T4--ISITE 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 roactrer vczAel leeRCS level. If R.CS/rlzatzr ;',zzz! !cAIRCS level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans bTyietily-provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The differcnee in thc specified R.CS/rc-acFtr ;':zz!

AIevA-lRCS level-+4 in EALe 1.b emd-2-b-reflects the fact that with CONTAINMENT CLOSURE not established, there is a ewef an raised probability of a fission product release to the environment.

hi-EAL.#42.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 performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor (reactor vessel/RCS [P....R J o

PrRP

[W4,jr

) 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 (reaeter-vieseMRCS jPW4RJ of t*P IWRI).

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.

Escalation of the eme'rgeeny ca!Rzz*.-;if4ctin !v EMERGENCY CLASSIFICATION LEVEL would be via IC CG1 or-AG-RG1.

79 1 P a P-e

CSI: EAL-I Selection Basis:

Per NEI 99-01 Rev. 6, the RCS level indication should be six inches (6") below the bottom inside diameter of the RCS loop. Six inches (6") below the bottom inside diameter of the RCS hot leg nozzle (31 '-0.5") is 30'-6.5" per OPOP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram. Per OPOP02-11-0002, RVWL Monitoring System, Addendum 1, RVWL Sensor Elevations, if water level is below 31 '-0.5" (dry) and above 30'-1.6" (wet), the Reactor Vessel Water Level Monitoring System indication would be 31 '-0.5" which corresponds to 33% of plenum and a water level of "Bottom of the Hot Len Nozzle".

CSI: EAL-2 Selection Basis:

Per NEI 99-01 Rev. 6, the RCS level indication should be approximately the top of active fuel. The RCS level which corresponds to the top of the active fuel is 28'-2" (0POP03-ZG-0009, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram). The Reactor Vessel Water Level Monitoring System is capable of indicating down to 29'-2.7". During Mid-Loop Operations a temporary level gauge is placed in service but has a bottom scale of 31 '-0". Neither of the level indicators will provide indication when level drops to/or below the 28'-2" elevation. Core uncover is determined by the secondary indications listed in this EAL. The secondary indicators of inventory loss include a list of tanks/sumps found in 0POP04-RC-0003, Excessive RCS Leakage.

Since other system leaks could rise levels in various tanks and sumps. the list has been limited to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.

As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C 129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. A rising trend on these monitors can be an indication that core uncovery is occurring. Additionally, erratic source range monitor indications, or lame level rises in the tanks listed can give further indication of core uncovera.

The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOCO 13-CALC-006 Rev. 1. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the top of active fuel. The high range of these monitors is 10 R/hr. The value of 9,000 niR/hr was selected to ensure that the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 mR/hr corresponds to approximately 8 inches above the top of the active fuel with the reactor head on: which provides an additional indication that RCS levels are near the point of fuel uncovery. These monitor readings in conjunction with the other threshold values allow for an accurate assessment of the EAL.

REFERENCES:

1.

Calculation No: STPNOCO 13-CALC-006 Rev. 1, Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds

2.

OPOP03-ZG-0009, Rev. 59, Mid-Loop Operation, Addendum 1, RCS/RHR Simplified Elevation Diagram

3.

USFAR, Rev. 15, Chapter 12, Table 12.3.4-1

4.

OPOP02-II-0002, Rev. 15, RVWL Monitoring System

5.

OPOP04-RC-0003, Rev 18, Excessive RCS Leakage

6.

Drawing 9C 129A81105, Re. 3, Radiation Zones, Reactor Containment Building, Plan at E. 68' - 0" 80ILPag e

poveloper-Notes-Accident analyses suggest that Awl damage may occur mithin Onle hour of tmeever-y depending upon the aount fticincshutdoym; r-efer. toA Gener-icA Letterf 88 1P, SEC 4 91 283, NIU RG 1119 And-4161-4 9 1 06.

The t"e and r-ange of RCS level instrumentto ma way during an ouitage as the plant moves through various opeatig mdes and rfufielinig cvolutiefns, particutlar-ly for-a PWR. AS. approepriate toe the plant des'gn alternate means of determinting RCS level mre installd to assur~e that the ability to moenitor level wvithint the ragereuied by operating proeedur-es ;.ill n ot be A interrupted. The instrumentation range necessary to suportimlementation of operating proeedures in the Cold Shutdown and Refuteling moedes may-be different (e.g., narrower) than that requirfed during moodes higher: than Celd Shutdown.

PWR For.EAL41t.b the "site speci fic level" is 6-" b-lelow ohc bettom WD of the RC Sloop. This is the l eve! at 6" beow the bottoma 1D of the rcaetorese sJpenelfatien and not the low point of the loop. If the availability of on scale level indieation is sueh that this level valuce ean be dctcrmined dtwing some shtttdeoA modes or conditiefns, but noet etheors, then qpecify, the mode dependWent ani'or. configumtion states during whieb the level indication is applieable. if the design and opefrAtion of water- !eye! inistrumentation is suceh that this level value cannot be determined at any time dur-ing Cold Shutdown or Refuieling modes, then do noet ineluide EAL 41 (classification %ill be aeeomplished in aceorfdanec with EAL 43).

Fmr 1A 4;2bh Thel "site specific level" should be approximately the top of active Afuel. If the availabilit)y Of Ont scale1R levMoel in~dicationlf is-q-uch that this lcvol valucp PAnR b dctcnnined dur-ing seme Shutdov~c m~doso conditions, but not othcrs, theft specify the mode dependent and/er configufation states dtffing which the level indlication is applicable. if the design and oper-ation o-fiwater. le-ve Inmntio is; siuh that this level value cannot be detcrmincd at any time dufnfg Cold Shutdeow or: Refueling modcs, then do noet ineludc EAL 42 (classifieation will be accomplishcd in accor~danec with EAL 43).

For EAL #3.b fir-st bullet As watcr level in the reactor-vessel lowcrs, the dose rate above the eeore will incr~ease. Eniter a "site specific r-adiation monitor-" that coulid be used to-de0tmet conrc tmeocr-y and the asocatd site specific Value" indic'ative of core uncover~y. it is Feecogized that thee con-aditionl deScribed by this IC may restilt in a r-adiation v~altt beyond the epcr-ating Or display r-ange of the installed radiationt moenitor-. Int those cases, EAL values should be determined with a margin suiffieint fe enstre that an accurate monitor reading is available. For: example, an EAL monitor-reading might be set at 90% to 95% of the highest accurate monitor-reading. This prvsonn 4'thstanding, if the estimate~'caleulated moneito reading is greater-than approx~imately 110Q% of t-he highest accurte mnitreadng then developers May cheeseR -not to-include-40 the monPitorf _as an indlication Mand-identify an alternate EAL, threslhold.

To further promote accuirate classification, developers Ishou-ld-conMsider1f if.-AA9 soecmiainoVoiors could be0 classification assessment.

For ELb 43.b second bullet Pest MI4 accident studies indicated that ~e installed PWR nuclear instfumentation enl orfate effatiefliv when the eere is oneovered annd that this qhAHuld hPe used WIQ aR tRAI ___

mA..kin..

such determinat.ions.

81 IPage

Fo EL f3b hir-d bullct Entcr: any 'site speei fie Sump and/ore tanke" lev'els that could be cxpcceted to ehanfg-e if ther~e were a less of RCS,'reaeter-vessel invcntery of suffieient magnitude to indicate eorc tneevery.

Specifie level values may be included if desir-ed.

For EAL ft3.b fiawh bullet DeveloperS Shouild detefmine if other-reliable indicators exist to identify fuel uneovery (eg. rmtviwing using eamcr-as). The goal is to identtify, any unique or-site speeifie indicationis, net already used elsew~her-e, that will pr-emote timely and aeeumate emer-geney classification.

Fer EAL 41l.b "site specific level" is the Low Low Low EGGS actato sepin / Level 1. The BNVR Low Loew Low ECCS actuation setpcint I Level 1 was chosen because it is a standard operatienally significant setpintat hich some (typically low pr-essurfe ECCS) injection systems would autematieally start and attempt to restor-e "V level. This is a RPV water-level value that is cbscr.'able below the Lew LowAIcvel 2 valuie specified in WC GAl, but significantly above the Top ef Active Fuel (T-OAF)4 threshold speeified in EAL #2.

For-EAL ft2.b-The "si te speci fie level" shouild be fer: the tep of active file!.

For EAL#13.b first bullet As water-level in the reactor-vessel lower-s, the dose rate above the core will iincrease. Enter a "site specific r-adiation moenitor-" that could be used to detect core uneover-y and the associated "site specific value" iniatv ofe~ meeyt is rteognized that the condition descr-ibed by this IC may r-esuilt int a r-adiation value beyond the oper-ating or-display range of the installed radiation monitor. In those eases, EA1 vallues;t shold be detemMined with a mar-gin sufficient te ensure that an accurate_

monfitor rceadinig is available. For. example, ant EAL monitor: reading fmght be set at 90% to 954 of the highest accurate monitor reading. This. prvso 4owthstanding, if the estimated/caleulated manitoi reading is greater than approxyimately I110% of the highest accurate Monitor. reading, then deVelopers May choose not to incelude the monfitor as an indicationt and ideniti6y ant alternate EAL thr-eshold.

TO further prOMote acceur-ate classificaition, de-velopeff should consider. if some combination of meaites couildb specified in the EAL to build int Em approepriate le-vel of coffobor-ationt between monitor-readings into the classification assessment.-

For-BWVRs that do not have instaled fediation moenitor~s capable of indictn irun cry, alternate site specific level indications ofecore uncover; shouild be used if available-.

For: EAL 4f3.b sceond bullet Because i3WR source range monitor- (SM4) nuclear instmumentatien defetectrs are typically located below cor-e mid plane, this may not b-e -a vi-able indic-ator f corefe oncover;y for-BXX4Rs.

For-EAL 4f3.b thir~d bullet Enter-any "site specific sump andor-tank" levels that could be expected to chang if there wer-e a less of RPNV inventory, of su-fficient maignitude to indicate core uncoer~iy. Specific level v-alues" may be included if desired-.

FoL~r P LA fti 43 u

hbuihhllet]

vlpr hua~triei te eibemiaoseitt dniyte uncover,' (e.g., r-emote viewing using camer-as). I he goal is to identity any tmiqlue or-site speciti

.I "11 indicatiEns notiaeady. used e._lshr. t-hat AAill rmtotimelyh. and1 auaefnemrienev elaissiticAtion-T"* W"* t A

A *

!1 "1 I T't N=IL A'ssieinment A'ttfleutes: A.144 821 P a e

CGI I ECL: Genera Emr.g.c.y. GENERAL EMERGENCY Initiating Condition: Loss of ffeaeteI-*essRCS [PWRj or-RPM *[BWR,) inventory affecting fuel clad integrity with containment challenged.

Operating Mode Applicability: Cold Shutdown.., Ref.eling. --

SE-anmple-Emergency Action Levels:

F-2)

Note: The Emergency Director should declare the Gner-al Emergency GENERAL EMERGENCY-promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.

'.JR1tAV 1 fl~V1t J J&VZROO~AOARA

~3t~

JS.,tt

• t V

loniger-AND~

F (1) a.

(RPaetf, r c*OSSRCS 1Pj OFr RPM IBJ1I) level cannot be monitored for 30 minutes or-longer.

(2) a.

(RoeatOr ;v'csg/ciCSi vIW r RPM t

BW-Rii,),*I l

elf cant bef Anitod for *30m mtinute or-longFr.

L J

w AND

b.

Core uncovery is indicated by ANY of the following:

(Site

.pecf.c radiation moenitr) Reactor Containment Building, 68'-0" Area Radiation Monitors RE-8055 or RE-8099 reading greater than 9,000 mR/hr.. (site sp. ifi'..

lue)

Erratic source range monitor indication

-4.WRj

_ UNPLANNED inefeaserise in (SITE SPECIFIC SUMP AND/OR TANK) ANY of the following sump or tank levels in Table C2 of sufficient magnitude to indicate core uncovery

• (Other-site speeifie indieation Table C2: RCS Leakage

" Containment Normal Sump

" Pressurizer Relief Tank (PRT)

Reactor Coolant Drain Tank (RCDT)

• MAB Sumps I thru 4 Containment Penetration Area Sump SIS/CSS Pump Compartment Sump AND

c.

ANY indication from Table C1 the Centnaii iment Challengc Table (see belew).

Table Cl: Containment Challenge 4aWle

.. CONTAINMENT CLOSURE not established within 30 minutes OR 83 IPag-e

_"(Eplcsiv-i, tc*.) %

-hydrogen exists inside containment OR UNPLANNED neFease*rise in containment pressure Seezfidaiy eentainmcnt r.adiatien mnfit. reading above,

( Qitz in P ifIMP v

1uz9[WR4 I if CONTAINMENT CLOSURE is r-e established p~

c fing 1he 30 minute timce limit, then dcclaratn of a General Emergency GENERA.L is not required.

Basis:

This IC addresses the inability to restore and maintain..ae......s.se.RCS 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 reacter vesse!4e! eeRCS level. If R.CS/reactzr vesel leveIRCS 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 GENERALGENERAL EMERGENCY is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the ether. li*ted indications in Table C l to assess whether or not containment is challenged.

SI--EAI-L-2:b l.a-7 the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitorrýreactor vessel/RCS t..WR].

rP.l !B"KR) 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 84IP a & e

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 (reactor vessel/RCS [R o-R, or RPV [BWRI).

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.

ICGI: EAL-1 Selection Basis:

C 1: EAL-1 Selection Basis:

The secondary indicators of inventory loss include a list of tanks/sumps found in OPOP04-RC-0003. Excessive RCS I ka~e Sinc'p other ~v~tPm lpnk'~ could rke l~vi~k in v2riAiu~ innI~ nnd th~ lid hn~ h~~n Iirnitpd to the tanks and sumps that would have the highest probability of indicating RCS leakage inside the Reactor Containment Building.

As RCS level drops the dose rates above the core will rise. Area Radiation Monitors RE-8055 and RE-8099 are located on the 68'-0" elevation of the reactor containment building. Their locations are identified on drawing 9C129A81105. Their range (0.1 mR/hr to 10,000 mR/hr) is identified in Table 12.3.4-1 of Section 12 of the UFSAR. Rises on these monitors can be can be an indication that core uncover is occurring. Additionally, erratic source range monitor indications, or large level rises in the tanks listed can give further indication of core uncovery.

The threshold value for radiation monitors RE-8055 and RE-8099 was based on Calculation STPNOCO13-CALC-006 Rev. 1. The calculated monitor response is 22.4 R/hr when RCS level is at the top of the active fuel and 6 R/hr at one foot above the ton of active fuel. The high ranue of these monitors is 10 R/hr. The value of 6

b v

c to

)

f ac iv f

e

.T h nz

--- n e

o f........

/......

9,000 mR/hr was selected for this threshold to ensure the threshold is readily assessable and within the calibrated range of the monitor. The threshold value of 9,000 mR/hr with the reactor head on corresponds to approximately 8 inches above the tov of the active fuel which provides an additional indication that RCS levels are near the noint of fuel uncoverx. These monitor readinius in coniunction with the other threshold values allow for an point of fuel uncoverv These moni r readings in conjunction with the o er threshold values allow for an accurate assessment of the EAL.

This EAL is similar to EAL-2 for IC CS 1 but also adds a parameter for challenging containment and potentially having a direct release path to the environment.

REFERENCES:

1.

Calculation No. STPNOCO 1 3-CALC-006 Rev. 1, Dose Rate Evaluation of Reactor Vessel Water Levels during Refueling for EAL Thresholds

2.

OPOP03-ZG-0009, Rev. 59, Mid-Loop Operations

3.

Drawing 9C129A81105, Rev. 3, Radiation Zones, Reactor Containment Building Plan at El. 68'-0"

4.

USFAR. Rev. 15, Chapter 12, Table 12.3.4-1, Area Radiation Monitors

5.

OPOP05-EO-E0 10. Rev. 21, Loss of Reactor or Secondary Coolant

6.

OPOP04-RC-0003, Rev. 18, Excessive RCS Leakage DeveloeF Notes; Acceidontli- -analyses suggest that fuel damage may eecur-within one hceur. of uneover-' depending upon the aunAR1t ofiftim sinzc shutdown; refer-to Gz*nc ie Letter 8.

17,*

SEC 91 283, NURET 1119 and NkARC 91 06.

85 1 P a g e

The type and range of RCS level instr~umnentto ma ay during an euawge as the plant moves thr-ough variouts opertin moes and refueling evolutions, -particular-ly for-a PWR. As appropriate to the plant design, alternate means of d-terfpming RCS l..

r.

IcR intailled tA a..ur. tat the ability to m..nitr level wit.hin th. e ranfige r.equir-ed-by operating proeeduwr-es will no;t be interrupted. The instrumenatmin range necessary to su~pport implementation ef opefetinig proeedurfes in the Cold Shutdown and Refuieling modes may be different (e.g., ntarrcwcr) than tha required dur-ing modes higher than Cold Shutdo-wn.

Forf A 0 lI-A T-h-e "site specific level" should be approxkimately the top of active ful f

tevailability of on+

seale level indieatien is such that this level vale carn be detennmined durinig some shutdo;; modes or conditions, but noet ethefs, then speei6, the mode dependent and/or configuration states during w-hich the level indicationi applicable. if the design and oper-ation of water level insitumentation is such that this level value cannot be detennined at any timne dur-ing Cold Shutdevff or-Refudelintg modes, thent do not include EAL lii (classificationl will be accomplished in accor~dancee with EAL ft2).

For EAL l42.b firlst bullet As water level inl the eaetr lvessel lowers, the dose rate above the core will incr-ease. Enter-a "site specific r-adiation monitor-" that coulid be used to detect cor-e uncover;y and the associated

" site specific value" indieative of cor-e uncover;. it is recognized that the condition descr-ibed by this lc may result in a r~adiation value beyond the operating or-display r-ange of the installed r-adiation monitor. in those easesý,

EAI. values should be determined with a mnargin suifficient to ensure that an accurfate monitor-readinfgis available. For example, an EAL moenlitor r

.eading might be set at 90-6 to 95-%,o

.f the highest accur.ate moenitor readinig. Thi s poiso nouvithstanding, if the estimatedcalculated monitor-reading is greater than approxy/imately 1107% of the highest accurate moitor reading, then developers may choose not to inlude th moonsifter Mas

-an-inid-ication-if -anda ideatifý' an alterna-te A threshold.

To fHther: promote accurate classification, developers should consider if some combination of mfonitorfs co-uldb specified in the EAL to build int ant atppropriate level of co*roborationi between monlitor. r eadings into the classification assessment.

For-BNVRs that do not have iist~dled radiation monitors capable of inidictn coeeuncover;, alternate site specific level inidications of cor~e uncover; should be used if available.

Feo EAL 42.b second ballet Post T-M! accident studies indicated that the inistalled PWR nuclear.

in.st.umentation will per-ate erratically when the

.core is uncover.ed and that this should be used as a toel fr.

making such determinationis. Because BWR Soce Range Monitor. (SRM) nu.lea. instrumentation detectors are typica*l loated below core mid plane, this may not be a viable i.die.A... f core unco.er;. for BnRs.

For-

• A*L.

2.hb th-ird-buillet Enter any "site specific sump aa&VV Ier k" levels that could be exipected to change if there were a loss of inventoy of sufficient magnitude to indicatcoe uncover;. Specific level values mafty be included if desired.

For-EAL 4f2.b fourth bullet De~veloper~s should determine if other-reliable inidicator-s exist to identify fuiel uneover-y (e.g. emt vieing using camneras). The goal is to identify any unique or site specific indications,no already used elsewh cc, that will prom*te timely and accrate emergency classificat For-the Containment C-hallenge Table:

Site shutdown otnec plans typically, prov-.ide for-Fe establishing CONT-AINMENT CLOSURE foallowina less of RCS heat r-emo-val or-iniventeory control functions.i 86IP a g e

Fer- "E~plesivc maigetur", de-.,elepefrs may efntef the minimfuma eentainmcent atmespherie hydrogen ecneentreAtizn neoessefy to suppei4 a hydrogen buma (ije., the lcmwc deflagration limnit). A coemuffent Cetamnntoygen cznccnf-tr:at-ionR mmay be ineluded ifthe plant h-az-th-is iniztjgAtiAn aval.AbicW in the Gontm] Reom.

Fer-BNVR9, the uise of seeendaf' eontainment ratdiattioni moneitors sheuld prov.ide inidicationt of incervascd r-elcuo that may be indieative of a ehallenge to seeendaf' containment. The "site speeifie valuce" shoulid be based efn the EPo mawimum safe valucs bozausc, thAso A'.alucs ar-A saivr i

abic n haWA A dfinzdP~ bAqis ECL Assienment Attributes-i 3.1.14.B 87 1 P a gze

8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

ICS/EALS Table E-1: Recognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E-HUI Damage to a loaded cask CONFINEMENT BOUNDARY.

Op. Modes: 44ALL 88 1 P a gz e

E-HUI ECL: Ntifiatin of Unusual E;cnt UNUSUAL EVENT Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability: AIMALL I E ni!mnpe-Emergency Action Level*:

(LD

_Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than: (2 ti...s the site p.. ifie eask

.p..

ifi, t..hni.al

.p..

ifi.atie allewabic

.adiaticn l.eel) ref the sufamma fe te srpent fuel task.

a.

60 mrem/hr (gamma + neutron) on the top surface of the spent fuel cask OR

b. 600 mrem/hr (gamma + neutron) on the side surface of the spent fuel cask OR

c.

7000 mrem/hr (zamma + neutron) on the side surface of the transfer cask.

Basis:

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 values for this EAL are 2 times the Technical Specification allowable radiation levels. The technical specification multiple of "2 times", which is also used in Recognition Category RAJIC AU-IRU 1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under ICs HU1 and HAl.

E-HUI1 - EAL-I Selection Basis:

NEI 99-01 Rev.6 states that the dose rate limits are 2 times the Cask Technical Specification Limits. Section 5.3.2 of the "Certificate of Compliance No. 1032, Appendix A, Technical Specifications For The HI-STORM FW MPC Storage System"' states:

89 1 P a gPe

5.3.4 Notwithstanding the limits established in Section 5.3.3, the measured dose rates on a loaded 0 VERPACK or TRANSFER CASK shall not exceed the followinga values:

a. 30 mrem/hr (gamma f neutron) on the top of the 0 VERPA CK

b. 300 mrem/hr (zamma f neutron) on the side of the OVERPACK.

excludingz inlet and outlet ducts

c. 3500 mrem/hr (gamma i neutron) on the side of the TRANSFER CASK

REFERENCES:

1.

Certificate of Compliance no. 1032, Appendix A, Technical Specifications For The HI-STORM FW MPC Storage System, Section 5.3, Radiation Protection Program. 10 CFR 72.104, Criteria For Radioactive Materials In Effluents And Direct Radiation From An ISFSI or MRS 90 1 P a g e

9 FISSION PRODUCT BARRIER ICS/EALS Table 9-F-I:

Recognition Category "F" Initiating Condition Matrix ALERT FAI An-yANY Loss or eayANY Potential Loss of either the Fuel Clad or RCS barrier.

Op. Modes. Power Operation, Hot Standby, Startup, Mgt S.4.dwn. 1,2,3,4 SITE AREA EMERGENCY FS1 Loss or Potential Loss of &nyANY two barriers.

Op. Modes: Poer Oiperation, Hgt Sandbýy, Start., Hot Sh..down 1,2,3,4 GENERAL EMERGENCY FG1 Loss of afyANY two barriers and Loss or Potential Loss of the third barrier.

Op. Modes: Power Operation, H*t Standby,

__ _ S.ta.up, HotSh*u.down 1,2,3,4 See Table 9 F 2 for-BWR E 'A Is Sec Table 9 F 3 fo PWR EALS Developer-Note: The adjacent legie flew diagram is fer use by Ide;'cloemrs and is not r-eauir-ed for site sneeific imnlementation:

however-, a site specific sehem-e, Must inclelude some type of user aid to faci litate timely and aceumtte elassifieationt of fissien producet barier losses a r potential losses. Sue i

r

.m.r.ised of logic flew diagrafms, "seeo... g" crliter-ia or k.h.eko typ marics.The user aid logie must be eansistent with that of the I

911Page

Developer oes 1 ", "

4

,1 t'

11

4. I n ici uc fAor ta-rnecinaie oaiin roPriocts MnR toliWAInoM nicr

° I

I 1

"1

.1

,11 Tnc Fuel Clad Riafrier and tne KG-fuaffier are Fegnc mr ncavily than the tontainmenit Ba~fiff.

J _ -.

L -

1 I

Gateger-y S.

~m ii..z

~ucia~e' im4 Fii~iOn rac anrsaepcrseni jcgi thcsols ill neoa-d-toA bc pcrformfed in nutIn wit-h dosA-qP -AmeSF~tS to eniSure corEct:P and timcily' cScalaAto ff thc c-Mergcncy clalssificeationl. FOr eXamfpic, anf eValution~f 8f the fiSSion prOdutl barrier-thrcshklds may fcsutlt in a Site Arca Emergency classification while a dose assessment may in-h~iPt" thA*An AT.

n inZ41FmI~nn-'I AC-h"1 "n'-ci-1

".1

-4 i nc tirsiefi or-eai]t thnfr-~ thrcstinkR sneetfie within a seheme amf exneete te rctlcct Wannt sneetc

-I-------14e~

o

°

,° ml *

,1 1

I acsiirn ana oncratina enaractcristic~ i nis may rc~1rnr2 tnat acvcicncr' crzats oittcrznt mrzsnoias d*.......

r

• 1

.1 T 1 J

T toan tnose pro;~aca in tnc generic guiaance.

4.

'A 4oaiv prsnAtiON mithodIS for the Reeognition Category F W~s andl fission pr-oducat barrier-thrcsholds aprc accAhepb anid includc flow charts, block diagrams, and ceceklist type tables.

Devedoper-s must cnsurc that the site specific mcthed addresses all pessibic thrcshold combinations and classification etutcomcs shown in the BWR or PWR EAL fission proeduct barrficr tables. The N1RC--

st-aff considers9 the

aenttonmthod of the Rccoegnitiot Category F ifrtinto be an impotant uiser-aid and may rcqucst a changt partclar proposcd mcethod if-, amoeng othc rcasens, the changc is necessar-y to proemote conisistency across the inidutryfý.

5.

As used int this Rccogeition Category, the term RCS lcakagc encopaee no ust these f"~c dlefinled in,; Teehmieal SpeceifficatFs-ioen;s bu it also inceludes the leoss,- A off -RCS mass to any locatesion inisid coent-ainmffent, a secondar~y side system (i.e., PWR steam gcncr-ater-tubc ckgca ifitcrfaiing system, or-outsidc of containmcnt. The fclcasc of liquid or-steam mass ffom the RCS due to the as desiencd/expected aoncrationli ofa cif vl is-4nopt csdrdto be, RC-S le.Aiagio.

cl K~

`At thn '-ite A-ea Lmer4p'cncv level'c enaiticntm ;ea ie~sn fnnI'cr- 'sh'ui'd mnantni X'-nýfccs v.

how farf proenet cVonditions arce from m~eetinig a thfeShOld that would require a Genleral Emefr-gency declarationH. For cxamffplc, if thcit Fuci1-A Cla4d and RCS fissionl pf:droduct ba

'er ere both lo)st, t-hen thoro-e-should be frcqucnt assessments of containment r-adioaetive iniventory and integrity. Altematively, if beth the Fuel Clad and RCS fission pr-oduct baffiers wcr-e potentially lost, the Emor~gencey Diretor would haVc morfe ft~iass ranec-th4A-t t-herc4-asn cdiAtA need to csAqAtc to A Genral EImcrgeney.

-,1

• 1

~

1 -

7 1

I

ý

. I I

i nc apiur; to escalate to a nwncr cmcrc~

,..ia.~.,incarIon icvci in rcsnonsc to acaraamg conaitions t1I 1---~J r~------------

lSC~C~ 1~ ~1 lJ

,fl~JtiU Ve mafffffintaet. rorf oxampfivl, a steaev morcf~ase in RC icaKaee weul eia rouesentt ant incesn 1

f 921 P a g e

Table 9 IF 2:, BWAR EPAL FiSSiOn Product Barricr Table Fl~iAl.FAI SITFE AMREA-EMR_=_RGCENC FG1 GENERAL EMERGENCY4 A"

I

ýg II

• -tflIt.I T

qII

a.

LOSS or: Potential IoS of afly t;AbA hifrfpr

b.

LOSS Of any tWO baffler-m n

- C~l third bamien eFel Clad B.rrier

d. RAS Bo.i.*r

e. Containment BaAMrer f-4LOSS g........h.

LOSS

i.

rOTn LOSS POTENTIAL PALt PAL LOSS LOSS LOSS 4

fPims"

^

Containment Radiati

4. rPrimaF y Containment Radiation

k. 4. Prime",Containment Radiation A....................

Not A.................

No

l.

notn t

A. Pimac t

radiation monitor Applieab radFiation monitor.

App.a Applieanb r

monitor rcading groater than; 4e roadinig greater-than Wle e

reading greater-than (site speEifi v.aluc).

.(site sp.i Ea c.

(site speific valu).

e. 5. Other-ndicnti!ns

p.

54. Otherf indie-tionsg

5. Othctn dications A. (site specific as A. (site speia A. (site speia A. (site speei A. (site speia A. (site specific as appheabýe I

phab pplieable)

I vpkwable appie

~

~ppfieable r-. 6. Emner-gency Director judgment

s. 6. Emner-gcney Director judgment

t.

6. Emcrgcncy Direetor-judgment A. ANY condition in A. ANY condition in the A. ANY conidition in the A. ANY condition in A. ANY eondition in the A. ANY condition in theoplfienion fhe opinionefthe Opinionof te te p~iein 04b Opinion f~b tp t

heOpno Emoregency Dircotor Emfergefncy DircotMr smer-geney Dircotor Emer-geney Dircotors Emarfgeney Diroctor Emcrfgency Dir-ctor.

that indieates Less thatifidieates that ifidieate Les e that ndieates that inidicatcs Less of theAtifidieetes of the Fuel CladA se Pot ia LoA II.;Af theLA the RCS Barrfr Potential Loss of thc the Containmen Potential Less of the uIýml Cland BARri RGS BafFiein BeFfi~e4FCnaimn Br I

BHwiS !InFomation FOr milD E'ITVI £'I Ark DADDIVD TIJDVC~I1C~I 1W.

I r'~ rii'-I I Ifil nrrncr consists ci tnc zircanov or stain:ess steel luci cunaic WDCS mat contain tnc luci PelletS.

R.

Aci vAit beTg&4-A This thr-eshold inidicates that RCS radieaetP't enntatio is greater-than 300 tt~i/gm dose equiv.alent 1 13 1. Reaetor coolant activity above this level is grcatcr than that ex~pected for iodine spikes and corresponds to anapoiae range of 21%4 to 5% -fu-el clad damage. Since this conidition in-dicates that a significant amount of fiael clad damage has occurred, it represenits a less ef the Fuel Clad Barriefr.

Theft is no Potential Less thr-esheld associated with RCS Aetivity.

Develper-Notes, Thrcsbold values should be detefminted assuming RCS r-adioactivity eoncentration equals 300 ttCi/gm desc equivalent 1 13 1. Other-site spccifie units may be used (e.g., #Gii/ee.

Depending upo sit spific capabilities, this threshold may have a sample analysis component an4'or-a madiatio moitor-r-eadfing component.

Add this paragraph (er similar. Wording) to the Basis if the threshold inceludes a sample analysis component, "it is-recognized that sample eolleetieni -and-anialysis of r-eactor. coolant with highly elevated activity levels couild rcq~uirce scver-l heur-9 to completc. Nefnctheless, a sample rcelatcd thfeshold is ifieludcd as a backup to other-indications."

2.

R-PM Water: Level The Less threshold represents the EOP rcequircrncnt for prmayctinment flooding. This is identified in the B"OG EP s/

N hen the phrase, "Pr-imary Cont-ainment -Flooding is Requir-ed," appears. Singe a site specific RPV water level is noet specified her~e, the Loss throshold phfasc, "Primary conitainment flooding requfired," also acconmmodates tho EOP need to flood the pr-imar ct imct whcn RPM* watcr

level cannot be determninod and corc damage duet iniadequate coe tolnsble edt be occuring Potential Loss 2 -A This water level corresponds to the top of the activc fuel and is used in the EOPs to indicate a ehailefigete er-e eling.

The RP7 waAter level thrcsqhold i;sthe same asq RCS barrier LoAss thr-eshold 2.A. Thus, this threashold indicates A Potenitial Loss of the Fuel Clad barrier-and a Loss of the RCS ba~r-er that approepr-iately escalates the emergencey classification level to a Site Afea Emergcncy.

This thfeshold is consider-ed to be exceeded when, as specified in the site specific EONs, RPV. water eamet be r-estoede and maintained above the specified level following depr-essurization of the RPMI (either manually, automatically or-by failureeoftheR RSbaricre) or-hen proceedural guidanceeor-a lack of low pressu~e RPM injection sources preclude Emer~gency RPM depressurizationi. EONs allow the operator-a wide choice of "VM injection sources to consider-when r-estoring RPMZ water level to withn pr~escr-ibed limaits. EOPs also specify depr-essur~ization of the RP' in order to facilitate

_RPM w.ffater. level control w.ith lowpesr injectio sorces. Inhom events, elevated RPMP pressure may pr-event rcpswstomtno ACRPMAW water level til preassure droeps below the shutoff heads of availatble injetion souFres. Thferefr-e, this Fuel Clad baffier Potcntial Less is met only after either: 1) the R.PM has been depressur-iied, or requtired emnergencey RPM depreSSurizationH has been-4 -attempted, giving the operator anopruiy to assess the capability of low pr-essure injection sources to r~estor-e RPM water level or-2) noe low pressure RPMI injeto Fytmse available, pr-ecluding RPM depressuriziation in an attemp to Fiie loss of RPM inventor-y.

The tenn "cannoqt be restored anod maintained above" meains the value o RPMRI wateir. level is not abl to be bogtave the specified limit (top of active fuel). The determinatioreursa evaluation of system per-forancee and availability in r-elation to the "NM water: level value and Ifead. A threshold prescr-ibing declar-ation when a threshold value caimna1 be reAstorfed0 anId-maint-ainedA above-a specified limit does noa eur mmdaeato simply because the current value1 is below the top of active fuael, but does noet permit extended oper-ation below the limit; the threshold must be considee r-eached as son asi ws paent that the top of active fuel cannoit be atuaified.

in high power-ATqWSF/l'fA_;i_;ure to semam events, EOPs may dir-ect the operator to deliberately lowýer R-PM. wate-r le-vel to the top of active fuel in or~der-to reduee r-eactor power. RPM water: level is then eentfelled between the top of active feel and 94 1 P a p e

the Minimum Steafm-Coo-ling RPNV Water-Level (N4SCRWL). Although such action is a ehallenge to er-eeling and the Fuel Clad bafier-, the iwmmediate need to reduce r~eactor power-is the higher-priority. PFr sueh events, W~s SaA-5 Afr 8S95 400ll dictatc the need fcr: emer-gency classification Since the lcess of ability to d-ete-rmine i-fadequate eor-e eooling is being provided presents a significant challenge to the fuel clad bafnier, a potential loss f thc fuwl clad ba~rrier-is speeified.

DevepeF oesi The phrase, "Primary containmcnt floodinig rcquifed," sheuld be moedified te agree with the site specific EOP phrase indieating exit from all EOPs and enti-,~ to the SAGs (e.g., dfywell floodinig roquir-ed, etc.).

Potential Loss 2.-A The decision that "RPV water level cannot be dctcnnined" is dir-ectcd by guidanec given int the RPV water-level contrle sections of he EOPs.

3.Not Applicable (incluided for numbering consistency bc~eohnp baffler; tables)

The r-adiatien menitor-reading cofrresponds to an instantaeeus release of all reactor-eoolant mass inito the pr-imoa conai~ntassuming that r-eactor: colant activity equals 300 PI&Qgm doesc equivalent 1 13 1. Reactor coolant activity abovti leve~l is gfeeter than that expected for iodine spikes and corresponds to an aprmmt agce ef 2% te 5-, fiiel elad damage. Sinee this conditiont indicatcs that a significant amoiunt cf fucel clad damage has occurrfed, it Fcpr~escnts a loss of the Fuel Clad Barrier:.

The radiation monitor reading in this thrcshold is higher thant that spccificd for RCGS Barrier Loss thrcshold 4.A since it indicates. a loss.. ofboth the Fuel Clad Barrier and the RCS Barrier. Note that a combiniation of the two moneitor readings-a"pprpiately cseal-atess th-R e mergencey classificeation level to A-Site A-rea-Em~ergeney.

There is no Potential Less thrcshold associated with Primary Containment Radiation.z DevelopeF Noesis The rcadinig should be dcterminced assuming the instantaneous release and dispersal of the rceaeter coolant noble gas and iodinc inventor-y, With RCS radioactivity eetenifnatien equal to 300 tt~i/m ds equivalentl1 13 1, in to the pr-imr centaifmnent atmoespher-e.

5. Other-indicationis Loss an4'or-Potential Less 5. A This subeategory addes es te site specifie thresholds that may be included to inidicate loss Or potenltial loss f the Fuel Clad barrier-based eft plant specific design characteristics not conisidered in the genieric guidance.

DeveopeF Notesi Loss and/er-Potential Less 5.A Develope~s should determinte if other r~eliable ifidieateo: exist to evalutte the statuis of this fissien producet bamFer (e-.g-.,

review accident analyses de-scribepd in th-e site Final S-afety Analysis Repeor, as updated). The goal is to identify any unique or-site specific-inAdicaltionRs that will promote timely and accurate assessmaent of barrier: stattas.

Affi) added thresholds should represent approximoately, the same r-elativ~e threat to the bafrrier-as the other thresholds in this column. Basis in-formationm Ifor thee other thrfesholds may be used to gauge the r-elative ba~rrier-threat level.

6. Emer-gency Dir-ector judgment Leess 6A This thre shold addresses any other: factors that are to be used by the Emer-gency Diirector in d-eterfmining whether. the Fuel Clad Baonier: is lost.

Potential Loss 6.

This threshold-ad-d-resses; any other factor-s that may be used by the Emer-gency Dir-ectorF in d-eterm-mining whether the Fuelt Clad BaiFer-is potentially lost. The Emergency Director-should also consider whether or not to declar-e the ba~r-i potentially lost in the event that bm:Fer status cannot be montitor-ed.

DevelpeF Notes-:

9511 P a R e

None The RGS Barrier-is the r-eaetor-coolant system pr~essurfe boundary and includes the RPM and all roaeter eoolant system pipinguptoand ineluding the isolation valves.

• Primarey ConAm-inmcn Pessur be&94-A The (site specifie Yaluce) primr cntanent pressure is the drywell high pr-essurfe setpoint whieh inidicates a LOCA by autematieally initiating the EGGS or equivalent makeup system.

Thero is no Potential Less throshold asseeiated with Primnary Containment Pr-essir-e.-

Developer-Notes.

Nene LRAI Water Level This water level eerrcspcnds to the top of active fuel and is used in the EOPs to inidicate challenge to cor-e oolinig-.

The pV RP

,water lc-vcl thrcshold is the samc as Fuce Clad barrier Potential Loss threshold 2.A. Thus, this throshold indiatgosp a1 Lo of; Athg RCS hilfcr. And Potential LoAss of tho Fuel Clad barrier and that appr-opriately escalates th emer-gency classification leyel to a Site ~AJe Emcr~gency.

This thrceshold is eonsider-ed to be cxccceded when, as speeified in the site speeific EOPs, RPM water cannolt be restoroed

-Fned-miliaint-ainofid above the speeified level fellowing depr-essurization Of the RP V (eith fr manulally', automiaticallyor by' failuroe of the RCS barrier) or-when pr-eeedural guidanec or-a lack of lcw proessure RPV injootioni sourocs pr-eelud Emer-geney RPM depr-essurization EOPs allow the operator-a wide ehoiee of RPV injoctioni sourcs to eeonsider-when rostoring RPX' water-level to withint proescribed limnits. EOPs also speci6y depr-essurization of the RPV int order: to faeilitat R-PM water-level eontrol with low prcssur inato oroces. int seme events, elevated RPM prcssurc m~ay proent rpestomtfi-ion of PM ator. level until pressure drops belowmi the shlutoff9 heads of available injetinl -sourcoe-s. Th-erefore, this RCSF barripr. Loss is met on[' after-either:. 1) the RPM has boon depr-essurized, Or roqutired emergeney RPMZ depressurization has beon attempted, giving the opcr-ator anl opportunity to assess the eapability of low pr-essrinotn sources to restor-e -RPMl~ woat-e-r level or-2) no low pr-essur-e RPM injection systems are available, precluding RPM deprossurization in an aamp tomniie less of RPM inventor-y.

The tcnn, "cannot bo rostorod And-m-ainftained above," moans the value of RPM water-level is noet able to bo broeught abovo, the specified limit (top of active foe!). The d-eterminAmatsio reui-n evalufation f Osyqtefm porfrmnanpe And availability in rolation to the RPM water-lovl valuce and nd. A threshold proscribing declar-atiotn when a threshold value.am!

be rostor-ed and maintained aboew a specified limit does noa our immdiate actiont simply beeause the eurront valuce is belo%, the top of active fuel, but does not pefmit extended operation beyond the limait; the thfeshol d-must be considerwed roachod as soon asiis paet that the top of aetix'c fficl cannot be attained.

int high power AT-WS/faiiurc to ser-am events, EOPs may dircet the operator to deliberately lower RPM w-ater level to the top of aetive fuel in or-der to reduee roactor-power. R!PM water level is thefn controlled botween the top of active fuel and the Mintimum Steam Cooling RPM Water Level (M4SCRWL). Although suceh action is a ehallefnge to eor-e cooling and the Ful-0 Claid barier., the ifmmediate neepd to reduce riector-power-is the higher priority. F-Or such events, W~s SA5 or SS5 will ddict-ate the need for. emaergency classification There is no RCS Potential Loss threshold associated mith RPM Water-Level.

3.

RCS Leak Rate Loss Thr~eshold 3.A Lar-ge high energy linoes that rupur-e outside pr-imnay eontainment can dischar-ge significant amounts of inventorty and jeepardiize the pressurie retainling capability of the RCS tmtil they arc isolated. Wfit iss ddeterfmined that the ruptur-ed line catnnot be promffptly isolated from the Control Room, the RCS bafdFef Less throshold is met.

Loss Threshold 3.B Emergency RPM Depr-essuriz-ation in-accor0dance. With t-heEPs isincaveoa loss of the RCS barrier. If Emergency RPM Deprossuri.tin is. peformed, the plant opefrator-s arc direeted to opent safety r-elief valves (SR~s) and keep them occa. Event thouueh the RCGS is being vented into the supr-ession peel. a Loss of the RCS barrier-exists duo to th I

dimfinfished etizgetiveness of the RC-S to r-etainl fssieon proeducts; within its bounfdar;.

Po1tntial Loss T-hroshold 3.A 96 1 P a p e

PotPntial less of RCS based n pimat Item leakage outside the primary containeat is determined from EOP tefmpefrafe Or radiationk Mwi Normal Operating salucs in areas sueh as mraini stemam linie tunnol, RCOC, H4PCI, ete., whieh indicate a direct path ffom the RCS to areas eutside pimr co tainent.

A MA,4x* leNormnal Operating value is the highest value of the identified parameter expecete-d tcA occeur during normal plan!

Operatinig 8eondit:AIonsq A0ith all1 dircctly aSSEoeiatcj SUPPOrt andJ conitrOl Sy'Stcs ftmctiening properly.

The indiceatrs r-eaehing the throsheld barriers and confirmed to be eaused by RCS leakage fromf a rmr ytn arant an Alert classification. A primr stmis defined to be the pipes, valves, and other equipment wh~ieh connect direetly to the RPV sueh that a rcduction in RPV prcssurc will effect a decrcasc in the steamn or water being discharged through an tmiselated break in the systerm.

.4an NSAL leafk-wlhich is indicated by M~a* Nonnal Operating values esealates to a Site Area Emerfgenc~y when combined with Containment Barrier Loss threshold 3.A (after a eefntaiffnmct isolation) and a General Ermergency whern the Fuel Clad Barrier criteria is also exeeeded.

DeVel~peF Note.

Less T-hrshold 3.A The list of systems inceluded int this threshold sheuld be the high encrgy linies which, if rudptured an;d-rcmftain uisolated, c-an rapidly depreswurizc the RPY. These linies arc typieally isolated by actuation ef the Leak Dctcctien systerm.

Large high energy line breaks suceh as Main Steam Linte (MSL), High Pressure Coolant Injection (MPG!), Fccdwatcr, Rcacter Water C-lcanup (RWC-U), isclatiefn Coidcntscr (IC) or Rcactor Gore Iselation Cooling (RCIC) that arc UNISOLABLE. r-cpresert a significanit lass; -f thc -RC-S barrier.

The madiatien moinitor reading eerrcespendS to ani insitanltaneous release of all rceaetor coolant mrass intoe the primariy continmntassuming that reactor coolant activity equals Tccehnical Specification allowable limits. This -value is lower thin ta p

Id -foer-Fu.

C lad 1

- Ba rri*e r L-o-1 thr;eshoId-

..A sin e it indicates a loss of thea RCS Ba

....er only.

Trcis'- noA --PotetialJ Los hreshoAld 40scit4dwih Pr-imary Coentaiinment Radiation.f Develope-r Note:

The reading Shoul1 1d-heA deterinfied assuming the instantaneous release and dispersal o-f-the re-actor coolant nobeh1 gas and iodine inivenitry, with RCS activity at T-echnical Specificationt allowablc limits, into the primfar enanct atmosphere.

Using RCS aetivity at Tcchnical Spccificatien allowable limnits aligns this thrcshold with 1C SU3. Also, RCS activi~i at this lcvcl wi....ll typially result in prfimryý m..

raddi ation Rem.

tha can ib m.r.. roadily detected by primary conitainmenlt radiation monitors, and morie r-eadily, differentiated ferom these caused by pip ing or copncn "shine" sourccs. if desired, a plant may uisc a lesser value of RC S activity for dteormining this value.

lIn som-e cases, the site specific physical loceation -and sensitivib, of the primary containment radiation monitor(s) may be such that radiation fomA a cloud of.

.l.as.d RS gases earnnt be distinguished from radiation emanating from piping and 1om.pon.ents cntaining elevated rca.tor oeolant activity. if so, refer to the Dcv.l.p.r Guidanec for LossPefottial Loss 5.A and determine if an altenate indication is available.

5.

Other indications Losand4tor Potential Loss,ý A This subeategory addresses other site specific threshelds that may be includcRd to indicate loss or potential less of the RCS barrier based on plant specifie design character-istics not considered in the gencric guidance.

Dev.elopeF Notes:

Less and/er Potcntial Loss 5..A Dcvcelopcrs shoulid dctermninc if ether reliable inidicators exist to evalueat the status of this fission product bafcr-i(c.g-.,

rcvuiew aceidcat analyses described in; thc sitc Final Safety Analysis Rcpert, as updated). The goal is to idcnftify, any unique or-site spccific indicationts that will promoite timely and accuratc assesssmont of barFicr-status.

Any added thresholds should represent appreximatcly the same r-elative threat to the barrier as the ether thesholds in this column. Basis information for the ot1her. thr-esholds may be used to gauge the r-el-ative barriter threat leel 6.

.e*eeA

.r=

u tla

+

lI III

+*I

]+VVII*ý]*++iI**

+

• +

+

++

+ ++.~£ V

+ +I+

+ +

97 1 P a m e

This thfeshold addresses any ether-factors that are to be used by the Emcr-gcney Dirccter-in detcfrmining whcther the RCS Potential Loss&-"

ThiS threShold addresses any, other-factor-s that may be used by the Emerfgenc~y Director-in detennining whether-the RCS Ba&cr-e is potentially lost. The Emcrgene)y D9ireter. sh-oulId False emcnsidcrm A6h;etherf er. not to deelarc the bar-ieF potentially lost in the event that bafrrier staus eannet be monitored.

Dveloper-Notesi Noeii The Pr-imary Containmcnt Barr-ier-ineludes the dry-w li, the wetwell, thcir r-espective interconnecting paths, and ether 0o0 ctos pt and including the outennost containment isolation valves. Contaimment Barrier-thr-esholids are:p uwsed a cierwiafor escalaftion of the ECL from Alet4 to a Siteo Arca Emergeney or a General Emcrfgency.

1.

Pr-imar-y Containment Conditions9 Lo-A-Il.A-m a- -1. B Rapid UNPLANNED loss of pr-imary cotanmn prsturc (i.e., noet attributablc to drywcll spr-ay or-eendcnsation cffects) following an initial prcesstirc incrcasc inidicatcs a less of primfaryJ cotanmnt itgity. Primary contaimcent prcsr shculd inocas FasI1 a cs fmass and cncrfgy release into the pr~imary eontainment ro as LOQC-A. Thus, pr-imfary conaimn pesuro not increasing under these coniditions indicatcs a less of pr-imar-y containmcnt intcrty.

Thcse thcsodsrly on epcratrrcontn of an tmcxpcctcd rcspensc for the coniditien and thcrcfefrc a specific value is noet assigned. The uncxpcctcdl (tNPLANNED) rcSPORs is imprtn bcas t isAh indicator. fir mctffifnmcnt b5-pass eendifieft.

Potential Loss l.A The thrcshold prceSSUrc is the primary containmcnt internal design prceSSUrc. Strutuaeifl acceptance testing dcmonestratcs the capability of the primnary contaifnment to citprsue greater-than the internal deRsignprcssurc. A prcessure of ti magnitude is greater than these cxpccted to rsult from anmy design basis accident aRndI thus, repr-esent aR Potential Loss;-. of the-Cont-ainmfenit badfer..

Potential Loss I.

if hydrogen concent*Ratioreace orf exceeds. the lower-flammffability limit, as defined in plant EOPs, in an oxygen rich cnvromcta potcntially cXplesivc mnixtur eitists. if the combustiblc mnixture ignites inside the primnary containmcnft, los oef the Conta~inment bafier-could oecur.

Potcntial Loss I1.C The Heat Capacity T-empcrature Limit (4C-TL) is the highest supprcssion pool tcmper-atur fromn which EmcrgcnieyRP Dcprcssur-iai%.

will noet raisc:

"Suppr-ession ch-amberf temper-ature above the Maximum temperature capability of the suppr-ession chVhambranld eqipen ithin the suippr-essiont chamaber-which may be rcqutircd to opcratm when the RPV is prcesstr-izcd, OR "Suppression chamber: pressurel Rhabovep Pr-imary Contain-m-ent PrsueLimit A, while the rate of ener-gy transfer from the RPV to the containment is grcatcr than the capacity of the containment vnmt.

The HCGTL1 is a frnctien of R!P3, pressure, Eupcsin ol tcmpcratur-e and suippr-essiont pool water-level. it is uitilized to preclude failure of he cona~infment and equi'fipent in the-conaMAinmenfft neceOssary for t-he safe;huHtdoVn Of the plant and thcrc@ferc, the inlabilit to mafintafin plant parFFAmc9tcrfs bclcw-.A the limit Aosiuc a potni-al loss ofeefitaipmcnt.

De~epeF oefr Potcntial Loss 1.

BWR EPGs/SAGs spccifically, dcfince the limnits associated with explosive mixtures in tcnns of dcflagmtion eonccentr-atiens of hydrogcn and wxygcn. For. 44k L1'1I enamcts hck dctfagfRationt limfiit's aFrc "6%04 hydrogen and 5% Oxygcnff in th dr"'cll or-supprcssien chamber". For-MkE ill eentainmcnts, the limnit is the "H4ydrogen Deflagration Over-rcesSUrc Limi" The thrcsheld tcrm "cxiplesivc mnixturce" isinn~u with the EPG/SAG "deflagrationt limits".

Potential Loss 1.C Sincc the H4C-TL is defined aswig age of suippression pool water-levels as low as the clevation of the dewcoe epeffifgs in Mk 1/11 containmcnts, or 2 feet abovc the clcvation of the horizontal vcnts in a NNE ill eentainmn.tI it is unncccssmry to consider scp-arate Conl-iAMenA-t 1-09ic 4

Losor PROSA1tc-Al L-oss thresholds for-Abncna suppressiont Pool 98.1 P a g e

water level conditions. if dcsir.d, developers may inlude a sepa.ate C.ntainment Potential Less thr.esh.ld based on the inability to maitn uprSsiOn fl POIwater level1 abov'e thP dcwncomerF opnig Min 141k AAiotainen~ftS, Or 2 feet above the elevation of the herizontal vents in a NQ II on1 mnn with RPV prcssurce above the minifmum decay heat r~emeval pr-essure, if it will simplify the assessment of the suppression pool level component of the "C-3.

T-hcre is no Less throeshold associated with RPVX Water-Level.

Potential Loss 2.A The Potential Loss thrcsheld is identical to the Fuel Clad Less RAI Waere Level threshold 2.A. The Potential Ls rcquir-ement fcr-Pr-imary Containment Flooding inidicates adequate corce cooling eannot be restefrcd and maintained and that core damage is possible. BWR EP~s/SAGs specify the coAnditions that require pr-imary containment flooding. When primay cotainn-icn flooding is rcquir-ed, the EPGs are exited and SAGs arc entremd. Entr into SAGS is a logiea esealation in

.espens^

to the inability to r.storce and maintain adcquatc corc cooling. PRA studies indicatc that eondition cof this Potential Loss threshold could be a coere melt sequnene which, if not coffected, couild lead to PvV failuree and ifiercascd potential fcr pr-imr co ta ent failure. In cojnto ith the RPV water-level Loss thrcsholds in th Fuel Clad and RCS baffier-eeltifs, this theshold rosults in the deelaration of a General Emerge-ncy.

DeelpeF N4te§ý The phrase, "Primary containmcnt floodinig roquiroed," should be moedified to Weec wit the site spccific EOP phrs indicating xidt fromn all EOPs and efntry to the SAGs (e.g., drywell flooding roequir-ed, ete.).

Thcs theshlds ddrss ncomlct cotainmcnt isolation that allow.s an UNISOLABLE dir-eet rclcasc te thc The utsc of the modifier- "dir-eeCt in defining the rclcasc path discr-imintates against rclcase paths throuigh infteffacing liquid sysemsor inor-release pathways, sueh as inastrument lines, not protected by the Primary Containment Isolation System The existonee-e o-f a filter is net conisider-ed in the threshold assessment. Filters do not remove fission producet noble gases.

in addition, a filter-ceuld become ineect

• iedueto iodine an*do*rpaticulate loading beyond design limits (i.e., retenti ability has been exceeded) or-water satur-ation from steaw~igh humHidity in the r~elease stream.

Following the leakage of RCS mass into rmr co im-ent And ai Arie inpiaycnanetpessuire, there mnay be m~inor ratdiological releases associated wAith; allowable pr-imar coinet leakhage through various penerations or system components. Minofrreleases may also occur-if a prmar co ainet isolation valve(s) fails to close but the pffmary contiainiment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential less of pf~fir-my conaimetlutshoulld be evaluated using the Recogaition Categor~y A IC-,.,

EOPs may dirc pr)'cntainmnent isolation v~alve logic(s) to be initentionially bypassed, even if offiite r-adieaetivity release rate limits will be exeded. Under these conditions with a valid primnary containment isolation signal,th containment should also be consider-ed lost if pr~imar-y containmffent venting is actually per-femed.

Inentionlal ventinlg opif piar Fotinet for prim~ary con~talffinent pressure Or combustible gas con-;trol to the secondar conainmenmt and/or the eniviropnment is a Loss of tP he Containmm ent. Venting for priay ctinmn prsueeAtrl* when noet int an accident situation (e.g., to controel pressufe below the dr-ywell high pressure scaFpint) does not meet the bess-3ý The Ma. Safe Operating Temperature and the Max Safe OpcrAting Radiation Level are each the highest value of these paramffeter-s at whl'ich neither:i (1) equtipment ntecessar-y for-the safe shutdown of the plant will fatil, nOr (2) per-sonnelaccs neessaf' for-the safe shutdown of the plant will be pr~ecluded. EOPs uitilize these temper-atures and radiation levels to establish conditions under-which RPV dep uiain is required.

The temper-aturwes anid radiatio levels shul beconr4;med to be caused by RCS leakage fromn a primary system. A primry ystm is defined to be the pipes, valves, and other-equipment which eeofflet directly to the RPV such ta r-edutiotin int RA' pressure will effect a decr-ease in the steam or-water-being dischar-ged throeugh an uinisolated break in the in combination with RCS potential less 3.A this threshold wouild result in a Site Area Emnergency.-

99 1 P a g e

J7 I t

,l 1

1 1 7

• I

,l

?

I If*

1 nlere !S HlO raiftin Deveoper-Notes; ai LOSS inr~esnold assoeeatca with rnmna~ry Lonftanment Iseleait ionraurte.

Consideration may be given to speeifying the Specific proeedural Step Within the Prfimary COntainmfent Control EOP that de-finesq inften-tiOnal Venting of the Primar-y Cent-Rainment r-egardless of effsitc radi@actiVity rclcaiSc raF I.Pr-imnar Contaifnment Radiatien Themc is no Loss threshold associated with Pr-imary Containment Radiation.

Petcntial Loss&lA-The radiation monitor-reading corresponds to an instman~eeos release of all r-eaetor eoolant mass inite the primary cotanenassuming that 220% -f-the -fu-el cladding h-as failed. This level of fuel calad failurlie is well above that used to dter'm-ine. the analogous Fuel Clad Barrier Loss and RCS Bafrrier Less thr~esholds.

NUREG 1228, Somrme Estmatiens During Inientm Response toSýerNclaPwr PlanAeeiden.'s, indieatcs the fuel cl-ad fa-iire-must be gr-eater than approeximately 20,% in or;der. -foer there to be a major. releamse o-f r-adioactivity rce ui i g offsite proetetiv aetiens. For-this eondition to exist, there must alr-eady have been a loss of the RCS Bafier-and the Fuel Clad B3aFer-. it is ther-efore ppmdent to treat this conditiont as a potential less of contaiftment which would then escalate the emergeney classification le-vol to a Genieral Emnergency.

DcVclOpeF NOteS:

NJUREG 1228, Seurce Estimations Duwrng Incident Respeov-se to Severe NuclPIe-ar Peower Plant Acidents, provides the basis for-using the 20% fule! cladding failure value. Unless there is a site specific aan~ysis justifying a different value, the reading shouild be deteffined assuming the instanitaneouis release and dispef~al of the r~eaetor coolant noeble gas and iodine inventor; associated with -20% fuel clad failure into the

,fma;ot inet atmoesufhere.

5-.

0 ;therl1A Ind icP at6fIonS T

-i Al-.,

flT.;^ i; I

LOasS antger r aten ati Iass -

J. -

This subeategor-y addresses other-site specviffic thr-esholds that may be inceluded to indicate loss or-potential loss of the Cont-ain-menot barrier. based on plant specific design cha-r-acteristic-s noot cosdrdin the generic guidanee.

DeeopeF Notes.

Los anor otetia Los 5A,;

Developers should detennine if ether-reliable inidicators exist to evaluate the status of this fission proeductbnereg.

r-eview accident anfalyses described int the site Final Safety Analysis Repeot, as updated). The goal is to identify any migue or: site specific indlications; that will promote timely and accuraite assessmffent of PbonFier status.

Any added thfeshelds should represent approximately the same r-elative threat to the baF~ie as the other-dweshelds in this eeolefm. Basis infonnationt for-the other thr~esholds may be used to gauge the relative baF~ie threat level.

E. mergency Director-judgment Less 6.A This threshold addresses any ether-faetors that arc to be used by the Emer-gency Dir-eetr in detefnining whether-the Containment barrier is lost.

Potential Loss-".

This threshold adese any ot-her-factors that may' be used by the Emergency Director in detonnining whether-th Conauinment Ba~cr-i is potentially lost. The Emergencey Direetor-should -also cons1ider-Mwhete or-not to declareAd; the A~

petentialy test in the event athe barer-status cannot be moniter-ea.

Developer Notes:

None 100 1Paue

I Table 9-F-2.: P-WR-EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAl ALERT FS1 SITE AREA EMERGENCY FGI GENERAL EMERGENCY A-yANY Loss or ayANY Potential Loss of either Loss or Potential Loss of a*yANY two Loss of aayANY two barriers and Loss or the Fuel Clad or RCS barrier.

barriers.

Potential Loss of the third 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. RCS/reactir vev, e A. An automatic or A. Operation of a standby A. A leaking or Not Applicable 4v[4 w; than (site manual ECCS (SI) charging fm.kep.*

RUPTURED SG speei-f.eleve4-Core actuation is required pump is required by is FAULTED Cooling - Orange entry by EITHER of the EITHER of the outside of conditions met following:

following:

containment.

ti,_UNISOLABLE

1. UNISOLABLE RCS leakage RCS leakage V.

OR OR

-,_SG tube

2. SG tube leakage.

RUPTURE.

OR

• .B.

RCS cocld'wn raitc greatcr thanl (Site

..peeifie przzzur-izcd the~ffia-sheek Aritf-Win'liMik def~ined by ie speeifie idieatien,*

-Integrity

- Red entry conditions met 1011 P a g e

I I

I

(-E Fuel Clad Barrier (F)

RCS Barrier (4)

Containment Barrier SLOSS

('-1 POTENTIAL (4-)

LOSS J

POTENTIAL

(

LOSS 4

POTENTIAL

____LO_;

j LOSS LOSS LOSS

ýN4

2. Inadequate Heat Removal (0.)

2. Inadequate Heat Removal

-P4

2. Inadequate Heat Removal A.

or-e-e0Oi A. cer.-eN Not Applicable A. !nad"*Wate Not Applicable A..(Siw i

heileeee teneeople heat renfo;'a4 er4ef 3-eny Feedi~gSgr-gIeet FLadiflgS g~eatef eapabi!it, ia Steamf inte eefe-e eeeing thm(site re e

(sit@i pee geR@ea~er Festeratien temperaturfe aue)...

temperatre value).

i.die..e. by (site.

Core Cooling - Red Core Cooling -

speifrin*ea,..c...

2.RentEoriaticot entry conditions met Orange entry Heat Sink - Red pree4*t not conditions met entry conditions cf

..tive.within 15 OR met.

• 4wate&.-Core B. !Rt PCooling

- Red entry

.conditions met for heat removal

.15 minutes or capabiliy Via 408am Ionizer.

ifidie:ated"' b5,4t-spercifie indications).

Heat Sink - Red entry conditions met

3. RCS Activity / Containment Radiation Fed

3. RCS Activity / Containment Radiation PX-8

3. RCS Activity / Containment Radiation Clad BEIrrer B

A 1. Gentqnmei4 Not Applicable A. Goeanmpen Not Applicable Not Applicable A 1. Gentainment ratdiation monitOr radiatiOn m'-fitfOr radiatiOn MOntz eaW4.-RCB Rad va4Mt-.RCB Rad vete.-RCB Rad Monitor RT-8050 Monitor RT-8050 Monitor RT-8050 or RT-8051 or RT-8051 or RT-8051 greater than 140 greater than 2 R/hr greater than 550 R/hr R/hr OR OR

2. HATCH

2. HATCH MONITOR MONITOR greater than 300 greater than 1200 mR/hr mR/hr 1021 Page

OR B. Site speeifiee indieaticns tdat feaetor eeelant activity is greater analysis indicates that reactor coolant activity is greater than 300 1tCi/gm dose equivalent I-131.

I I

103 Page

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 A. Centainmen isolation is required prezsure grc.ater-than.

AND EITHER of (site specxfic -'alue) the following:

Containment - Red entry conditions met

1. Containment O

integrity has OR been lost based B. Explosive mixture on Emergency exists inside Director containment judgment.

(H2 > 4%)

OR OR

2. UNISOLABLE Cl. Containment pathway from pressure greater the containment than 9.5 psig.

to the speeifie press environment seoont exists.

AND OR

2. Less than one full B. Indications of RCS train of *ste-specifie-s.........

lea k ag e o u tsid e o f sy s......

containment.

eupet Containment Spray is operating per design for 15 minutes or longer.

I I

1041P a g e

Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS

5. Other Indications

5. Other Indications

5. Other Indications A. (site-speeiie -as A. (site speeifie asi A. (sie peeifie-as A (site speei~fie-as A. (sie-speeifie-as A. (stespeeifie as applie*h4*4_N/A app"iea4k)N/A applNe/be)N/A

/aieab4e)-NLA Npplieab4+h-/A aN/A

6. Emergency Director Judgment

6. Emergency Director Judgment

6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the the opinion of the Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director Emergency Director that indicates Loss that indicates that indicates Loss that indicates that indicates Loss that indicates of the Fuel Clad Potential Loss of the of the RCS Barrier.

Potential Loss of the of the Containment Potential Loss of the Barrier.

Fuel Clad Barrier.

RCS Barrier.

Barrier.

Containment Barrier.

105 1P agze

Basis Information For P-WR-EAL Fission Product Barrier Table 9-F-2_

De~eOpeF Notesi FL....L 1A "l.

A*.

.A ifs Each P"R ownlr-'s groeup has devecloped a methedolegy for-guiding the devclepment And implecmcntatiefn ef EONs (i.e., assessing plant par-ametcr-s, and determining and priorit.izin operator aetiOnS). Many of the thrcshelds eentained in the PWR EAL F-issiont Pr-eduet BarrFier Table roflect eenditiefns that r specifieally addrcessed in EONs (e.g., a less of heat r-emoval capability by the s10amn generators). When develepina icpcfi threshold, doeveleperS ShOUld uON the par:amotors: And v~alues specified within their EONs that align with the conditien described by the generic threshold and basis, and rcelatc developer. noteps. This approach will ensurveeonsistency between the site specific EOsan-d emer-gneyl) cl1assific-ation seheme, -And-thus hAcilit~at mo~ro timcly and aecuratc elassifiation assessments.

a d-efin-ed-Set of Crit-ical Safety Functions as part cf their-Emcrgeney Response Guidelines. The WOG approeaeh stmcturces EOPs to maintain andler. rostoroe these Critical Safety Funtioitins, and to do oi Aroiic and systcmatic mnanncr. The WAOG Critical Safe(), Functionis are presented be!ow

" SS 4befiti eality

"-Weat Sink Containment 0

_RCS_ Inventor-y The WO10G ERGs proevidc a moethedolegy for: mefnitering the status Of the CrFitical SafetyFuntlfeions and classifying the significancc of a challcngc to a funcetion; this rncthodology is rcefcrrcd to as the Critical Safety Function StaUS TrFees (CSFSTs). For plants that have implemented the WOIG ERGs, the guidance

-in-N-E-99 0 1 allows' for: usc; of certain CSFS"T assessment roesults as EAI~s and fission product barrior loss/potcntial less thrcesholds. Int this manncre, an emergency classification asscssmcnt may flow dircotly

&AoM A CSýFST assset it is important to tinder-stand that the C SFST-9 arce evaluated using plant par-amcters, and that they arc simply a vendor speeifie mcthod for-eelItivy ev aluating a set of paraeter-s for-pur-pses of driving emergenc oprain poedur usage. For. thl emrAegency conditionis of intcrcest, the genferic thrcsholds..

within the PL 9A1. Fission Proeduct Barioref Table spccify the plant par-ametcrs that define a potential loss or: loss of a fission product ba riF;howve'r-,as dcscr-ibcd in the associatod Devclopcr-Notes, a C-SFST teminus may be used as well. For this reason, inclusion of the CSF-ST related thresholds would be redundant to the par-amctc based thrcesholds for plants that employ, the WOG ERGs.

Sites that employ the WOG ERGs may, at thcir discrction, inelude the CSFST-based loss and potcntial lossý t~chrolApds as dpsoribod in the Developer-Notes. Dovelopers at these sites shoAuld-con-sult with their.

classification dccision makcrS to dctcminc if inclusion wouild assist with timcly and accurtot emer-gency classification. T his dpecisionf should cons;ider: the, effects Of an' Site specific ehanfgeS Wo the genieric WOG C-SFST evaluiation logic and setpeints, as well as those vaising from uiser rulcs applieable to cmcrgcney pcaigproccedurcs (e.g., cxccptions to proccdurc~ cnaty or-transietio due to spccifie accident conditions-Or loss of a support Sy'Stemf).

The CSF-ST thrcsholds, may be addi-esscd int eac of 3 ways:

1.

Not inoer-poratod; thrcshelds will use par-aotors and values as discussed in the Dcvcleper Notes.

2.

Incor-porated along with paramete and value thresholds (e.g., a fulel clad loss wovuld have 2 thrcsholds such as "CETs9> 120OeF" and "Corce Cooling Red entry coniditiefns mct"L.

3.

Used int lieu eof pafameters9 and values for. all thfcsholds.

W'it-h one cxccptien, if a; dcc~vis,_ion is-Made to includeA the-C28F-STV based thresholds, thent all such allowed musl e use in the ishle e Q, ii is ftel peffokRifile le use effiv Me anaete

..1 106 1 P a g e

petential les of the fueol elad barriefr throzsheld and disrcgard all other. C-SFST-based thrczshelds). The ene

~opio is the RCS integrity (P) CSFST-. Beeause ef the eewmplexity ef the P Red deeision point that AMlo mn assessoment a pr-essaro temperatero ctwve, a P Red eefnditicn may be used as an RCS potential

]l-o thrzP-hald v.ithout the need tre in eeR-rarate the other. QSFST-hnRzd hz-hd r*

STP is part of the Westinghouse Owners Group (WOG) and has adopted the WOG Emergency Response Guidelines (ERG). These guidelines employ the use of Critical Safety Function Status Trees (CSFST).

Since STP has implemented the WOG ERGs, the guidance in NEI 99-01 allows the use of certain CSFST assessment results as EALs and fission product barrier loss/potential loss thresholds. This approach allows consistency between EOPs and emergency classifications.

I 107 1Page

PAW--FUEL CLAD BARRIER THRESHOLDS The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

I.

RCS or SG Tube Leakage Loss 1 There is no Loss threshold associated with RCS or SG Tube Leakage.

Potential Loss L.A ThiC reading indieate n

a reduction in retor-vCesel waterF level suffieient to allow the onset of heat induced cladding damage.

Core Cooling - Orange entry conditions-fo~et (CETs > 7080 F) are sufficient to allow the onset of heat-induced cladding damage.

2. Developer Notes:

3.

Petential Less !.A II.

r.i arIy~zroacr eeoling eenditien (e.g., r-euis prmpestorationt aetien). The fveatef.vessel level thatc eefrcspond to aproxmatly the top of active futel may also be used-.

5. For plants that have implemented Westingheouse Ow.ncr-s Groutp Emaergency Response Guiidelines, entcrF the reactor-vessel leve!(9) used-br li t.h--

GrcA G00ifig GtORage Path (iftewltidi depcn~dcnisue the states of RC-Ps, if applicable).

_. Inadequate Heat Removal Loss 2.A This reading indicates temp.ra.. res ithin the eore Core Cooling - Red entry conditions meýt CETs>

1200' F) are sufficient to cause significant superheating of reactor coolant.

Potential Loss 2.A This rading indiates temp.r.a..es within the

.core Core Cooling - Orange entry conditions (CETs > 7080 Fare sufficient to allow the onset of heat-induced cladding damage.

Potential Loss 2.B Heat Sink - Red entry conditions met (4*NR level in all SG < 14% 134%] OR2 pre--ure in at least.one SC >

132 PSIG SG) -AND total AFW flow to SG < 576 GPM). 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 EmergeneySITE 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 SITE AREA EMERGENCY declaration because inadequate RCS heat 108 1 Page

I PW-R-FUEL CLAD BARRIER THRESHOLDS removal may result in fuel heat-up sufficient to damage the cladding and iner-eseraise RCS pressure to the point where mass will be lost from the system.

So-me sise sp-eifi" E0DP s

e.

a o

.P user guid.lin s may establitsh dciien m..aking

-rite-ia onc...n the number er ether attributes of thermoeoupie roadings neeessmy to drive aefetso (e.g., 5 CET-s reading g..ate. than 1,200cF is... uir-ed beferce transitiening to an inadequeat coro cooling procoduro). To maintain cossoc ith EOP9, thes deeisien makinag er-iteria mnay be used in the eere exit thrmce.ouplc r..ading th..sh.lds.

E.... a site sp".ifi t

.mp

.r.atr value that.. rr.sponds te significant in.er.

superh.ating f mroa.t.

coolant. i,200oF ma.y als, be used.

For plants that hae.. implemented Westinghouse Owners Group Emer.geney Respenase Guid.lines, ente the pammoterS and valucs uscd in the Core Coeling Red Path.

Potential Less 2.A Enter a Site Speeific teffPecratUrc valuce that ccrrspends to coro cond-itions at she wnset of heat induced eladding damag (ce.g., the tm..r.atr. all.win. g for. the fer.mation of sup.

h.at.d steam assumf.ing that the RCS is inta.t). 7O..F may also be u.s.d..

Forf: PlMts that Uhpn tcam iIMatr0d WcSti gheus-Owne.s Group

.m...g.. y Rcspens. Guidelines, ontco the parameter-s and valuces used in the CGrm Cooling Change Path.

Petential Loss 2.B3 Enter-th site specific paramoet-r's an.d values that dfi;nc an etxeme halienge to the ability to romove heoat from tho RS via the steam g,.

c.ra-t.s.

These will typically be parameters and values that weuld re

.iPr p ters to take prmnpt action o addr.ess this. c-,nditn.

For: plants that have implemented Westinghouse Owners Grouip Emcr-gcncy Response Guidelines, enter.

the paramcters anid vailucs used in the Hemt Sink Rod Path.

As a less indicatin, dIvelepcrs should onsidcr1 in.luding a thr;- shold the same as, or, similar to, "C1r Cooling Red entif conditionts met" int accordancoe with the guidanoce at the frent of this scction As a potential loss indicati, dcvcl.p...

sh.old c..sidcr. ine.l.ding a th..sh.ld the same as, or. sim.. ilar. to, "Core Cooling Orangc*

nt. y conditionis met" int accordanc. e with the guidance at the f1nt of this se;tion.

As a potcntial loss indication, dcvclcpcr~s should considor-including a thrshold the same as, or-similef to, "Het ink Red entr1' econld-itions; met" in acer-ffdnance with the guidancep at the fron;t of this etin

-1

__RCS Activity / Containment Radiation Loss 3.A. I The radiaicn-readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond meniter rcading cerrcespnds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals 300gCi/grn 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% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier. The values for RT-8050 and RT-8051 were based on Calculation STPNOC013-004 Rev. 1. The threshold values used were rounded from the calculated values by approximately 2% to ensure the values were readily assessable.

I 1091 P a e

]PWR-FUEL CLAD BARRIER THRESHOLDS Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-uD monitor to the containment high range monitors (RT-8050 and RT-805 1). The HATCH MONITOR threshold value is based on I I I I I

I Calculation No. 03-ZE-003. This value corresponds to the calculated containment high range monitor readings for Fuel Clad Barrier Loss 3.A 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 levl EMERGENCY CLASSIFICATION LEVEL to a Site Area Emergency SITE AREA EMERGENCY.

Loss 3.B This threshold indicates that RCS radioactivity concentration is greater than 300 g.iCi/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% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

Potential Loss 3.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

This

.thr.eshold is the HATCH MONITOR. whi.. h is the ba. k up meniter to the ntai.. n.ent high rang monitorps (RT 9050 and RT-9051) and is located outside containment adjaeent to the per-sonnel hatc-h.

it is reeegniezcd that sarnpie e.Jlltien and analy#sis of reaeter-coolant wiih highly elekvated actiit 10cVOclS Pcould require scycral heurs to complctc. Nonetheless, a sample Felated thr-esheld-is inclusde as a backup te ethcr indications.

There is nie potcntial less flhrcsheld associated with RCS Activity / Contaiinmcnt Radiation, Deeloer-News-Th--h rc-.....i..g Shuld be d...

1assuming the nr.e.lease and dispersal f the rt*;r..

D colant noeble gas and i.dine inv1nty, with RCS fdieactiit c.n..ntratin equal te 300

.i..g.

dose equivalent 1 13_1, into the ontainment atmosphee.y Thr~esheld values shetuld be deteffpned assuming RCS radioactiit cocntain equals 300 P*Ci/gm dose equi-valent 1 13 1. Other-site specific units may be used (e.gCie.

Depcndin upnie secific capabilities, this thr-eshold may have a samople analysis componient andler-a radiation monitor reading ceompcnent.

Add this paragraiph (Or simfilar WOrdling) to the Basis-if the th-reshoeld inc-lud-es -a sample analysis, eompentent, "it is r-ecognized that sampl cIllctio and aftalysis of reactor-coolant with highly eleyated activity levels could r-equire sevcral hourfs to complete. Nonetheless, a sample related threshodi included as a backup to other indioatiens."~

I.

Containmnent Integrity or Bypass I

1l0 Page

IW-R--FUEL CLAD BARRIER THRESHOLDS Not Applicable (included for numbering consistency) 4-qL. Other Indications I Loss and/or Potential Loss 5.A This subeategcry addrcsses etheF site speeifie thrcshelds that may be ineluided te indieate lcs 4rptct less ef the Fuel CPad bafrier-based en plant spcoifie design eh-aractcriztir#4z not oznqidzread in the gefnznc gu:ianee.N/A DevelpeF Nots Less and/zr-Potential Le%4.-A Develepers should determine if ethefr rciable indieatecrs ecsist to evaluiate the status of this fissien przduet barrier. (c~. -vic eidefft analyses dc-scribed-in sthe site -Final1 Safety Analysis Repzrt, as updatzd). The geal is to identify an), unique er sites peeifie indieations that will pr-emote timely and accur-ate assessment of bmarrier status.

Any added thresholds should represent approximat he, thc sina relati;'e throat to t"h barrfior as thP Athcr throeshelds in this columfn. Basis infermatien fer-the ether thrcsholds may beo used te gauge the r-elative barrficr throat level.

6,.

Emergency Director Judgment 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 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 should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

I 111lIPage

IW*R-FUEL CLAD BARRIER THRESHOLDS Nefie 112 1 P a g e

-P;R-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.

I.

RCS or SG Tube Leakage Loss L.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 EmcrgencySITE AREA EMERGENCY-since the Containment Barrier Loss threshold I.A will also be met.

Potential Loss L.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 gite Area Emcrg neySITE AREA EMERGENCY-since the Containment Barrier Loss threshold L.A will also be met.

Potential Loss I.B Integrity - Red entry conditions This eefiditieii-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).

Less.A

.Actulationfl of the E-CC"S may also be Fr&Aferrcd_ W -AS Safety Injeetion (SIl) actuation or other approraezt speeifiete~m.

Petential Less !..A-113 I P a g e

SPWR-RCS BARRIER THRESHOLDS DOeponding upon charging8 pump flo caaitcendRSw'lncnrl par-ametors, develepcr-9 may use-an RCS leak ratovalue of 50 gpm, er

'nepopit site speeific valuc, as an altemate Potential Less throsheld. If used, th threshold wording should reflecmt thAt th4e dterntion fthc leak rate aluAe exeludes no-Rnnal routosin -RCS inventery (c.g.. by thc letdevn System 8r R C= _ea s A-A -A-f6&.

Potential Less 1.B Enter-the site speeifie indicatiens that define an extrofme challenge to the integrity of the RCS pr-essufe boundafy due to prvssur-izod thermal shock a transieat the at casesmrpid RCS eeeldewnt while the C is in Mede 3 or higher- (i.e., het and proessurfiizd). These will typically be par-ametefs andv;aluoes that w.uld freuir. p...... to take prompt action to address a press*uized thenal hAIPk* "Andition.

Developeors shoul-1dail-so deteffnot if tho throsheld needs to rofleet any depondeneios used as EOP tr-afsitiep~zentf decisin pons rcndition validation er-iteria (e.g., an EOP used to respead to an excesive CS coldew may not be onerouefd-or immffediately exited if RCS pre-SSURe0 is oo ac-i For plants that have implemented Westinghouse Ownors Group Emergonoy Rosponse Guidelines, entor:

the paramoetors and values used in the RCS lfntegrit)'Red Path. Beeauso ef the comploxity of Regain decisi.n points within the Red Path of this CSFST, developets at th*se plants may oloot to not inelude the spociflo parameterS and -.aluces, and instead follow the guidanoc below.

Westin",houe ERG Plafftq As a potential loss indication, developors should eonsiderincluding a

-esh-ld the same as, or similar to, "RCS Int.g.ity Red.nt.y conditionis Met" in aecordane. with the guidanee a the front of this seetien. As no.t.d above, d

.v...

pe.s should ensur that the thr.sh.ld wording r.fl..ts any EOP transition/ontry deeision.

points or. condition validation cr-it.".ia. For e-aflple, a throshold might road "RCS" lnteg*i.y (P)

Red eatr.' saditiews mzt; wih RCS nr~e~sur-e>300 nsig'

2.

Inadequate Heat Removal j*..........

E" -- "*"

Loss 2.A There is no Loss threshold associated with Inadequate Heat Removal.

Potential Loss 2.A Heat Sink - Red entry conditions met ({4NR level in allA-- SG < 14% [34%] OR pressure in at least efne SG-> 1325 PSIG SG, AND total AFW flow to SGsSG < 576 GPM).

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 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 SITE AREA EMERGENCY declaration because inadequate RCS heat removal may result in fuel heat-up sufficient to damage the cladding and me esseraise RCS pressure to the point where mass will be lost from the system.

Deveoe-Notesi Potential Less 2.A 114 I P a g e

I PWR-RCS BARRIER THRESHOLDS Enter the site speeific pearamcters and-vallues that dc64fine Min eotr-ema ehallengc to the abilify to rcmciev heat ffom the RCS via the steam gefncrator~s. These will tyically be paramctcrefs and values that wetuld ropuuire onr ter to1e tare rmet acItion 4to address tq his. condition.

A I-O orpeAntS Ofat nave impiesmcntca w cstingnourie Wwncrs rCuftp IMregefncy *eSPOfl55 "uiainOS, Ontcr the P rn'meti-' and vanluesuse iiv' n the~ Heat Sink RedI Pat Westifiehouse ERG Nam.&

• a I

• q Uselpe-s.hetld consider ineitidig a throsheld the same as, or-stmilar te, "14eAt Sink Red eatry

3.

RCS Activity / Containment Radiation Loss 3.A.

The r-adiatin menitorr readigCalculation STPNOC013-CALC-004 provided a value that corresponds to an instantaneous release of all reactor coolant mass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. The calculation result was that containment radiation monitors would be readina 450 mR/hr. These monitors have an average backgzround reading of 1.5 R/hr due to the presence of a "keep-alive" source. The value of 2 R/hr was selected because that would be the expected response from the "keep-alive" source and the calculated source term from a loss of RCS (450 mRihr).

The threshold value of 2 R/hr should not be used until approximately 40 minutes after a suspected loss of RCS. For approximately 40 minutes following a loss of RCS, RT-8050 and RT-8051 readings are expected to be influenced by a temperature induced current as described in DCP 04-8245-33, Replace RCB HRRM Cables With Cables Less Susceptible to TIC. This effect may cause variations in RT-8050 and RT-8051 reading from 0 R/hr to 4 R/hr. After approximately 40 minutes. RT-8050 and RT-8051 are not expected to experience a temperature induced current and the threshold value should be used.

If a secondary system break inside containment is suspected, this threshold should not be used to determine a concurrent loss of RCS for approximately 90 minutes. This -auc is loe-r tkhan that spccified for Fuel Clad Buarriecr Less thrcelshold I.A s incc it indicatcs A Aqlo o*

th, RCS; BRnior-anl.

Potential Loss 3.

There is no Potential Loss threshold associated with RCS Activity / Containment Radiation.

Develope*- Notes.,

l ee-3 A The rcading shouild be determined assuming the inistwntancous releses and dispersal ef the roaeter-eoolant nTble gas and iodine invenateoy, with RCS activity at Tehaical S i

ia a

limits, into the nAW;i.m.nt atmsphc... Using RCS a1tivity at Teehnical Spccifieatien afllwabie limaits aligns this thrcshold with WC SU3. Also, RCS activity at this level will typically r-esult in contai*nment raidiation levels that Pan be mor-Asv raily deteeted by Gontainmente raditio moito, ad more roadily differ-entiated from these caused by piping er-eempenent "shins"'

seur-ees. if desirsid, a plant may use a lessefr value of RCS activity for deter-mining this value.

b; soeme eases, the site speeifie physical location and sensitivity of the Giontainmont ratd-iatisin monitor(s) may be sueh that radiatien from a cloud ef released RCS gases carmoit be distinguished from radiationi I

115 1Page

I PWR-RCS BARRIER THRESHOLDS 44.

efmanating 48oM Piping ean eComp~flefnts eontinfinfig elcx'ated rootetr eoolanft acti;'1t. 498c, rotor 4c the Y

P i

TNýyrýiý ýý XT-*-.

jr-T T AQq 1Z A Rnd EleteFmme 4 aft altemate indiee4tan is Fivailfible Containment Integrity or Bypass Not Applicable (included for numbering consistency)

5.

Other Indications Loss and/or Potential Loss 5.A This sibeateger-y addrcesses other Site speeifie throsholds that mnay be inceluded to indicate los 1rpota los; Af the RCS barriAr haSEd Onf planlt speeifie design eharacteristics met Pensider-ed in the gen---ric guianaee.-N/A Developer-Notes:s Lcss and.'or Potential Leass 5.

Dcv clopers should determine if other reliable indicators exist to c;'aluatc the staws of this fissiont produtfe barrier- (cgr eo cident analyses deser-ibed in the site Final Safety Analysis Report, as updated). The goal is to idhcnify anfy uni ritspoifie intdieaticnS that Will Prolmte timely and to c

uarer at esscsson of barrier-staws.

Any added throesholds shouild r-epr-escn

!pr1iaoy the samoe relative throat to the barrier as the otbor throsholds in this column. Basis inforemation fer-the other throesholds may be used to gauite the rolative

6.

Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS 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 RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

I 116 1Pagie

I RIW-R-CONTAINMENT BARRIER THRESHOLDS NetneThe 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 Alei4-ALERT to a Site Area Emergency SITE AREA EMERGENCY or a General Emergency 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.

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 FA ULTED definition] and the faulted 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 (and are thus 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.

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 -AR ICs.

117 P agc

I PWR-CONTAINMENT BARRIER THRESHOLDS I The emergezey elaszifieatien levelz 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 (er-ether-valt.. per SU4 Developer-Notes)

Greater than 25 gpm (eF ethcr ;'aluc per SU4 DeveepeF Nete*)

Requires operation of a standby charging (makeup) pump (RCS Barrier Potential Loss)

Requires an automatic or manual ECCS (SI) actuation (RCS Barrier Loss)

No classification Unauual Event UNUSUAL EVENT per SU4 Site Ar*a Emergency SITE AREA EMERGENCY per FSI Site Area Emergency SITE AREA EMERGENCY per FS1 No classification Unusual E--nt-UNUSUAL EVENT per SU4 Al4ALERT per FA1 Alof ALERT per FA I IPotential Loss 1.

There is no Potential Loss threshold associated with RCS or SG Tube Leakage.

Devl1ope Note IA.,r I

A:

~

~

~

~

L--:-

A seamgeaf~aef~ewe~ep teifeielvalemyalevetefetleasnatesptertesea hip valve efr other-aPPrzPriate site speeifie term.

Dz-vzlepOrama MfiinlU-dz an addditi@1nal site speeifie thrzzheld(s) to addresS prelcngcd ztaM FARcSLzAS neeessitated by epeeAtienal eensidefratiefis if AOPs er EOP9 eetuld rzguir-e that a leaking er RUPTURED steam gen~rtcO-Rr-beA uzcSd W3 Suppcrt plaft coaldOWn.

Dc;We1@P0rz may Wizb to-z-Anzidcrf iAnccP~fmt6ng the -Ahb3;' tablz into user-aids (e.g., a wallbeffd) er-othcr-loAt~izn3q Within thzir-bAqiq dAAcumzntj.

2.

Inadequate Heat Removal Loss 2 There is no Loss threshold associated with Inadequate Heat Removal.

I 118 Page

I P--R-CONTAINMENT BARRIER THRESHOLDS Potential Loss 2.A Core Cooling - Red entry conditions met for 15 minutes or longer. This condition represents an IMMINENT core melt sequence which, if not corrected, could lead to vessel failure and an iner-easedhigher 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 and/or if reactor vessel leveIRCS level is increasing. Whether or not 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 is appropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.

Developer Noesei Some site speeifie EOPs andler-EQP user-guidelines maey establish deeisien maaking er-iter-ia osm the nimbr-or-ether-attfibutes of thefrmosouple readings neeessafy to drive aetions (e.g., 5 CETs r-eading greater than !,200oF is required be--fef tranaitioning te an inadequate core cooling pro.edure).-: T ma......... sonsistsnc ith EOPS, th.S. dcsis.io making FAR ma. be uS. d in th. c.r. o.it thinmccouple reading thPasholds.

Entefr site speeifie eritsi reuiinentfy itot a e-r-eling Festeratien prooedurs~ er-pr-empt implementationt ef eer-e cooling rostefratiefn actions. A reading of 1,2OeF en the CET-s may also be used.

For. plants; th-at hasimplemented AWsstinighetuse Owners Group Emsir-geney Response Gutidelincs, entsr the par-amster-s and values used in the Corse Cooling Red Path.

DeVelOperS shOUld eeonsidfr insrluiding a thresheld the same as, Or-similar to, "Coree Cooling Red ent",

eenditions mcit for 15 minutifs or-lenger" in aecordanec with the guidanco at the front of this seetien.

I 119I iP a g e

I #W-R-CONTAINMENT BARRIER THRESHOLDS

3.

RCS Activity / Containment Radiation Loss 3 There is no Loss threshold associated with RCS Activity / Containment Radiation.

Potential Loss 3.A. 1 The fadieaen-readings for the containment high range area monitors (RT-8050 and RT-805 1) correspond m.niter. r.ading e...

.p.nd. to an instantaneous release of all rcactcFr coolant mass-the radioactive material inventory of the reactor coolant system (i.e., All the RCS coolant mass) into the containment, assuming that 20% of the fuel cladding has failed. The values for RT-8050 and RT-8051 were based on Calculation No. STPNOC0 13-004 Rev. 1. The threshold values used were rounded from the calculated values by approximately 2% to ensure the values were readily assessable. This level of assumed fuel clad failure is well abee--beyond that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the....

g....y.lafiifieatic.

lt.le EMERGENCY CLASSIFICATION LEVEL to a General Emergency GENERAL EMERGENCY.

Potential Loss 3.A.2 The HATCH MONITOR is located outside containment and is the back-up monitor to the containment high range monitors (RT-8050 and RT-8051). The HATCH MONITOR threshold value is based on Calculation No. 03-ZE-003. This value corresvonds to the calculated containment high range monitor readingzs for Containment Barrier Threshold Potential Loss 3.A. 1.

Develper Natoq.

Petential Less 3.A NUREG 1228. Nei~rc-EstM06efqns Driwfg lmeidwon Rospnse te!ovn Sevr mi/ar Power. PWPant Accidens, pro;'Videq the baziz for: uz5ing the 20%'4 fucl elatddinig fatihurc N'010e. Unflcc therc iS & Site Speeifie analyc5'i9 justifying a diffcr-ent value, the rceading should be determined assuming the intstantaneous rclease and dispersal of the fvaetcr-eeelant neble gas and iodine inctr, cciated with 20% fidel elad failtwe inte thP-P-ntainAiMcnt AtMFA.ihere.

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. Users are reminded that 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 (or sometimes referred to as design leakage). Following 120 1 P a g e

I PWR-CONTAINMENT BARRIER THRESHOLDS 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.).

Refer to the middle piping run of Figure 9-F-34. Two simplified examples are provided. 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 would be 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 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 AR 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.

Refer to the top piping run of Figure 9-F-344 in Addendum 3, Containment Integrity or Bypass Examples. In this simplified example, 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.

Refer to the bottom piping run of Figure 9-F-43. In this simplified example, 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 clocd w-Aetr cooling Component Cooling Water 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 I

121 1Page

I PWR-CONTAINMENT BARRIER THRESHOLDS 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 AR ICs.

The status of the containment barrier during an event involving steam generator tube leakage is assessed using Loss Threshold l.A.

Loss 4.B Containment sump, temperature, pressure and/or radiation levels will i-ieFearise if reactor coolant mass is leaking into the containment. If these parameters have not iaereasedrisen, then the reactor coolant mass may be leaking outside of containment (i.e., a containment bypass sequence). nefeaseRises 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 should 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 ier-seaerise 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.

Refer to the middle piping run of Figure 9-F-344 in Addendum 3. Containment

,nt.gr.ity or-Bypa 9*eamiples. In this simplified example, 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. 1 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 Containment - Red entry conditions met (containment pressure > 56.5 PSIG). 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. Thus, this threshold is a discriminator between a Site Area Emerwgeaey-SITE AREA EMERGENCY and General Emergency GENERAL EMERGENCY since there is now a potential to lose the third barrier.

122 1Pagc

I PWR-CONTAINMENT BARRIER THRESHOLDS 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 4% )). 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 potential loss of the Containment Barrier.

Potential Loss 4.C This threshold describes a condition where containment pressure is greater than the setpoint (9.5 PSIG4SP.S4G) at which eontainmcnt c

• gy (heat)..m.val systems arc Containment Sprav is 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 in that conitainment-hftat-rmVai'dcprcssu0r-iat ytm (e.g., contaffinenit SPrays, iee condcnscrF fans, cte., but noet inicluding conftainment ;cnting stratcgies) arc Containment Sprav is either lost or performing in a degraded manner.

Developer-Notesi Less-4.A4 Develeper-9 may inelude a list ef site speeifie radiation monitor-s to better-definc this thr-esheld. Expeeted menitor-alanus or: r-eadings may also be included.

Potential Less 4.A The site speeifie pre ssirve is thP -Ae Aon ta;I imnenat deg rsue FOr pleAtS that hgA;e impliemfenfted WeStifghOUsc OY,.Hr-Grou1p Emfergency Response Guidclincs, the pr-esSure v~alue in Potential Less 4.A is that used for-the Containmcnt Red Path. if the Containment C-.

SS

.t-

• mcre than I ne Red Path due to A t t er dependencies (e..,.

.tatus of

t.

c tainn, t i* olation),

eater-the highest containmcnt prcessure value shownf en the trcee. This is typically the eontainmcnat design Potential Less 4.B Dev elepcrs may enter the minimuem containmcnt atmesphcr-ie hydroegen eenecnitratien flccossary to supprt.

a hydog*cn burn (i.e., the l.wcr-deflag.tion* lim;it). A encrtrcrnfet c

ntg concentration may be includcd ifthe pflanths this I;-ind-ica-tionA ava-lable0 in t-he Controzl RAPm; Potential Lss I.C-Enter the site specific pressure se.4ointvaluc that actuates e.ntai n psue contro.l systems (e.g.,

eentainment spray). Also enter-the site speeifie coe imn p Hrcsr contoel systewa'egutipcnt tat should be oprtn Fe design if the eotimctpessur~e setpet t is roeahed. if desired, spceie A-A-1;d-itio Afindications such as parameter va-lues can-R -also;A be enter:ed (e.g., a conainm tspa flow raFte lessN t-h-an1 -A eertai lue).

This thr-eshold is noet applicable to the U.S. Evolutioinary Powcr-Reactor (EPR) design.

As a potential loss indication, dcvelepers should consider including a thr-eshold the some as, or. simnilar to, "Containment Red eat",~ conditions met" in acor-dance with the waidaftee at the front of this section.

5.

Other Indications I

Loss and/or Potential Loss 5.A This ubaeoyaddr~esses other site specific th-re-sholds that may be included to indicate loss or-potential l8Ss of the Conitainmffent barrier based on plant specific design characteristics not consider~ed int the generic guidfmeeý-N/A I

123 1 P a R e

I W-CONTAINMENT BARRIER THRESHOLDS if Site emfergency operalting procodUreS prov~ide for-venting of the coentainment as a mneass of peetn eanastope thre hclue shrepsholdpshould Pe imnlu"ded ferte renatimetrt htfler. This threshold would be met as sren as sceh venting is IMMINENT. Containment ven furdevery atin classified in accor-dancee with the radiolegical effluent W~s.

Developers should determinc if other reliable indieator-s exist to evalutate the stats of this fission proeduct barrier-(~. riew cident analyses deser-ibed in the site Final Safety Analysis Repeil, as updated). The goal is to identify s any u

tespeific indicatios that will proemote timely and aecurate assessment ef batier stains.

Any added thresholds should r-eprcescn aproiatl the same relative threat to the barrier as the ether threSholds1-in; th-i colu0-0mn. BasiS infoermation for the-o-therO t-hresholds mfay be usEd tW gaufge the rel ative barrier-threat level.

C.

Emergency Director Judgment Loss 6.A This threshold addresses any other factors that may 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 should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

Develop" Netes!

Nenie I

124TP a g e

I PWWCONTAINMENT BARRIER THRESHOLDS:

I I

125 1Page

I Figure 9-F-43: -W-R-Containment Integrity or Bypass Examples RCP Seal Cooling NOTES: Only Supplemental Purge is a filtered release and STPEGS Component Cooling Water is equivalent to Closed Cooling Water I

126 1 P a g e

10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Table H-1: Recognition Category "H" Initiating Condition Matrix UNUSUAL EVENT HU1 Confirmed SECURITY CONDITION or threat.

Op ] Modes:,4t-ALL HU2 Seismic event greater tha+ OBE levels.

Op. Modes: 44ALL HIL 3 Hazardous event.

Op, Modes:,4#ALL HI 4 FIRE potentially deg fading the level of safety of the plant.

Op.[ Modes. 41/ALL ALERT SITE AREA EMERGENCY GENERAL EMERGENCY HA1 HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Op. Modes: 4,-ALL Note:

4.4 See SA9 or CA6 for escalation of these events HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown.

Op. Modes:,4/ALL HA6 Control Room evacuation resulting in transfer of plant control to alternate locations. Op.

Modes: 41/ALL HA7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Al4,ALERT.

Op. Modes: 41ALL HSI HOSTILE ACTION within the PROTECTED AREA.

Op. Modes: 4I4ALL HG1 HOSTILE ACTION resulting in loss of physical control of the facility.

Op. Modes: A44ALL HU7 Other conditions exist which in the judgment of the Emergency Director wai rant declaration of an (Mr NW )

UNUSUAL EVENT.

Op. Modes: 4//ALL HS6 Inability to control a key safety function from outside the Control Room.

Op. Modes: 444ALL HS7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a Site Arzn Emer-geney 9--SITE AREA EMERGENCY.

Op. Modes: -4/ALL HG7 Other conditions exist which in the judgment of the Emergency Director warrant declaration of a GENEPRALGENERAL EMERGENCY.

Op. Modes:,41ALL 127 1 P a ge

HUI ECL: Netificaticn cf Unusual Event UNUSUAL EVENT Initiating Condition: Confirmed SECURITY CONDITION or threat.

Operating Mode Applicability: AItALL I E*.mp!e-Emergency Action Levels: (1 or 2 or 3)

I1LA SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by speeifie

.....r.ity shift upc...isi)* ANY of the following personnel in Table HI:

Table HI: Security Supervision Security Force Supervisor Acting Security Manager Security Manager I fL2-L-Notification of a cr-d:iblc zccurity threar CREDIBLE SECURITY THREAT directed at the site.

[ --

ý 3i.A validated notification from the NRC providing information of an aircraft threat.

Basis:

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..ee....y

.e.ent,-sSECURITY EVENTS which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.. Seefiity-*evens-SECURITY EVENTS assessed as HOSTILE ACTIONS are classifiable under ICs HAl, HSI and HGI.

Timely and accurate communications between Security Shfl-Force 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 !n-*p3xid.u! Sp;ni

• ;fe! Ssrgc, g

s,*.,aAoib, INDEPENDENT SPENT FUEL STORAGE INSTALLATION Security Program].

EAL #1-references (site spe'ifie zcc ur-ity.hifl.

upervisizn) Security Force Supervisor because these are the individuals trained to confirm that a

-eeurity-event-SECURITY EVENT is occurring or has occurred.

Training on

-eei~4 y,'cet SECURITY EVENT confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39039 information.

EAL #2-addresses the receipt of a crcdible 3ccurithy thratCREDIBLE SECURITY THREAT. The credibility of the threat is assessed in accordance with (site spceific procedur) OOSDPOI-ZS-001 1.

Implementing Procedure For Safeguards Contingency Events.

1281Page

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 (site specific procduro) 9POP04-ZO-SEC4. Guideline For Airbome (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to Airborne Threat.

Emergency plans and implementing procedures are public documents; therefore, EALs sheti4ddo 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 Ahe,-u be is contained in nen public documents such as the Security Plan.

Escalation of the emergeney via IC HAl.

eiasstiticatten level&EMERGENCY CLASSIFICATION LEVEL would be HUI: EAL-1 Selection Basis:

For EAL-1. the nosition of Security Force Sunervisor was included since it is a 24-hour nosition_

Normally the event would not be reported bv the Acting Security Manager or Security Manager because the Acting Security Manager position is not normally activated until after an UNUSUAL EVENT has been declared, and the Security Manager position is not normally activated until after an ALERT has been declared. However, reporting by the Acting Security Manager or Security Manager was included in the event these positions are staffed under unusual circumstances.

REFERENCEs:

1.

OERPOI -ZV-SH03, Rev. 12, Acting Security Manager

2.

OERPO1-ZV-TS08, Rev. 16, Security Manager

3.

OPOP04-ZO-SEC4, Rev. 10, Guideline For Airborne (Aircraft) Threat (SUNSI)

4.

0SDPO I -ZS-00 11, Implementing Procedure For Safeguards Contingency Events (Safeguardsl 5

Security Force Instruction SI 2700 Security Resnonse to Airborne Threat (SU TNSI J

Develper-Notes-IThe (site speeitce secufity shilt supcr.~isieon) is the title ef the on AIRt injdividual rosponsib!cb The (site speeifie proccdur-e) is the proccdurc(s) usced by Controel Room and/er-Seeturity per-sennel to defermine if a seettrity throat is erodibic, and to -validate r-eeeipt of aircraft throeat infermation+.

Emergency plans and imoplementing prfecedafts are public documents; thremfore, EA4.s should not

.n rirte Socur~ity sensitive inforemation. This incluides information that may be ad;,fantnfgeous to a potential ad;'r-sary, such as the pariticulas conccrnfing a speeific hrAtor-thrnoat location. Socurty sensilive in-fosrmation should-be contained in non publie documents such as the Secur-ity Plan With due considefratiefn given to the abov dvlepeF nee EALs may contain alpha or nluffbcr rcfcrcnees to selcctcdcevents deser-ibed in the Scour~ity Plan and associatcd implementing pr-eeedurs.

I cron cessnoujan otcont ain aroccgn i zawcacsorwu on01mccyc nt.r orcr.a m~i c.antrwma yo cwo rsb od wef:Eled as "Seetir-ity event 42, 45 er-49 is r-epened by e (site speeifie seettr-iw shift 9ttj3eFvi9iefW-"

,- A -

A-,,..

^

I} I I

A

• T k

I I

II 129 1P age,

HU2 ECL: Notification of Unusual Evcnt UNUSUAL EVENT Initiating Condition: Seismic event greater than OBE levels.

Operating Mode Applicability: A41ALL I Etmampie Emergency Action Levels:

(1) a. EITHER of the following conditions exist:

1.

Seismnie event greater-than Operating Basis Earthquake (013E) as indieated by: (site speeifie i.ndication that a* seimic e'ent met Or exc*eeded OBE limit"SEISMIC EVENT" alarm in Unit 1 Control Room (Lampbox 9M01, Window E-8)

OR

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

AND

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

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 safety of the plant.

Although the "SEISMIC EVENT" alarm (0.02 g) in EAL l.a is set below an O.B.E earthquake (0.05 g)P it does provide an indication that a seismic event has occurred. In order to determine whether an O.B.E.

earthquake occurred, additional indications may be needed. Determination per OPOP04-SY-001, Seismic Event is not practical if it takes longer than 15 minutes to perform.

Indications described in the EAL should be limited to those that are immediately available to Control Room personnel and which can be readily assessed. Indications available outside the Control Room and/or which require lengthy times to assess (e.g., processing of scratch plates or recorded data) should not be used. The goal is to specify indications that can be assessed within 15-minutes of the actual or suspected seismic event.

The EAL L.b-statement is included to ensure that a declaration does not result from felt vibrations caused by a non-seismic source (e.g., a dropped heavy load). Event verification with extemal sources shold net be niecessafry dur-ing or-fellewing an OBE. Ear-thquakes of this magnitude shetuld be readily Mct by en site p......l and rceegnizcd as a seismie event (e.g.,

ie.l latcral

.e.eleratiens are in eecss of 0.08g). The Shift Manager or Emergency Director may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely 130 1 Page

emergency declaration. It is recognized that this alternate EAL wording may cause a site to declare an UNUSUAL EVENT while another site, similarly affected but with readily assessable OBE indications in the Control Room, may not.

Depending upon the plant mode at the time of the event, escalation of the ei EMERGENCY CLASSIFICATION LEVEL would be via IC CA6 or SA9.

nsrgsnsy siassuisaucn isvsi HU2: EAL-1 Selection Basis:

STP does not have a readily available indication in the Control Room for determining if the site has experienced an OBE. The Seismic Event Alarm setpoint is 0.02g in the vertical or horizontal position and the station design basis value for an OBE is 0.05g. Since the Seismic Event alarm is set at less than half of the OBE value, it cannot be used as the sole threshold value for determining whether or not STP has experienced an OBE.

STP has implemented the alternative EAL described in NEI 99-01 Developer Notes in coniunction with using the installed indication. EAL-1, b. allows the Shift Manager or Emergency Director to determine if a seismic event has taken place, taking into consideration the Seismic Event alarm. Control Room personnel feeling an actual or potential seismic event and other indications deemed appropriate.

REFERENCES:

1.

OPOP04-SY-000 1, Rev. 8, Seismic Event

2.

NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors.

Developer-Noesez T-his- "Site Specific ind~icAtion th-at -A 5eismic event met A;r exceeded OBE limits" qhou~ld be bmaed on the indi-ati.ns, alarms and displays ef site sp-cific scimic mCnitenng cguiprcnt.

indieatiecns deseribcd in the EAL sheuld be limited te these that arc immediately avatilable te Ccntrsl Room persenncl and 'hich-esan be r~eadily assessed. Indieations -avail-able Rcutsidce the-Contrzel Room andicr-Which r-Cgu11-irc lengthy times to assess (e.g., przcsssing of serateh plates or-reeerded data) sheuld nzAt be u-scd. The goal is to specify, indicA-tions that ean be a sessd within 15 Minutes of the actual or Suspcstcd spismip AA.cnft.

For-sites that de net have readily acssessable OBE indiestiens withint the Ccnftrcl Reem, develepecrs should use the fellewing altsrnatc EAL (er simiarF Werditg).

a. Centrel Reem peFSennslj feel an actuial or pctential seismie evelut.

AND

b. The eeeuf~enec Cf a sismie event is ccnfifmed in mwmsrF deemed apprspr-iatsby the Shift Manager or-Emcr~gency Dircster-.

The E--

I.b statement is in"luded to ensure that a de-l-e aration doss not

r 1

f..M ft viratims C'ausepd by a non secism-icsue-roce (e.g., a drcppcd heavy load). The Shift Manager or-Emsrgsnsy Dirccter-may seek e.t.mal ver.ifi.atio..

if de...d appr-pr.iat. (e.g., a all to the USGS.. ehcck intcmrnt news souroes, ete.); however-, the ver-ifieation action maust not preclude a timcly emaer-gency d-eclafr-ation. 1t is recognized that this altemate EAL wor-ding may sause a site to deelafe an Unusual Event while another-sits, similarly affccted but with rcadily assessable OBE indieatiefns in the Centr-el RCem, fmay net.

T Rho @0P 4bv le ats rdinig mayfý also be used te dcvelep a compsnsatary EAL--

Cfor use durfing perieds when a seismie menitern sytmspable ef dstssting an OBE is out ef ssr.'iee for-maintenancr 131 IPaue

I EGI, Assi gpimcn Attfibutesz: -3.1.1 i.A I

132 1 P a g e

HU3 ECL: Nctifizatic, n cf Un'uala Evnt UNUSUAL EVENT Initiating Condition: Hazardous event.

Operating Mode Applicability: A4lALL Exam.eple Emergency Action Levels: (1 or 2 or 3 or 4 oi--or 5)

Note: EAL #4 does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

L(L-A tornado strike within the PROTECTED AREA.

I (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. 14)

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).

I-_2-*4)A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles.

(3)

(Sit s.epcifie list. f nmaural er-teehnlgical hazard evcnts) Predicted or actual breach of Main Cooling Reservoir retaining dike along North Wall kS.L._

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

EAL #1 - addresses a tornado striking (touching down) within the Pr-,teeted AreaPROTECTED AREA.

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.

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, etc., or an on-site train derailment blocking the access road. This EAL is not intended apply to 133 1Page

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 45 addrcsscz (site specifie descr-iptien).

EAL#5-the Main Cooling Reservoir breach along the north wall which was included because it is a credible hazard and analyzed in the STPEGS UFSAR.

Escalation of the emergency classificatlin level EMER based on ICs in Recognition Categories AR, F, S or C.

GENCY CLASSIFICATION LEVEL would be HU3: EAL-1. EAL-2. EAL-3, EAL-4 Selection Basis:

N/A

REFERENCE:

1.

STPEGS UFSAR, Section 3.4. 1, Flood Protection Developer-Notes, The "Site speeifie list af natural or. technolegical hazar;jfd cvcnc"q qhul incldA Lthcr-cvcnts that maBy be a proceur~er toamr inificant event or-condition, and that are prpraot the site location and ehafaeterites.

Nctv~xithstandiag the events speei f i ally included a sV -A-Is aboAeve, a "S ite speeif-rleo li t Af ntlOr technoelegieal haizard events" need net inelude chort lived events fcr-which the cxtcnt of the damage and the rccsultn con'cgucnccc an be determincd within a relaively short timoe famfn. in these eases, a damage assessment can be pcrfcrimed seoon aftefr the event, and the plant staff will be able to identify petential or-actual impacts to plant systems and siteructrc. This will enable prompt definitionan i mplemen t-atio o-

-f eompencate ry OF V-,

Ofte qIiv e Fne-asurco smA with n A approc iab! e incrfea sc in r-i sk to the pub lic.

To the cxtcnt that a short lived event does Pause in~nndiate And significant damnage to plant systems and stfuetur-es, it will be elassifi-abic nd the Reeognitien Categor-y F, S and C Wsc and EPLS. Events e Iesscer. impact would be expected to cause enly s-mall an;-d-locealized damage. The conseguencec from thoes additiont, the... uf...nc. or-

.ec.t of the event may bo rope.t-abl under the r..uir-em.nt. of 10 C=FtR 50.72.

A A*

"1 I l

A

• "1 1 I

P eq=i.,k9stemment Attfihtttesý 6. 1. I.A and A. 1. 14-2 I

134PI ag e

HU4 I ECL: Nefifieati Oft 0'f Unusual Event UNUSUAL EVENT Initiating Condition: FIRE potentially degrading the level of safety of the plant.

Operating Mode Applicability: A41ALL Ex-ampile Emergency Action Levels: (I or 2 or 3 or 4)

Note: The Emergency Director should declare the Unusual Event 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 felewkng-plant rooms or areas in Table H3:

k--C~pCII 119t 01 Pianft rc8cmS OF afeaSI Table H3: Plant Rooms/Areas Mechanical/Electrical Auxiliary Building (MEAB)

Fuel Handling Building (FHB)

Reactor Containment Building (RCB)

Essential Cooling Water Intake Structure (ECWIS)

Isolation Valve Cubicle (IVC)

Diesel Generator Building (DGB) switehyard Turbine Genefator Bidn IviccnafflefiýAerizcrial pAumxifiarvH1

"'ia Fuel Handling Butilding Reaeter-Crentainment Building J J I I

-t.

I C 1 -

sol-'tio HlI.'c Cuiclee Diesel Generoicr Build"in2 Gifeulating Wawc Intake Strueturc (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 f4!4ewinrgplant rooms or areas in Table H3:

135 1 P a g e

(S40z Spftifie lict @f plant rzzmS OF aFRca)

Switehyafd TbneGener-ator Bui ldinig MccaniabElctrialAuxiliar,'ý Bulijlding Fuel Handling Building Reaeter-Czntninment BuildHing iselatiin Valve Cubiele Diesel Generator Building Circulating Water lntnkce Str-uetur-e AND

c.

The existence of a FIRE is not verified within 30-minutes of alarm receipt.

(3)

A FIRE within the plant or 1SFS!VW [fer-.....m w.. an

• -AFS!*

ep.-sd !tha* : Protac'cd Arca]

PROTECTED AREA not extinguished within 60-minutes of the initial report, alarm or indication.

(4)

A FIRE within the plant or 18F81 [for-plants with an 18F91.utsid. the plant Pr.te.t.d Areal PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.

EAL #1 The intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarm, indication, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarm, indication or report.

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 1361 P age

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 if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

EAL #3 In addition to a FIRE addressed by EAL #1 or EAL #2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This hasis extends

.- a FIP4 eu*rring wisihn !w PROTECTEDhl P 129f 0

!SFS! !oeated R.-uts4dr the *lan:' PROTECTED

,4494.

44491eefrpan Oihn I1SFSI boutsid-e 9he platPrpoteeted Area]

EAL #4 If a FIRE within the plant or ISFI fr plans i.h an S-- S!. outiside the pianot Potected.rea]

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-I extinguishment recovery REGOVE RY 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 e*pleeieft&-EXPLOSIONS."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in EAL #2, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period.

137 1Page

Depending upon the plant mode at the time of the event, escalation of the efir-geney EMERGENCY CLASSIFICATION -woe!dLEVEL would be via IC CA6 or SA9.

HU4: EAL-I.b. EAL-2.b Selection Basis:

The plant areas or rooms listed contain SAFETY SYSTEM equipment.

claifm.A-izAtn W, -!

REFERENCES:

1.

OPGPO3-ZF-O001, Rev. 26, Fire Protection Program

2.

STPEGS UFSAR, Rev. 16, Section 7.4, Systems Required for Safe Shutdown I

138 1 P a a e

The "site specifie list @f plant r-oems or-afeas" should speeify, th@Se Frzcm2 Or. arceas that m if;tai SA FETY3 SYýSTEM equipment.

As neted in the EALs and Basis seetieft, ineittdle the tefm ISF-SI if the site has an ISFS! outside the plant Przteeted Arca.

Irt'T A

lilA I

139J Page

HU7 ECL: Nctifiatizn of Unuzual Event UNUSUAL EVENT Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a (NO)UE.

I Operating Mode Applicability: A4IALL I Example-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 fei ity 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 emcrgcnfcy classificatin Ic.'cl EMERGENCY CLASSIFICATION LEVEL description for an NOUEUE.

HU7: EAL-1 Selection Basis:

N/A

REFERENCES:

N/A I

140IPa g-e

HAI ECL:A Ae4 ALERT Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes.

Operating Mode Applicability: A41ALL I E9mp*e-Emergency Action Levels: (1 or 2)

(1

-A OST-ILE ACTION is ee.urring er has

,rr-ed within the OWNER CONTROLLED AMA as

.epeit.d by the (site.pe.ific.ecur.ity.hit

.upe....iai)

Securit' F*or.e Super',i"er OR Actinc S

re eA-i1a.--r-. A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by ANY of the following personnel in Table H 1:

Table HI: Security Supervision Security Force Supervisor Acting Security Manager Security Manager

] L.2*A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between 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 Spewt V:U S1or1ge X

nsto_....,

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 Alei4 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 141 ]Page

aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

EAL #1-is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This in.lud.s any a.tie. dicte..d against

.n

!SF-S! that is le.ated.ut..de 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 (site spe.ifit prc-c.du...)

OPOP04-ZO-SEC4, Guidelines for Airborne (Aircraft) Threat, and Security Force Instruction SI 2700, Security Response to Airborne Threat.

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 he14.do 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 sheuld be is contained in nez public documents such a-the Security Plan.

Escalation of the emergenty el-assifiAcatizn level EMERGENCY CLASSIFICATION -wwoW4LEVEL would be via IC HSI.

HAl: EAL-1 and EAL-2 Selection Basis:

The EALs are taken from NEI 99-01. Rev. 6. For EAL-1, the positions of Security Force Supervisor OR Acting Security Manager were included because either of these positions could be activated prior to meeting this EAL. The Security Force Supervisor is a 24-hour position and the normally the Acting Security Manager is activated after an UNUSUAL EVENT has been declared. The Security Manager is also included 4o-although this position is normally activated after an ALERT.

REFERENCES:

I.

OERPO1-ZV-SH03, Rev. 12, Acting Security Manager

2.

OERPO1-ZV-TS08, Rev. 16, Security Manager

3.

OPOP04-ZO-SEC4, Rev. 10, Guideline For Airborne (Aircraft) Threat (SUNSI)

4.

Security Force Instruction SI 2700, Security Response to Airborne Threat (SUNSI)

Developer. Noteis I

142 1Page

The (Site Speeifie Seeur-ity Shift Supervisiefn) is the titic of the An shift indi;vidmal rczpenzible c euc.in oftbe oft shift seeir-ity fcrcz.

Emoir-geR6c planz anid imfplemfenlting prOeedur-es a.-e public. deeuments; theocfer-e, 9ALs sheukl net

.n rcrte Security Sensitive infeormation. This iffeludes infecrmatiecn that m~ay be adft~kaflgeoucs to a petcntial ad-veFzar-y, such as the partieularz eenccmaing a speeifie threat er-thrcat Iccatien. Seeuit sensitive infermatien shetuld be ccntained int non publie dcumcnts stteb as the Scur-ity Plan.-

With due eeftsider-ation givefn te the abeve dcvcleper. nete, EALs may contain alpha or: numbcrc Suchrefronoc ho;d;nt Gontain a r-@eogniZable d@eoiptiOn of the-even-t. F149 Oxample, an EA4. May be wofrded as "Seeur-it event 42, 45 or: 49 is rcpofted by the (site speeifie geetffit shift stuperviciafn)."

See the rclated Develeper-Nete in Appendix B, Definitions, for guidanee efn the development ef a sehcrn d-e-finiti on for the OWNEiR CONTROLLED AREA.

• T A

I

]-;, C-A q q i Q-,"

M--

P's; s 4%

t4*1 ml I tpq - A I ;, Q I

143 IPage

HA5 ECL: A4e ALERT Initiating Condition: Gaseous release impeding access to equipment necessary for normal plant n operations, cooldown or shutdown.

Operating Mode Applicability: A-lALL I E*ample-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 A-NANY of the following plant rooms or areas:

(Site Speeiffe lzt of plAnt rozmzS OF areas with cntr-y rzl6Atd mfidO applieftbility identified)

Mechanical/Electrical Auxiliary Building (MEAB)

Turbine Generator Building (TGB)

Isolation Valve Cubicle (IVC)

Fuel Handling Building (FHB)

Reactor Containment Building (RCB)

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 safety of the plant.

An A4ert 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).

144-LP a g e

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 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 merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or t-e-in-itnetinal rtingf.. ntaim.nt. (IBSR ^nly).

Escalation of the cmergecny clasifieatin level EMERGENCY CLASSIFICATION LEVEL would be via Recognition Category AR, C or F ICs.

HA5: EAL-1 Selection Basis:

The areas listed in EAL-1 amply to areas that contain equipment necessary for plant operations, cooldown, or shutdown.

REFERENCES:

1.

OPGP03-ZF-0001, Rev. 26. Fire Protection Program

2.

STPEGS UFSAR. Rev. 16, Section 7.4. Systems Required for Safe Shutdown

3.

OPOP03-ZG-0008, Rev. 56, Power Operations

4.

OPOP03-ZG-0006, Rev. 54, Plant Shutdown from 100% to Hot Standby

5.

OPOP03-ZG-0007, Rev. 71, Plant Cooldown Deveelopr. Natm:

The site speti fis elPlnni mes er areas va~

eafwy r-etiate mode appllczwiw 7 iacflhiticp sheald speeifý thonosi roomzls efr ffroan th-at enonancgim whieh rogquirr- -A Mmeanual/loal aotiet as speeified in oraigpr-eeedurons unod fer-normal plant eper-atien, cooldown and shutdev.n. Do noet ineludcroer an of n r al r emerg E

eny co enitio eucr a emer-gea na cy v etp i, c re ctie Fmea uc org.

an morgftcncy e

eperatiens). in addition, the list should speeify the plant mode(s) dur~ing whieh en"~ would be required fer eabh room or arca.

145 IPage

The liSt ShOUld n8t inoeludl rOOMS Or: arcatS fOr-whieh entry is Fecttircd selely to perfcrma aetions of An administrative or-r-eeord keeping niatur-e (e.g., normal rounmds or rotn istions).

The list nooed not ineludoe the Controffl Room if adequt cnizrd safety/ldesign features are in placo to procelude a Centroel Room evacutatien d"e to the relumise cf a- -hazardous gas. Sueh featuros may incelude, bu are not limited to, 1apability to draw air. f rom multiple air-intakes at different and scparate locations, innci and eutcr-atmospheric boundefics, efr the eapability to acquiro and manai ystv pressurcwihi the Control Roomo cnvelepe-.

if the equipment in the listed rooem or area was airceady inoperable, or-eut of ser.iee, bcforc the event A,

.+

L L.

A..*-

. I.,1*

+.

A L.

• 1 A*

L..

.. : L1, t

7 t en BE) efnef:genes, s ou

-A WAWA-A-MAR MA AN'PAt N*___ MA.'a no A-Vor-Ra

-PAPAPI PA'Atl t

At aicat iowea by I conniciat Speeiftaieauos at the time of the event.

I E(JL Assimnmcnt Afttnbutesi 3.1.2.B3 I

146 1 P a g e