ML20202G851

Proposed Tech Specs Establishing New Improved TSs (ITS) Surveillance Requirement for Performance of Periodic Integrated Leak Test of Cche Boundary & Revising ITS Bases 3.7.12 to Define Operability of Cche

ML20202G851
Person / Time
Site:	Crystal River
Issue date:	12/05/1997
From:	FLORIDA POWER CORP.
To:
Shared Package
ML20202G831	List: 3F1297-19, Application for Amend to License DPR-72,proposing Changes to Improved TSs for CREVS & to Vftp.List of Commitments,Summary of Changes,Reason & Justification for Request,Control Room post-accident Dose Calculations & Analysis Also Encl ML20202G851
References
NUDOCS 9712100123
Download: ML20202G851 (57)

Text

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4 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUh1BER 50-302 / LICENSE NUh115ER DPR-72 LICENSE Ah1ENDh1ENT REQUEST (LAR) #222, REVISION 0 CONTROL ROONI Eh1ERGENCY AND Eh1ERGENCY FILTERS ATTACilalENT C STRIKEOUT / SIIADOW PAGES Technical SpeciGcations Each change is indicatted by a shadow box.

Deletions are indicated by strikeout.

Additional and replacement text are Indicated by shading.

p PDR

CREVS 3.7,12 SURVEILLANCE REQUIREMENIS SURVEILLANCE FREQUENCY SR 3.7.12.1 Optrate each CREVS train for 31 days 2 15 minutes.

SR 3.7.12.?

Perform required CREVS filter testing in In acenrdance accordance with the Ventilation hiter with the Testing Program.

Vent 11atinn Filter Testing Program SR 3.7.12.3 Verify each CREVS train actuates to the 24 months emergency recirculation mode on an actual or simulated actuation signal.

.SR.h. 3.?. 7.".It_WTT. Vfriff'CCli0boUnddy leakigefdossTot

2. 4 m6nt.h._s

,m performance:cfnanl integrated lea 9. age teit asespecified;1nJpproyedcontrolffro% dose calculat1.ons.)

i Crystal River Unit 3 3.7-26 Amendment No. M 9

l Proceduros, Prograns and k nuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.11 Secondary Waur Chemistry Program (continuco) c.

Identification of process sampling points, which shall include monitoring the discharge of the condentate pumps for evidence of condenser in leakage; d.

Procedures for the recording and management of data; e.

Procedures defining corrective actions for all off control point chemistry conditions; and f.

A procedure identifying the authority responsible for the Interpretation of the data and the sequence and timing of administrative evants, which is required to initiate corrective action.

5.6.2.12 Ventilation filter Testing Program (VFTP)

/ nrogram shall be established to implement the following required test ng of the Control Roo'n Emergency Ventila ion S/ stem (CREVS) per the requirements specified in Regulatory buide 1.52, 4

Revision 2,1978, and in accordance with MME ANSI N510-1975 and A6PE-44609-4976 ASTM Dl3C03189l(ReTappfsedfl995){ ^

^

a.

Demonstrate for each train of the CREVS that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration end-eystem-bypass < 0.05% when tested in r

accordance with Regulatory Guide 1.52 Revision 2,1978, cnd in accordance with A M AN R N510-1975 at a thEsystem floweste of betWeen 4hM0:sfm+M4 37T500[andf4.7;8503fs.

b.

Demonstrate for each train of the CRCVS that an inplace test of the chaeceaf carb6n adsorber shows a peneirat4en-and system bypass < 0.05% when tested in accordance with and A M ANsl Regulatory Guide-l.52 Revision 2,1978l betiveeii 4?.crM9-efs N510-1989: 1975 at the system flowrate"of ~ ~~

i

• --10% 3h800landM7s350;cfs.

c.

Demonstrate for each train of the CREVS that a laboratory test of a sample of the chewat harbbri adsorbor, when obtained as described in RegulatoEy Guide 1.62, Revision 2g 1978, letetsithijlaboritoryItost ingirlteriaiofJASINID!38034 89pR44pproVedil995 pat aLtemperaturolof 30*Cf andtrelative hue ditylofi95@ith;sethy1 %did0penet'ratibngofilassithah h5f,( thews-the-methl-4ed4de-*enet+at4+*-less--Gen-4%-* Wen (continuod)

'.0-18 Amendment No. 149 Crystal River linit 3 s

Procedures, Prograns and Manuals

\\

i 5.6 5.6 Procedures. Programs and Manuals tested-4*-euerdance-wiu-14Me-4-+f-itegulatory-Guide h 52T Rev4 Mon-E-and-ASME-NSO9-1976-at-*-temperature-of-80'C :nd 7M-telet4ve-huMdityr 5.6.2.12 VFTP (continued) d.

Demonstrate for each train of 4e CREYS that the pressure drop across the combined roughing 7 filters; liiPA filters and the charcoal barbon adsorbers is 1-AP-6E s APa4" water @6gs when tested in accordance with Regulatory Guide 1.52, Revision 2, 1976, and ASME ANSI N510 1975 at the system flowrate of betweel 4h600-efm-+-lM 37,800; and 47,850: cfm.

The provisions of SR 3.0.2 and Sk 3.0.3 are applicable to the VFTP test frequencies.

5.6.2.13 Explosive Gas and Storage ?ank Rsdioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Radioactive Waste Disposal (WD) System, the quantity of radioactivity contained in gas storage tanke or fed into the offgas treatment system. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, " Postulated Radioactive Release due to Waste Gas System Leak or failure".

The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, " Postulated Radioactive Release due to Tank Failures".

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the Radioactive Waste Olsposal (WD) System and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria, (i.e..

whether or not the system is designed to withstand a hydrogenexplosion),

b.

A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank and fed into the offgas treatment system is less than the amount that would result in a whole bod, exposure of 10.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

(continued)

Crystal River Unit 3 5.0 19 Amendment No. 449

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMHER 50 302 / LICENSE NUMHER DPR-72 LICENSE AMENDMENT REQUEST (LAR) t!222, REVISION 0 CONTROL ROOM EMERGENCY AND EMERGENCY FILTERS i

P ATI'ACilMENT C STRIKEOUT / SIIADOW PAGES Hases Each change is indicated by a shadow box, Deletions are indicated by strikeout, Additional and replacement text are indicated by shading.

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CREVS 8 3.7.12 C 3.7 Pt. ANT SYSTEMS 8 3.7.12 Control Room En:argency Ventilation System (CREVS)

BASEy BACKCROUND The principal function of the Control Room Emergency Ventilation System (CREVS) is to provide an enclosed environmant from which the plant can be operated following an uncontrolled release of radioactivity or toxic gas.

The CREVS consists of two trains with much of the non-safety related equipment common to both trains and with two independent,redundantcomponentssuppliedforeachinijoy items of safety related p4eee-ef equipment (Ref. 1).

The major equipment consists of the normal duty filter banks, the emergency filters, the normal duty and emergency duty 4

supply fans, and the return fans.

The normal duty filters consist of one bank of glass fiber roughing filters.

The emergency filters consist of three l mks each.

The first bank is a roughing filter similar tc d.e normal filters.

The second bank is a high efficiency particulate air (HEPA) filter. The third bank is an activated chorecal d erher EarbonFadsorbid for removal of gaseous activity (principally i'odine). The rest of the system, consisting of supply and is not return ductwork, dampers, and instrumentation,fedundsrit designed with redundant components. HowevirD dampersfare:provided}foFfsolatiosof;t.he) ventilation;systeh from;theisurroundt.ng;eny_tronmenti Theventilationexhaustductiscontinuouslytespedby 5 radiation monitor RM AS, which has a range of 10 to 10 counts per minute.

The monitor is set to alarm and initiate the emergency recirculation mode of operation when the airborne radioactivity and/or area radiation level reaches two times the background count rate.

ThelC6titrol! Comp 16f Habitab(11ty~ EnV41opeT(CCHE)?isitK6 space within;the; Control Complex served by CREVS.-:This-includes Control' Comp 1Wfloorielevationssfrom'108 throuWi 180 s fee t) and ; t he Es ta li! encl os ure f rom : al ev at i on l 95 T to;198 feet G Th'elelementsLwhich/ comprise 1the:CCliErare walls F

' oorsMal oof E floors ;4 floor drains J penetration 3eal s Earid d

ventilation 11solationidampers M Together the CCHEiand!CREVS

' rovide"an enclosedtenvironmentifrom:which theiplantrean'be i

pcontrolled"followingfan;uniontrolledlyeleaseiof " ~ ~ ~

N.idloactivityfoOoxicl gas {'^~'

^

(continued)

Crystal _ River Unit 3 8 3.7 60 Amendment No. 449

o CREV5 B 3Jl.12 BASES DeiWnTsal tsliti oni?daiefsnelt heimaximsm"111ssibli?1Wakigh Into tne:CCHEbelow;which contro1{ room oper4 tor:dosniand thin; approved! limits toxic' gas ? concentrations l rema h wi,deterwine T acidal rie Ak(ag Integrated lealttest cof!the CCHE?

The: difference:between allowed andractual leakage!is ' " ~

converteditolan' allowance for breaches:(in square int:hss)_

that may. exist inithe CCHE to acronmodate numalcoperating and maintenance' activities.O Breaches in' excess 20frthe ~~

tal cul ated >4rea3 renders the1CCHE l incapable 1cf. performisgiiis function sthereforelinoperable. Routineopeningfandclosing of CCHE?dcors'foN personnel passagerandtthe'movementtof~

equipment;1siacccunted for, iii the design; calculations..JX _

continuousLinsksge of/10 cubicifeetaper< minute is assumedito accountiforithis6 Holding orJblockingidoors'opentfor.short periodslofitime1does,not constitutela breach of the CCHE;ns' 16ngias(th'e doorsJcouldtbe: closed uponrnotification5cfJa~

rad}ologicalforitoxicfg g releasef ' " """~~ ~ ^ "~'

CREV$ has a normal operation mode and recirculation modes.

During normal operatioli, t!.e system provides filtered, conditioned air to the '.ontrol complex,-4he-centel-een.ple*

cenHeteef-the-eentwl-- r =, vaHous-ethee-of-f4tes-4n-Re area-e f-the-eenteel-ne*r-end-a-ce ntreHed-acces: cre:4GA}r inc1Uding"the?coisrolled accessTarsa7on)the195Tfobt elevation.J Whed switched to thi~r# 'rculation' modes?

isolation dampeisQseitiolitiingMhFdischarge' to the controlledaccessarea-anddsolatiegitheoutsideairiistike6'

-the-systes-+eekiee-Rse44-4 r= 'cbt elde-e4*-end" ~

eee4ecehtes-f4Meeed-Me-threegh-Be :.= cre;; w 4h-the except4en-+f-thecentreMed m e s s-a reef i ni t h i i"m6de i the flescrjbeddn.3helfo11owingjparagraph, gh,the CCHEij"This;is sys teW reifi rcul ate s i fil t6 red : a t h throu

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(continued)

Crystal River Unit 3 8 3.7-61 Amendment No, M9

1 CREVS B 3.7.12 BASES BACKGROUND The control com11er normal duty ventilation system is (continued) operated from tin control room and runs continuously.

During normal operation, the outside air intake damper is open the-44:ch:r-4e-wie44e-44r-46mpee-le-open partially..heric,! relief / disc l:ahle7dampenircloseddthe the atmosp discharge te theicentrolled access: ares;1.sicpen f and the system return damper is' throttled." This configuration

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allows a controlled amount of outside air to be admitted to the control compicx.

The design temperature maintained by the system is 75'F at a relative humidity of 50%.

Thrt signals will cause the system to automatically switch to th recirculation modes of opet ation.

1.

Engineered Safeguards Actuation System (ESAS) signal (high reactor building presture).

2.

High radiation signal from the return duct radiation monitor RM A5.

3.

Toxic gas signal (chlorine or sulfur dioxide)[

1he recirculation modes isolate the (+ntreWeem CCHE from outside air to ensure a habitable environment for~the safe shutdown of the plant.

In these modes of operation the controlled access area is isolated from the centre R, eem CCHC;-end-the-roma4ning-ares; ef tho wen W -e w l+ w Upon detection of ESAS or toxic gas signals,11 this mode, the system switches to,the, normal _ recirculation mode,.

dampers ; fo rithe,L out s ide41'elint ake i and i the_ex sau s t itoi t he controlled access;areamillTautomaticallysclose,Jisolatinh end LCCHE, fron t out $ 1de r ai r; ar. change.-the-oute4de-*1rMake the eleser4selat4*t-4he-sentrol-r::: = elope-frem-evt.s4de-44r pather and the system return damper will open thus allowing air in the centrebeemplex CCHE to be recirculated.

Additionally, the mechan 4 eel-etelpment r;= *haud-fant CA fume hood exhaust f.n, CA fume hood auxiliary supply fan, and CA exhaust fan are de-energized and their corresponding isolation dampers close.

The return fan, normal filters, normal fan and the cooling (or heating) colis reaatn in operationInarecirculatingmode.

Upon detection of high radiation by RM A5 the system switches to the emergency recirculation mode.

In this mode, the dampers that form thewentrol re:r o**lepe CCHE will

~

automatically close. The mechan 4eal-equ4pment-rb; ;xh=d f*nr CA fume hood exhaust fan, CA fume hood auxiliary supply fan, CA exhaust fan, normal supply fan, and return fan are (continued)-

Crystal River Unit 3 8 3.7-62 Amendment No. 449

e' CREVS B 3.7.12 BASES tripped and their corresponding isolation dampers close.

Manual action is required to restart the return fan and The place the emergency) fans and filters in operation.

cooling (or heating coils remain in operation.

APPLICABLE During emergency operations the design basis of the CREVS SAFETY ANALYSIS and;tfie:CCHE is to provide radiation protection to the control room operators. The limiting accident which may threnenthehabitabilityofthecontrolroom(i.e.ivity)is accidents resulting in release of airborno radioact the postulated maximum hypothetical accident (MHA), which is assumed to occur while in H00E 1.

The consequences of this event in H00E 1 envelope the results for HOOLS 2, 3, and 4, and results in the limiting radiological source term for the control room habitability evaluation A fuel handling accident (FHA) may also resuL(Ref. 2).

t in a challenge to and may occur in any MODE.

control room habitability \\ty of the HHA and the MODES in However, due to the sever which the postulated HHA can occur, the FHA is the limiting accident in MODES 5 and 6 only. The CREVS ind the CCHE ensures that the control room will remain habitable

~

following all postulated design basis esents, maintaining exposures to control room operators within the limits of GDC 19 of 10 CFR 50 Appendix A (Ref. 3).

The CREVS is not in the primary success path for any accident analysis. However, the Control Room Emeroency Ventilation System meets Criterinn 3 of the NRC Policy Statement since long term control room habitability is essential to mitigation of accidents resulting in atmospheric fission product release.

LCO Two trains of the control room emergency ventilation system are required to be OPERABLE to ensure that at least one is available assuming a single failure dit abling the other train.

Failure to meet the LLO could iesult in the control room becoming uninhabitable in the unlikely event of an accident.

The recuired CREVS trains must be independent to the extent allowec by the design which provides redundant components for the major equipment as discussed in the BACKGROUND section of this bases. OPERABILITY of the CREVS requtres the following as a minimum:

a.-he-emergeneywMy-fe+r 4s-OPERABLE + AControl0 Complex EmergencyiDuty;SupplyFan?is;0PERABLE;

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(tontinued)

Crystal River Unit 3 8 3.7 63 Amenoannt Nw / 4

CREVS B 3.7.12 i

- BASES (continued) b.

MC@N]lC6iMiDeth@fisljii0 TERA 8tfi c.DZ HEPA? filtiEind MiksiiWibees466"ssMes"idseiWF7srs cetteacessivelyJets}rtettag flowPand areigap"ablelof

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Performingsthomy11tratienjfunct1onsl r9PreC4cNeiittwiirdensew sser w itiWoPfunED~W~Iiw l

brdLi&didMidtitpla sajotainee) Land ^ ~

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iU2Thi?MNE1Datiitii@iEMsOilili#

T E CCHE % shdaffitid

~"TtbsMitsifttFTifItWiTdeiFil wa'In1roofPfloornB. oor; eainsgpenetrationJsealsWand Ventlationalsulatoon amperp' shbemaintained!withintWI

' assumptions?ofthedesigncachation.c8reachestinithe f

temajnsjyapab1plotppt orejpg11tstfunctionT~ ~"~~!CCH CCHE eust4be(contro11 to' provide'assurancesthatsthe t

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1 Complet The ability to maintain temperature in the Control ddressed i

is :t :ddr::::d in thh T: hak:1 P::if te:tica, a ft b :ddr ::: P ~

3ff::echnical[pe61ficat16n?317pl87 hte:tL,f si idf 0f Tihak:14::ific:ti;;;r l

j-i i

APPLICABILITY In MODES 1 2, 3, and 4 the CREVS must be OPERABLE to ensure thal the :::tr:1,:::-le:: C$HE will remain habitable during and following a postulated DBA. During movement of irradiated fuel assemblies, the CREVS must be OPERABLE to cope with a release due to a fuel handling accident.

h ACTIONS AJ With one CREVS train ino>erable, action must be taken to restere the train to OPERABLE status within 7 days.

In-this Condition, the remaining OPERABLE CREVS train is adequate to per form the :::te:1 re:: radiation )rotection function for'is contr41Troes!personhe1{ However, t1e overall reliability reduced ~because~a~ failure in the OPERABLE CREVS train could result in loss of CREVS function. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period and ability of the remaining train to providetheregulredcapability.

+

(continued)

Crystal River Unit 3 B~3.7-64 Amendment No. 449

CREVS B 3.7.12 BASES (continued)

ACTIONS B.1 and 8.2 (continued)

In MODE 1, 2, 3 or 4, if the inoperable CREVS train cannot berestoredtobPERABLEstatuswithintheassociated Completion Time, the plant must be placed in a MODE in which the LCO does not apply.

To achieve this status, the plant and in must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

TheallowedCompletionYimesare i

reasonable based on opersting experience, to reach the requiredplantconditionsfromfullpowerconditionsinan orderly manner and without challenging plant systems.

C.1 and C.2 if the During movement of irradiated fuel assemblies ino)erableCREVStraincannotberestoredto6PERABLEstatus wit 11n the associated Completion Time, the OPERABLE CREYS train myst immediately be placed in the emergency recirculation mode.

This action ensures that the remaining train is OPERABLE, that no failures prw enting automatic actuation will occur, and that any active failure will be readily detected. Required Action C.1 is modified by a Note indicating to place the system in the emergency mode if automatic transfer to emergency mode is inoperable.

An alternative to Required Action C.1 is to immediately suspend activities that could release radioactivity and require isolatian of the eentM1-Men CCHE.

This places the plant in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

D.d If both CREVS trains are inoperable in MODE 1, 2, 3, or 4, the CREVS may not be capable of performing the intended function and the plant is in a condition outside the accident analysis.

Therefore, LCO 3.0.3 must be entered immediately.

L1 During movement of irradiated fuel assemblies, when two action must be taken CREYS trains are inoperable,ities that could release immediately to suspend activ radioactivity that could enter the een4+el-reem CCHE. This places the plant in a condition that minimizes the' accident risk. This does not preclude the movement of fuel-to a safe position.

(continued)

Crystal River Unit 3 8 3.7-65 Amendment No. M9

CREVS B 3.7.12 BASES (continued)

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.

Since the environment and normal operating conditions on this system are not severe, testing each train once every month adequately checks proper function of this system.

Systems such as the CR 3 design without heaters need only be operated for ;t 15 minutes to demonstrate the function of the system.

The 31 day frequency is based on the known reliability of the equipment and the two train redundancy available.

SR 3.7.12.2 This SR verifies that the required CREVS testing is aerformed in accordance with the Ventilation Filter Testing

)rogram (VFTP. The CREVS filter tests are in accordance with Regulato)y Guide 1.52 (Ref. 4) as described in the r

VFTP Program descri) tion (fSAR, Section 9.7.4).

The VFTP includes testing HE)A filter performance, chareeal-abs +rber and carbonLadforber efficiency, minimum system flow rate,bori.

the'phisical properties of the activated ehaeced car Specific test frequencies and additional information ard discussed in detail in the VFTP.

SR 3.7.12.3 This SR verifies that each CREVS train actuates to place the control complex into the emergency recirculation mode on an actual or simulated actuation signal.

The Frequency of 24 months is consistent with the typical fuel cycle length.

(continued)

Crystal River Unit 3-B 3.7-66 Amendment No. 449

l l

CREVS l

B 3.7.12 BASES (continued)

SR~3lTl1214 This'SRisifies thsLintejiHty"of thi CCHE and the 'is'sumed _

.inleakage' rates of potenttally contaminated:airi During'the emergency mode ~of 0:eration ;the CCHE is designed to'be.a closedenvironmentsaving,Ilmitedaltexchangewithits" surroundings. Performance oft 4 periodic leak test yerifles..

the continuing integrity of the'CCHE'withiniacceptable~~ f limits.

The-f requency of 24 months:it; consistent with"the typical fuel cycle length.,.The acceptance criteria;for thd.

test isilnakate that does not exceed the value contalned in the, approved _c;ose;calculati.onss ^

~

REFERENCES 1.

FSAR, Section 9.7.2.1.g.

2.

CR-3 Control Room Habitability Evaluation Report, submitted to NRC on June 30, 1987.

3.

10 CFR 50, Appendix A, CDC 19.

4.

Regulatory Guide 1.52. Rev. 2,1978.

(continued)

Crystal River Unit 3 B 3.7 67 Amendment No. M9

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Fl.ORIDA POWEll CORPOllATION CitYSTAI, RIVER (JNIT 3 DOCKET NtJMllER 50 302 /1,1 CENSE NtJMilER DPR-72 1,1 CENSE AMENDMENT REQtJEST (1,AR) #222, REVISION 0 CONTROI, ROOh! EMERGENCY AND EMEllGENCY FILTERS ATTACilM ENT C REVISION llAR PAGES Technical Specifications

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AIJACHMENT 10 11 CENSE AMENDMENT NO.

LACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following pages of the Appendix "A" Techt' cal Specifications with the attached pages. 11e revised pages are identified by amendment number and contain vertical lines indicating the area of change. The corresponding

• spillover pages are also provided to maintain document completeness.

ILU1Y1 EnlEt 3.7-26 3.7-26 8 3.7-60 0 3.7-60 B 3.7-61 B 3.7 61 B 3.7-62 8 3.7-62 B 3.7-63 B 3.7-63 B 3.7 64 B 3.7-64 8 3.7-65 B 3.7 65 B 3.7-65A*

B 3.7 650 5.0-18 5.0-18 5.0 19 5.0-19

3.7.12

)

f IURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each CREVS train for 31 days 1 15 minutes.

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I SR 3.7.12.2 Perform required CREVS filter testing in In accordance accordance with the Ventilation Filter with the Tetting Program.

Ventilation Filter Testing Program SR 3.7.12.3 Verify each CREVS train actuates to the 24 months c'nergency recirculation mode on an actual or simulated actuation signal.

's

']

SR 3.7.12.4 Verify CCHE boundary leakage does not 24 months excet:d allowable limits as measured by performance of an integrated leakage test.

d t

Crystal River Unit 3 3.7-26 Amendiaent No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.11 Secondary Water Chemistry Program (continued) c.

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condensor in leakage

• d.

Procedures for the recording and management of Mr.a; e.

Procedures defining corrective actions for ali e control

{

point chemistry conditions; and f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrativo events, which is required to initiate corrective action.

5.6.2.12 Ventilation Filter Testing Program (VfTP)

A program shall be established to implement the foriowing required testing of the Control Room Emergency Ventilation System (CREVS) per the requirements specified in Regulatory Guide 1.52, Revision 2, 1978, and in accordance with ANSI N510-1975 and ASTM D 3803-89 (Re-approved 1995),

a.

Demonstrate for each train of the CREVS that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration < 0.05% when tested in accordance with l

Regulatory Guir e 1.52, Ruiston 2,1978, and in accordance with ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

b.

Demonstrate for each train of the CREVS that an inplace test of the carbon adsorber shows a system bypass < 0.05% when I

tested in accordance with Regulatory Guide 1.52, Revision 2, 1978, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

c.

Demonstrate for each train of the CREVS that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2,1978, meets the laboratory testing criteria of ASTM D 3803-89 (Re-approved 1995) at a temperature of 30*C and relative humidity of 95% with methyl iodide penetration of less than 2.5%.

(continued)

Crystal River Unit 3 5.0-18 Amendment No.

Procedures, Prograns and Manuals 5.6 i

5.6 Procedures, Programs and Manuals 5.6.2.12 VFTP (continued) d.

Demonstrate for each train of CREVS that the pressure drop across the combined roughing filters, HEPA filters and the carbon adsorbers is 1 AP-4" water gauge when tested in accordance with Regulatory Guide 1.52 Revision 2,1978, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VfTP test frequencies.

5.6.2.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Radioactive Waste Disposal (WD) System, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system.

The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, " Postulated Radioactivo Release due to Waste Gas System Leak or Failure". The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, " Postulated Radioactive Release due to Tank Failures".

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the Radioactive Waste Disposal (WD) Systa and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria, (i.e.,

whether or not the system is designed to withstand a hydrogen explosion).

b.

A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank and fed into the offgas treatment system is less than the amount that would result in a whole body exposure of ;t 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

(continued)

. Crystal River Unit 3 5.0-19 Amendment No.

l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMHER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST (LAR) #222, REVISION 0 CONTROL ROOM EMERGENCY AND EMERGENCY FILTERS ATTACl! MENT C l

REVISION llAR AGES Bases

1 CREVS

'B 3.7.12 j

B 3.7? PLANT SYSTEMS k

B'3.7.12 Control Room Emergency Ventilation System (CREVS)_

^ ' BASES 4

--BACKGROUNO-The principal function of the Control Room Emergency-Ventilation System (CREVS) is _to provide an enclosed _

i L

environment from which the plant can be operated-following

_ an uncontrolled release of radioactivity or toxic gas.

The CREVS consists of-two trains with much of the non-safety related equipment common to both trains and with two

• ^

independent, redundant components supplied for major items of safety related equipment (Ref.1). The major equipment consists of the normal duty filter banks, the emergency filters, the normal duty and emergency duty supply _ fans, and the return fans. The normal duty filters consist of one i

bank of glass fiber roughing filters.

The emergency filters consist of three banks each.

The first bank is a roughing filter similar. to the normal filters.: The second bank is a-c high efficiency particulate air (HEPA) filter. The third bank is_ an activated carbon adsorber for removal of gaseous l

activity (principally iodine). The rest of the system, consisting of supply and return ductwork,- dampers, and instrumontation, is not designed with redundant components.

-However, redundant dampers -are provided for isolation-of _ the ventilation system from the surrounding environment.

The-ventilation exhaust duct is continuously tested by 8

j radiation monitor RM-A5, which has a range of-10 to 10 -

counts per minute. The monitor is set to alarm and initiate n

the emergency recirculation mode of operation when the airborne radioactivity and/or area-radiation level reaches two times the background count rate.

o The Control Complex Habitability Envelope (CCHE) is the space within the-Control Complex served by CREVS. This includes Control-Complex = floor? elevations from 108 through

-180 feet and the stair enclosure from elevation 95-to 198 feet. The elements _which comprise the CCHE are walls, doors, a roof,-floors, floor drains, penetration-seals, and

-ventilation; isolation dampers.- Together the CCliE and CREVS.

_ provide:an enclosed environment from which the. plant can-be controlled following an-uncontrolled release of-radioactivity or. toxic gas.

(continued)

. Crystal' River Unit:3

'B 3.7-60

. Amendment No.-

i.

e

CREVS B 3.7.12 8ASES BACKGROUND Design calculations determine the maximum allowable leakage (continued) into the CCHE below which control room operator dose and toxic gas concentrations remain within approved limits.

Integrated leak tests of the CCHE determine actual leakage.

The difference between allowed and actual leakage is converted to in allowance for breaches (in square inches) that may exist in the CCHE to accomodate normal operating and maintenance activities. Breaches in excess of the calculated area renders the CCHE incapable of performing its function, therefore inoperable.

Routine opening and closing of CCHE doors for personnel passage and the movement of equipment is accounted for in the design calculations. A continuous leakage of 10 cubic feet per minute is assumed to account for this. Holding or blocking doors open for short periods of time does not constitute a breach of the CCHE as long as the doors could be closed upon notification of a radiological or toxic gas release.

CREVS has a normal operation mode and recirculation modes, I

During normal operation, the system provides filtered, conditioned air to the control complex, including the controlled access area on the 95 foot elevation. When switched to the recirculation mode, isolation dampers close isolating the discharge to the controlled access area and isolating the outside M r intake.

In this mode the system recirculates filtered air through the CCHE.

This is described in the following paragraph.

The control complex normal duty ventilation system is operated from the control room and runs continuously.

During normal operation, the outside air intake damper is partially open, the atmospheric relief discharge damper is closed, the discharge to the controlled access area is open, and the system return damper is throttled.

This configuration allows a controlled amount of outside air to be admitted to the control enmplex.

The design temperature maintained by the system is 75'F at a relative humidity of 50%.

Three signals will cause the system to automatically switch to the recircuiation modes of operation.

1.

Engineered Safeguards Actuation System (ESAS) signal (high reactor building pressure).

2.

High radiation signal from the return duct radiation monitor RM-AS.

(continued)

Crystal River Unit 3 B 3.7-61 Amendment No.

CREVS B 3.7.12 BASES BACKGROUND 3.

Toxic gas signal (chlorine or sulfur dioxide).

l (continued)

The recirculation modes isolate the CCHE from outside air to l ensure a habitable environment for the safe shutdown of the plant.

In these modes of operation, the controlled access area is isolated from the CCHE.

l Upon detection of ESAS or toxic gas signals, the system switches to the normal recirculation mode.

In this mode, dampers for the outside air intake and the exhaust to the controlled access area will automatically close, isolating the CCHE from outside air exchange, and the system return damper will open thus allowing air in the CCHE to be recirculated. Additionally, CA fume hood exhaust f an, CA fume hood auxiliary supply fan, and CA exhaust fan are de-energized and their corresponding isolation dampers close.

The. return fan, normal filters, normal fan, and the cooling (or heating) coils remain in operation in a recirculating mode.

Upon detection of high radiation by P.M-A5 the system switches to the emergency recirculation mode.

In this mode, the dampers that form the CCHE will automatically close.

The CA fume hood exhaust fan, CA fume hood auxiliary supply fan, CA exhaust fan, normal supply fan, and return fan are tripped and their corresponding isolation dampers close.

Manual action is required to restart the return fan and The place the emergency) fans and filters in operation.

cooling (or heating coils remain in operation.

APPLICABLE During emergency operations the design basis of the CREVS SAFETY ANALYSIS and the CCHE is to provide radiation protection to the l

control room operators. The limiting accident which may threaten the habitability of the control room (i.e.

accidentsresultinginreleaseofairborneradioactIvity)is the postulated maximum hypothetical-accident (MHA), which is assumed to occur while in MODE 1.

The consequences of this event in MODE 1 envelope the results for MODES 2, 3, and 4, and results in the. limiting radiological source term for the control room habitability evaluation LRef. 2).

A fuel FHAh niay also result in a challenge to handling accident (bilityity of the MHA and the MO control room habita and may occur in any MODE.

However, due to the sever which the postulated MHA can occur, the FHA is the limiting accident in MODES 5 and 6 only. The CREVS and the CCHE ensure that the control room will remain habitable following all postulated design basis events, maintaining (continued)

Crystal _ River Unit 3 8 3.7-62 Amendment No.

-m.

4 CREVS B-3.7,12 BASES exposures to control room op(erators within the limits of GDC APPLICABLE 19 of 10 CFR 50 Appendix A Ref. 3).

SAFETY ANALYSIS (continued)

The CREYS is not in the primary success path for any accident analysis. However the Control Room Emergency VentilationSystemmeetsCrIterion3oftheNRCPolicy Statement since long term control room habitability is essential to mitigation of accidents resulting in atmospheric fission product release.

LC0 Two trains of the control room emergency ventilation system are required to te OPERABLE to ensure that at least one is available assuming a single failure disabling the other train.

Failure to meet the LC0 could result in the control room becoming uninhabitable in the unlikely event of an accident.

The required CREVS trains must be independent to the extent a'ilowed by the design which provides redundant components for the major equipment as discussed in the BACKGROUND section of this bases. OPERABILITY of the CREVS requires the following as a minimum:

a.

A Control Complex Emergency Duty Supply Fan is OPERABLE; b.

A Control Complex Return Fan is OPERABLE; c.

HEPA filter and carbon adsorber are not excessively restricting flow, and are capable of performing their filtration functions; d.

Ductwork, valves, and dampers are OPERABI.E, and air circulation can be maintained; and e.

The CCHE is intact as discussed below.

The CCHE boundary including the integrity of the doors, walls, roof, floors, floor drains, 3enetration seals, and ventilation isolation dampers must se maintained within the assumptions of the design calculations.

Breaches in the CCHE in excess of allowed ' unidentified' leakage pathway sizes as specified in approved design calculations must be controlled to provide assurance that the CCHE remains capable of performing its function.

The ability to maintain temperature in the Control Complex is addressed in Technical Specification 3.7.18.

(continued)

Crystal River Unit 3 B 3.7-63 Amendment No.

s A

CREVS B 3.7.12 BASES APFLICABILITY In N0 DES 1, 2, 3, and 4, the CREVS must be OPERABLE to ensure that the CCllE will remain habitable during and l

following a postulated DBA. During movement of irradiated fuel assemblies, the CREVS must be OPERABLE to cope with a release due to a fuel handling accident.

ACTIONS 6,1 With one CREV5 train inoperable, action must be taken to restore the train to OPERABLE status within 7 days.

In this Condition, the remaining OPERABLE CREVS train is adequate to perfe a the radiation arotection function for control room l

>erso...iel. However, tie overall reliability is reduced l

3ecause a failure in the OPERABLE CREVS train could result in loss of CREVS function.

The 7 day Completion Time is based on the low crobability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

IL Land 8.2 In MODE 1, 2, 3, or 4, if the inoperable CREVS train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. To achieve this status, the plant must be placed in at least FDDE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in H0DE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable,basedonoperatingexperience,toreachthe recuired plant conditions from fu.1 power conditions in an orcerly manner and without challenging plant systems.

C.1 and C.2 During movement of irradiated fuel assemblies, if the ino)erable CREVS train cannot be restored to OPERABLE status wit 11n the associated Completion Time, the OPERABLE CREVS train must ihnediately be placed in the emergency recirculation mode.

This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that any active failure will be readily detected.

Required Action C.1 is modified by a Note indicating to place the system in the emergency mode if automatic transfer to emergency raode is incperable.

(continued)

Crystal River Unit 3 8 3.7-64 Amendment No.

CREVS B 3.7.12 CASES ACTIONS An alternative to Required Action C.) is to immediately (continued) suspend activities that could release radioactivity ard require isolation of the CCHE. This places the plant in a l

condition that minimizes the accident risk.

This does not preclude the movement of fuel to a safe position.

R.1 If both CREVS trains are inoperable in MODE 1, 2, 3, or 4, the CREVS ray not be capable of performing the intended t' unction and the plant is in a condition outside the accident analysis. Therefore, 100 3.0.3 must be entered immediately.

f.d During movement of irradiated fuel assemblies, when two CREVS trains are inoperable, action must be taken immediately to suspend activities that could release radioactivity that could enter the CCHE.

This places the l

plant in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby syttems should be checked periodically to ensure that they function properly. Since the environment and normal operating conditions on this system are not severe, testing each train once every month adequately checks proper function of this system.

Systems such as the CR-3 design without heaters need only be operated for 2 15 minutes to demonstrate the function of the system.

The 31 day Frequency is based on the known reliaoility of the equipment and the two train redundancy &vailable.

SR 3.7milta This SR verifies that the required CREVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CREVS filter tests are in accordance with Regulatory Guide 1.52, (Ref. 4) as described in the VFTP Program description (FSAR, Section 9.7.4).

The VFTP includes testing HEPA filter performance, carban adsorber l

efficiency, minimum system flow rate, and the physical (continued)

Crystal River Unit 3 B 3.7-65 Amendment No.

a CREVS B 3.7.12 BASES SURVEILLANCE 1R_.3.7.12.2 (continued)

REQUIREMENTS properties of the activated carbon.

Specific test and l

aoditional information are discussed in detail in the VFTP.

SR_) 7.12.J This SR verifles that each CREVS train actuates to place the control complex into the emergency recirculation mode on an actual or simulated actuation signal.

The Frequency of 24 months is consistent with the typical fuel cycle length.

SR 3.7.12 1 This SR v(r,fies the integrity of the CCHE and the assuned inleakage rates of potentially contaminated air.

During the emergency mode of o)eration, the CCHE is designed tn be a closed environment laving limited air exchange with its surroundings.

Performance of a periodic leak test verifies the continuing integrity of the CCHE within acceptable limits. The frequency of 24 months is censistent with the typical fuel cycle length.

REFERENCES 1.

FSAR, Section 9.7.2.1.g.

2.

CR-3 Centrol Room Habitability Evaluation Report, submitted to the NRC on June 30, 1987.

3.

10 CFR 50, Appendix A, G0C 19.

4.

Regulatory Guide 1.52, Rev. 2, 1978.

=

Crystal River Unit 3 B 3.7-65A Amendment No.

CREVS L

B 3.7.12 l

l BASIS THIS PACE INTENTIONALLY LEFT BLANK Crystal River Unit 3 B 3.7-65B Amendment No.

l A

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICF.NSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST (LAR) t/222, REVISION 0 CONTROL ROOM EMERGENCY AND EMERGENCY FILTE.RE ATTACIIMENT D CONTROL ROOM POST-ACCIDENT DOSE CALCULATIONS l

1 S Nuclem Regulatoiy Commission Attacament D 3F129719 Pagei NITACilMENT D CONTROL ROOM POST-ACCIDENT DOSE CA1.CULATIONS Ewns This attachment provides a discussion of calculations for post-accident dose to the control room operator. Specifically, iodine and beta skin doses are assessed. The Maximum Hypothetical Accident (MHA) is taken to be the limiting event from the perspective of Control Room habitabili'y and is initially assessed MHA with loss of offsite power (l.OOP) and MHA without LOOP secnarios are considered. The Steam Generator Tube Rupture (SGTR) is also assessed to evaluate events which might require radiation morutor PMAS (for Control Complex Habitability Envelope (CCHE) isolation), and the Fuel llandling Accident (FHA) is assessed to evaluate the source terms and mode' associated with this event. Inputs for these calculations are listed.

Pretection from toxic gas events is dependent on detection and isolation of the CCHE.

Calculations demonstrate that adequate detection and isolatior, capabilities exist for the toxic gas sources located near CR-3.

Once isola; ion is achieved, leakage into the CCHE is of no consequence, even at many times the Icakage allowed for radiological events. Since the focus of this document is the determination of allowed leakage foi radiological eventx toxic gas events are not discussed fudher in this document.

Iladammtd NRC Inspections and System Readiacss Reviews conducted in 1997 during the CR-3 Design Basis Outage identified several issues which potentially impacted control room habitability. The predominant issue peitained to the validity of assumptions ibr CCHE inleakage. Other notable items cf concem included Control Room Emergency Ventilation System (CREVS) recirculation flow rate and carbon filter ediciency. Modithatiens were made 'o reduce CCHE inleakage by improving the integrity of boundary elements.

Existing boundary dampers were replaced with zero leakage models, and redundant dampers provided at all boundary damper locations. 'lhe mechanical equipment room exhaust duct was spared in place, and the CCliE penetration ibr this duct scaled. Minor CREVS design changes were made to provide alternate means of mechanical equipment room ventilation and to improve system reliability. Programmatic changes were made to ensure that the assigned etliciency of the Control Complex carbon filters was consistent with regulatory guidance.

The findings and modifications arising fiom resolution ofidentified issues required that the Control Room operator dose calculations be revised to align inputs and assumption; with plant design. The basic methodology used in these resised caletdations is consistent with that found in regulatory guidaace and utilized in previous calculations Determination of CCHE inleakage, performed by tracer gas testing, is ditrerent than the previous methodology. Other inputs have also changed and are listed in Attachment E.

The tracer gas test was conducted at a differential pressure of 0.171 inches water gauge (in wg) using ASTM E 741-1993. The test was ba.;cd en measuring the inleakage across the CCHE. Tttis

4

'U.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 2 condition is representative of the inleakage mechanism applicable to the design of the CR-3 CCllE (zone isolation system with f!!tered, recirculated air). Using tracer gas test methods, it is i

possible 19 set up a test to measure inleakage under conditions which are representative of a specific postulated scenario.

The primary motive forces which might indure a significant difTerential pressure across the CCliE are taken as wind pressures (assuming a loss of of1 site power) and ventilation systems in adjoining structures (no loss of offsite power) In the loss of ofTsite power scenario, secondary effects such as thermal effects and localized pressures induced by the operation of CREVS become relatively significant and must be considered Significantly higher differential pressure would be expected assuming no loss of offsite power, but the source term would be lower given that this would necessarily require the Auxiliary Building Filtration System to be in operation.

To fully assess limiting post-accident conditions, both scenarios must be eva!uated. This is accomplished by measuring inleakage at a known differential pressure using tracer gas inethods, then analytically adjusting this value to correspond with postulated conditions. The wind speeds responsible for the driving CCliE differ,:ntial pressure were taken consistent with Murphy-Campe assumptions regarding the post-accident meteorology. This standard method for calculating dose utilizes these wind speeds to prohce a conservati'.e estimate of the control room operator dose.

Using the wind speed to calcula'.c the differential pressure at the four periods defined by Murphy-Campe Juring the accident, ;ontrol room opera'or dose during the accident is 26.5 REM, including 22.8 sq in. of breach margin in the envelope beyond the conservative inleakage measured during the tracer gas test.

FPC is using this mechanistic inleakage approach to demonstrate the operability of the CCHE and CREVS.

Analysis of the MHA was perfonned using the post accident model described in Regulatory Guide 1 A and source terms dec ved from TID-14844. SRP 6A ano the Murphy-Campe Paper on Control Room Ventilation System Design were used as guidance documents. The following lists specific assumptions associated with control room dose calculations for the MHA. Additional information penaining to selected parameters is prosided in the discussion that follows. A detailed set ofinputs and assumptions is pimided in Tables 1 and 2.

tisittLUplLoRf0LlhCAIMA This analysis uses 102% of the rated thermal power (2619 MWth).

The containment free volume is 2,000,000 cubic feet. The sprayed volume is 1,304,000 ft' and the remainder is unsprayed.

Containment air mixing rate is equal to 2 unsprayed volumes per hour between the sprayed and the unsprayed volumes.

The core tission products released to the free volume are 25% of the totaliodine and '00%

of the noble gases.

The core ftssion products released to the sump water is 50% of the total iodine.

The iodine species fractions for the free volume are: 0.91 Elemental & 0.05 PartiMate +

0 $4 Organic, The post LOOP /LOCA containment design leakage rate is 0.25% for the first day and 0.125% for the remaining post-accident recovery period.

4

LS. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 3 Modeling includes continuoutEmergency Core Cooling System (ECCS) Icakage outside the containment building.

For the MilA w/ LOOP, modeling includes 30 mmutes of 50 gpm ECCS leakaga outside

. the containment building beginning 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA initiaticn event.

For the MHA without LOOP, credit U taken for the Auxiliary Building Ventilation System carbon filters operating at 75% emeiency.

The vaporization fraction for the ECCS leakage is not less than 10%

The iodine removal efDciency for the 2" th!A CREVS carbon filter is 95% based on meeting the testing requirements of Reg. Guide 152.

Meteorology extrapolations and control room dose modeling and calculatious (when equilibrium conditions are present) are based on the Murphy-Campe methodology for meeting GDC 19 of 10CFR50.

The containment spray elemental iodine removal cut etr is based on a decontamination factor (DF) of 100, and the particulate iodine spray removal constant is based on a DF of 50 for reducing the constant by a factor of 10.

The spray starts 124 seconds (0.03444 hours) after containment isolation which is assumed to occur instantaneously. This time is more conscrystive than tull flow time specified in Tech Spec 3.6.6.

Drring the first 30 minutes post LOOP /LOCA, there is no ferced ventiletiou flow in the Control Complex.

Fission product solids that might be in the sump water are assumed to be non volatile and are r.ot released to the environment.

T1.e sump watei volume is assumed not to be reduced due to ECCS leakage.

DiscmiottnMeksteM!Mlalintonkauhuc_C#nklicabsts 1)

Dose Conversion Facters FPC applied revised dose conversion factors for Control Complex liabitatiility dose calculations.

Specifically, FPC changed from using international Commission on Radiological Protection Publication 2 (ICRP-2), pubikhed 1959, to Internatienal Commission on Radiological Protection Publication 30 (ARP-30), published 1979, and Federal Guidance Report #l1, published 1988, for calculating thyroid dose to control room operators.

Dos calculations for internal organs such as the thyroid ate performed usir,g dosimetnc and metabolic models contained in the ICRP publications The cunent NRC Safety Evaluation of FPC's control room habitability is based on the ' Control Room liabitability Evaluation Report" submitted to the NRC on June 30, 1987. At that time, ICRP 2 methodology was used for internal dose calculations-Rcvised methods for calculating organ dose and relating organ dose to whole body dose were published in ICRP-30, and endorsed for use in this county by the Environmental Protection Agency (EPA)in Federa! Guidance Repe t #11. These documents changed the dose conversion factors that are used to convert a quantity ofinbated radioactive material to organ dose. Fct the radionuclides of concern, w of ICRP-30 / Federal Guidance

. U.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 4 Report #11 dose conversion factors results in the accident thyroid dow to be ~30% lower than previously calculated.

The statutory authority for the use ofICO30 and Federal Guidance Report #11 can be found in the Statements Of Consideration for the publication of revised 10 CFR 20 as a Final Rule. In Federal Register 56 FR 23360, published May 5,1991, and efTective June 20,1991, the usa ofICRP 30 and Federal Guidance Report #11 for calculating internal doses is discussed. The NRC addressed a public comment regarding Section 20.1204,

' Determination of Internal Exposure," by stating that: " Appropriate parameters for calculating organ doses from radionuclide uptakes can be found in ICRP-30 and its supplements. Dose factors in Federal Guidance Report #11 are also acceptable for use in calculating occupational exposures for compliance with either {s20.1 - 20.601 or with

%b20.1001 - 20.2401, except that the individual organ dose values must be used for

} 20.1 - 20.601." (Note: sf20.1 - 20.601 were the former sections of 10 CFR 20 that remained in etTect concurrent with the revised sections, G}20.1001 - 20.2401, until January 1,1994, after which %Q20.1 - 20.601 were removed from federallaw.)

CR-3 Improved Technical Specifications (ITS) include specific actisity limits for primary and secondary coolant.

Specific activity is measured and reported as DOSE EQUIVALENT l-131, which is a defined term in the ITS. The ITS definition of DOSE EQUIVALENT I-131 specifies that the thyroid dose conversion factors used for this calculation shall be those from ICRP-30.

Evaluations of postulated accidents include estimation of offsite doses that could result from radioactwe material releases. A standard assumption applied in determining the amount of radioactive material released is that the reactor coolant activity is equal to the Technical Specification limit. Since.I'OSE EQUIVALENT l 131 is used as a measure of the permissible concentration of rr.dioactive iodine species in reactor coolant, ICRP-30 is cuaently being used in the calcu'ations of ofTsite dose consequences. Therefore, the use of ICRP-30 in revised control room dose calculations is consistent with the current licensing basis of CR 3.

4 2)

Accident Analysis Software (POSTDBA & AXIDENT)

POSTDBA is an inte.nally developed proprietary computer program of Sargent & Lundy for evaluating the radiological c msequences of design basis accidents.

POSTDBA has been used to support post-accident dose analyses licensing requirements and special dose studies for more than twenty years. It was used to support the initial Byron I & 2 and Braidwood I & 2 (Comed) FSAR submittal control room doses. Since B/B uses the NUREG-0800 (Standard Review Plan or SRP) format, dose analyses referenced compliance with the SRP and Regulatory Guides. This computer program was used in the 1980s to investigate Main Steam Isoiation Valve leakage on BWRs. Studies were performed for GE Mark 11 design (Comed's La Salle County) and GE Mark 111

' design (IPC's Clinton), Zion's current control room design modification is under NRC review, and the supporting post-accident dose analyses were prepared using POSTDBA.

At the present time (last quarter of 1997), La Salle and Clinton are having their post-accident doses reanalyzed using POSTDB A.

s U.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 5 Computer program-POSTDBA performs radiological dose calculations and related analyses for the LOCA in a pressurized water reactor (PWR) or a boiling water reactor (BWR). POSTDBA was originally developed to calculate PWR control room (CR) and offsite doses in accordance with requirements and recommendations of Regulatory Guide (RG) 1.4, RG 1.109, Standard Review Plan (SRP) Section 6.4, and SRP 6.5.2.

This program handles containment leakage, additional gaseous leakage (purge and MSIV), and can model liquid leakage (constant and intermittent reactor coolant (RC) boundary releases outside containment) as a separate case in the same computer run. In addition. to the dose evaluation, POSTDBA calculates the time dependent airborne concentrations of iodine (using spray removal modeling which handles all SRP 6 5.2 requirements and recommendations) and noble gases for the containment at the CR intet ar.d in the CR volume. The control room's potential outside air intrusions, iodine filtration, and gamma body, beta skin, and thyroid dose computations are based on the Murphy-Campe approach using time dependent integration techniques. The site meteorology can be entered as a joint frequency table, as predetermined 5th percentile x/Q values or as effective wind speeds as determined by the Murphy-Campe methodology.

POSTDBA is constmeted to allow the user to select the time steps and to control variable parameters for each time step Initial DBA isotopic iodine and noble gas sources (starting at time - 0.0 seconds) are individually entered, element family release fractions can be applied, and the iodine can be subdivided into specific fractions of chemically ditTerent types Any one or all of the three release pathway rates and the containment spray removal factors can be specified for each time step. Control room ventilation inputs include outside air makeup, recirculation, and unfiltered inleakage.

Separate filter elTiciencies can be specified for the makeup and recirculation filtert POSTDB A was revised in 1994 to make the iodine dose conversien factors referenced in Federal Guidance Report #11 (ICRP-30) the default convession factors. This program will allow manual entry of any of the default iodine conversion factors and/or depth dose conversion factors for gammas and betas Thus, if surface gamma or beta doses are needed, they can be determined using POSTDB A.

Similar to POSTDBA, Computer program AXIDENT is NUS - SCIENTECH proprietary software developed for radiological dose calculauons and related analyses consistent with

~

approved regulatory guidance and recommendations.

The AX1 DENT Code was developed in the mid-70s by NUS with the primary function to assess the habitability of control rooms during design basis accidents. The AXIDENT code has been used by NUS over the last 17 years (also used by SCIENTECH Inc. since the acquisition of NUS in 1996). In the early 80s, the AXIDENT code was used to prepare a large number of control room habitability studies that were submitted to the NRC in response to NUREG-0737, Action item Ill.D.3.4. These original habitability studies typically contained the AXfDENT manual (NUS-1954) which includes a detailed discussion of the code derivadon and design inputs (an example being CP&L's Brunswick and 11 B. Robinson subndttals). Since the original habitability submittal, the AXIDENT code has been used to resolve habitability issues throughout the industry, including the development of new

M.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 6 licensing basis analyses, The most recent analyses that were submitted to the NRC in 1997 were prepared for Dresden, Quad Cities, and Brunswick.

3)

Reactor lluilding Spray System The reactor building spray system was analyzed conservatively using only the spray pump recirculation rate of 1112 gpm for the entire time spray removal credit for iodine is permitted. For calculation purposes, the spray mode of operation is one header at 1112 gpm for the permitted duration of iodine removal credit. The volume of the sprayed region is 65.2% of the total building free volume and the unsprayed volume is 34.8%

The SPIRT Computer Code was used to evaluate the spray removal constants for elemental iodine. SRP 6.5.2 calculational methods were used to evaluate the spray removal constants for paniculate iodine. Organic iodine remains airborne for the duration of the accident. lodine removal censtants derived in CR-3 calculation # l-86-0002, Revision 5 was used as input ;o this analysis. Additional inputs and assumptions for RB Spray analysis are found in Tables 3 and 4.

4)

Control Room Emergency Ventilation System Recirculation Flow Rate LER 97-022 00 identified that past modifications were implemented vchich cdded resistance to Control Complex Ventilation System ductwork without fully assessing the impact on recirculation How rate. As a result, the system flow rate with clean filters is now somewhat lower than the 43,500 cfm nominal design flow rate. Previous calculations have assessed Control Room dose at values as low as 39,150 cfm (43,500 - 10%). A flow evaluation of current system perfonnance has established a calculated minimum recirculation flow rate of 37,300 with 4"wg across fouled filters This lower bound is conservatively taken forward for use in Operator dose calculations The 4" filter fouling value is less than the 6" currently refbered in the ITS and taken as the combined (llEPA and carbon) fitter fouling limit. Current procedures constrain operation within 43,500 cfin +/ 5% and ensure that the 37,800 cfm minimum flow rate requirement for dose calculations is met.

AnMysinfRUA_with LQQP SRP 6.4 identifies that locally high differential pressures at boundary damper locations within the control room sentilation system can have the effect of inducing significant leakage across the habitability boundary. The redundant dampers being installed at all CREVS boundary locaticas are tested to be bubble-tight at 15" wg, and effectively eliminates this concern for the CR-3 CCHE.

Subsequently, the only mechanism of getting unfiltered outide air flow into the controlled environment following a LOCA with a LOOP would be to induce it oy virtue of differential pressure across outside walls such as would be induced by wind pressure.

Wind speeds can be converted into differential pressure (Ap)in the following rdationship:

2 Pv c ap = 0.00642 p U inches of water column, where U is the wind speed in mph.

U.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 7 Air density "p"for this equation is conservatively taken as that at a temperature of 15" F, and is (pis = pn x Tm / Tis = 0.075 lb/ff x 530 R / 475'R =) 0.0837 lb/ff.

Thc relationship between inleakage and difTerential pressure for a fixed resistance can be conservatively expressed as Q = C (Ap)", where Q is the inleakage flow rate in cfm, with the flow coefucient C taken as 1.

For interpolation to values less than the test condition, the flow exponent 'h"is conservatively taken to be 0.5.

For extrapolation to values above the test cor dition, the use of n = 0.5 is non-conservative, and the more realistic value of n = 0.65 is taken from ASHRAE guidance. From this relationship the following equation can be obtained which determines corrected inicakage Qc at any differential pressure Ape based on test inleakage Qr at the corresponding test differential pressure Apr:

Oc = Qr /(Ape /Ap1)"

Finally, substituting the results from tracer gas testing (462 cfm at 0171"wg) yields the equation which predicts inleakage of the CR-3 CCHE at any differential pressure:

Q, = 462 x (Ap2 / 0.171)"

Since wind pressures are assumed to be the rimary motive force under MHA w/ LOOP conditions, inleakage fbr this scenario is determined by examining meteorological conditions associc.ted with event analysis. SRP 6.4 methodology assumes post-accident meteorological conditions corresponding to the 5% x/Q value during the critical initial stages of the event in order to minimize dispersien of the radioactive plume as it is carried fra the containment i

building to the Control Complex. The methodology then allows for three incrementalincreases in wind speed and direction over the duration of the accident due to the extreme improbability that these initial wind conditions would be sustained over an extended period of time.

11ased on these considerations, inleakage values are derived for each of the four time intervals over which x/Q values vary by correcting inleakage at the test difTerential pressure to the ditTerential pressure induced by the wind speed associated with that interval. These wind induced difTerential pressures were conservatively calculated using ASHRAE methods. Each of these inleakage va!ues is an input into the appropriate interval in the revised radiological dose j

calculations such that the wind speed associated with plume dispersion corresponds to that which drives inleakage through the Control Complex boundary.

I For the MHA w/ LOOP, it is noted that the use of!cw wind speeds provides relatively small j

motive force for inducing leakage through the CCHE. However, parametric studies show that, over the range ofinterest, increased wind speeds will tend to lower Control Room dose when it is applied uniformly to both x/Q values and building difTerential pressure. It is also noted that, at thesc. relatively low wind speeds, the potential effects of thermally indaced inleakage becomes significant.

DitTerential pressure across walls induced by ditTerences in inside and outside temperatures (i c., stack effect) can be presunced in tMI structures, as its magnitude is basically a

S Nuclear Regulatory Commission Attachment D 3F1297-19 Page 8 function of the difference in temperatures across a wall and the ditrerence in height from a given penetration to the building's neutral pressure level The temperature gradient between the Control Complex and adjacent areas at the outset of an accident would be relatively small. Given a source term model wherein the majority of exposure occurs during the initial stages of the event, leakage induced by the stack efTect would be minimal during this critical period.

The neutral pressure level of a building wall tends to be towards the elevation containing the largest leakage area, or in the case of uniform leakage, at the vertical center of the buildin, The majority of CCIIE penetrations are at or near the elevation of the cable spreading room, which is itself just below vertical center of the Control Complex clevation. Since the stack effect results in no appreciable differential pressure at the neutral pessure level and differential pressures which increase with distance from the neutral pressure level, the distribution of CCHE penetrations would tend to minimize the inleakage due to the stack efTect.

At higher wind speeds, the inleakage induced by wind pressure is dominant and stack etTect pressure provides a lesser relative contribution to inleakage.

For MilA w/ LOOP, the contribution of stack effect to inleakage was conurvatively considered by calculating stack etTect pressures during both winter and summer conditions. A uniform temperature of 31 F was assumed in adjacent areas for winter conditions, while 118 F was used tc, assess summertime conditions The Control Complex itself was assumed to remain at its design temperature of 75 F. These values were conservatively assumed to remain constant for the duration of the 30 day accident. An average stack efTect pressure was calculated, and inleakage associated with this value determined by application of relationship between building dilTerential pressure and inleakage derived from tracer gas test results. This value was then combined with wind induced inleakage using an ASHRAE formula, with a 10 cfm allowance added for access / egress, as follows:

Q... = (Q.2, q,2)" + 10 The MHA w/ LOOP analysis also gave consideration to CCHE leakage which might be induced by localized high and low pressure areas ir.duced withm the CCHE boundary by the operation of the CREVS. Obviously, any leakage which occurs as a result oflocal high pressure areas within the CCllE would be outleakage, and of no concern with regard to control room dose consequence. It is also reasonable to assume that inleakage caused solely by virtue of low pressures within rooms or elevations of the CCHE due to ventilation system operation would be induced into the system return ducting.

CREVS distributes air thmugh common supply and return headers From the common headers, discharge and return branches service each elevation individually to provide heat removal for operating equipment. It is significant that the majority of CCHE penetrations exist on the lower elevations (from the cable spread room elevation down ) Therefore, a relatively small percentage of CCHE inleakage occurs on the control room elevation, Air leaking into lower elevations is

b.S. Nuclear Regulatory Cammission Attachment D 3F1297-19 Page 9 induced into the ventilation return duct and passed through the carbon filters before reaching the control room The assumption that a_ sign!Qgar!Lp.pgentage of ajl CCHE in!cakage is actuaHy R!1gMdanpid nql.icxualinig, liowever FPC has treated all leakage due to wind pressure, stack effects, and access / egress as untiltered inicakage. This is extremely conservative.

FPC has not quantified the leakage induced by operation of CREVS, however it is conservatively accounted for as follows. Inleakage induced by CREVS operation is assessed by including an additional penalty of 125 cfm of filtered inleakage. Given that testing was performed with CREVS in operation such that this effect existed at the time, cla>sification of any portion of inleakage as filtered for this reason could be taken as a reduction in unfiltered inicakage. Instead, this filtered inleakage penalty is superimposed on unfiltered inleakage due to wind pressure, stack effects, and access / egress, and is applied for the entire 30 day duration of the event. This again is

- extremely conservative treatment ofinleakage assumptions since the penalty for filtered inleakage is taken both directly in tracer gas test measurements and analytically superimposed again in dose calculations.

AnalysMMUAsithnut. LOOP Given the occurrence of the MIIA without a loss of ofTsite power, the ventilation systems in adjacent buildings are assumed to continue to operate during and after the accident. Increasing levels of radiation in the Auxiliary Building as sensed by radiation monitor RM-A2 would result in a trip of the Auxiliary Building Ventilation System (ABVS) supply fans, resulting in a significantly greater negative pressure in the Auxiliary Building. The Turbine Building is considered to be essentially at atmospheric pressure due to the numerous large openings in that structure. Under these conditions, the post-accident leakage into the Control Room could be significantly higher (especially during the early time steps) than that postulated on the basis of wind pressures (i c.,

Mil A w/ LOOP)

The release path for this scenario is based on the activity being released from the Containment and subject to initial dispersion as it travels to the Turbine Building Ventilation System intakes and into the Turbine Building. From that point it ultimately enters the Control Room as unfiltered inleakage by the difTerential pressure induced across the Control Complex. This release path model considers dilution into the large Turbine Building volume as well as minor decay and holdup while the activity is in the Turbine Building.

The evaluation of MilA without L.OOP has four distinct changes from the version of the event which assumes LOOP; (1) given that the ABVS must be in operation to induce the

• ferential pressures of concern, then filtration by the ABVS carbon filters occurs and there is no requirement to assume an ECCS pump seal failure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident with a leak rate of 50 gpm for 30

- minutes,

~(2) the normal ECCS leakage which does occur is assumed to be filtered to 75% etliciency,

=

b.S. Nuclear Regulatory Commission Attachment D 3F129719 Page 10 (3) the activity will enter the Control Room via the Turbine Building and as such will be subject to some delay due to the buildup and decay in the volume of the Turbine Building, and (4) inleakage will be constant for the duration of the accident and will not be affected by the wind speed used in the dose analysis. This is conservative in that the wind direction n:cessary for transport towards the Turbine Building would tend to oppose inleakage through the CCllE towards the Auxiliary Building Temperature effects in this scenario are assumed to be insignificant given that continued operation of adjacent ventilation systems minimizes the temperature diH'erentials between these areas and the Control Complet The 75% elliciency assumed for the Auxiliary Building carbon litters is consistent with that recently allowed for these filters by the NRC in control room habitability analyses. Inleakage induced by CREVS operation is also ignored on the basia that conditions at the time of tracer gas testing are similar to those postulated under these post accident conditions, such that this factor was present during the tests As with the MllA w/ LOOP, analysis of this scenario assumes that inleakage is distributed evenly throughout the CCllE volume. CREVS design makes it probable that very little inleakage is introduced into the Control Room from the floors below it without being subject to filtration Given this and other conservatisms in the analysis, the treatment of MilA without LOOP described above is considered to be a very conservative treatment of this scenario.

ResthufM&iAnlyses The results of this analysis shows that the bounding version of the MilA is that associated wit'i the accident occurring with LOOP. Calculations show that a 26.5 REM dose limit can be 2

maintainec' in this scenario while allowing an a ilitional CCllE breach area of up to 22.8 in. The 26,5 REM value corresponds with that in the Control Room liabitability Evaluation report dated June 30,1987 (as referenced in the ITS Bases), and the NRC's SER in reply dated May 25,1989.

2 It is concluded that, given that CCHE breach ereas are maintained below the value of 22.8 in,

the level of CCilE integrity is suflicient to meet operabiluy requirements pertaining to radiological consequences of the MilA, AtlaI sisaffthcLDlWi J

A review of other design basis accidents for which CR-3 is licensed was performed to verify that the MilA as analyzed above is the limiting event. This review was based on (1) a review of source terms, (2) a review of the means by which isolation of the CC11E is achieved, and (3) conside ation of plant operating conditions (i.e., operating MODES) at the time of the event.

This review found that the MIIA source term exceeds that associated with all other DBAs as analyzed in Rev 23 of the Final Saf5ty Analysis Report (FSAR). Ilowever, MilA accident analysis assumes that CREVS boundary dampers are isolated essentially from the outset of the event by virtue of the 4 psi Reactor Building Ifigh Pressure Engineered Safeguards (ES) signal.

Since other DBAs might not proside isolation by this signal, a review of events which could rely on the radiation monitor or operator action to isolate has been performed. Based on this review, a detailed analysis of the Steam Generator Tube Rupture event was performed which demonstrated that isolation of the CREVS was r.ot necessary to maintain operator exposures less

U.S: Nuclear Regulatory Commission Attachment D 3F129719 Page11 than regulatory limits Further, given any reasonable isolation time either by the radiation monitor or operator action, the MHA remains the bounding event with regard to control room habitability.

The inputs, source terms and dose consequences of the SGTR, as analyzed, are presented in Tables 6 through 9.

e

.S. Nuclear Reguletory Commission Attachment D 3F1297-19 Page 12 Table 1 Significant Core lodine and Noble Gas Fission Products at the Start of a DB-LOCA(MilA)-

2619 MWT ISOTOPE FISSION Ci/MWT REACTOR HUILDING YlELD T=0 AIRHORNE INVENTORY in Ci 1131 0.029 2.508E+4 -

6.5685E+7 l-132 0.044 3.806E+4 9.9679E+7 l-133 0.065 5.622E+4 1.4724E r8 I-134 0.076 6.575E+4 1.7220E+8 I-135 0.059 5.103E+4 1.3365E+8 KR-83M 0.0048 4.152E+3 1.0874E+7 KR-85 0.0029 4.102E+2 1.0743E+6 KR-85M 0.015 1.297E+4 3.3968E+7 KR-87 0.027 2.335E+4

6. I 154E+7 AR-88 0.037 3.200E+4 8.3808E+7 KR-89 0.046 3.979E+4 1.0421 E+8 XE-3 IM 0.0003 2.595E+2 6.7963 E+5 XE-33M 0.0016 1.384E+3 3.6247E+6 XE-133 0.065 5.622E+4 1.4724E+8 XE-35M 0.018 1.557E+4 4.0778E+7 XE 135 0.062 5.363E+4 1.4046E+8 XE-137 0.059 5.103E+4 1.3365E+8 XE-138 0.0552 4.775E+4 l 1.2506E+ 8 W

.2

U.S. Nuclear Regulatory Commission ~

- Attachment D 3F129719

_ Page 13 s

. TABLE 2 =

List of Assumptions and Parameters to Model the Maximum Ilypothetical Accident for Control Room liahitability Analysis Parameter Value Thermal Power (MWt) 2619 Containment Free Volume (fP) 2 0 x 10'

~

% Sprayed Volume (ff) 65.2(1.304 x 10')

% Unsprayed Volume (fl')

34.8(6.96 x 10')

lodine Fraction Initially Dispersed In Sprayed Volume 0.652 lodine Fraction Initial Dispersed In Unsprayed Volume 0.348 Air Turnover Unsprayed to Sprayed Volumes 23,200 cfm Air Turnover Sprayed to Unsprayed Volumes 23,200 cfm Fraction of Airborne lodine Activity Released From the Core 0.25 Fraction of Airborne Noble Gases Released From the Core 1.0 Fraction of Sump lodines Released From the Core 0.5 Elemental lodine Species (%)

91 Organic lodine Species (%)

4 Particulate lodine Species (%)

5 Maximum Decontamination Factor For Removal of Elemental 100 lodines by Sprays Maximum Decontarnination Factor For Removal of Particulates 50 Maximum Decontamination Factor For Removal of Organics O

Containment Spray Flow Rate-One Header (gpm)

I112 Spray System Actuation Time Post LOCA(Seconds) 124 lodine Remova! Cutoff (hr) 4.4 Time to Sump Recirulation (Min) 29 95 Elemental Iodine Removal Constant br

20.46 (To a DF of 100)

Particulate Removal Constant hr

2.21 (To a DF of 50) 4 Particulate Removal Constant hr 0.221 (Afler a DF of 50 for 2.0I hours)

Containment Leak Rate (%/ Day) 0-24 hr 0.25 Containment Leak Rate (%/ Day) 1-30 Days 0.125 Recirculation Loop Leakage-Operational Leakage (ce/hr) 4510 cc/hr for duration of the accident Recirculation Loop SRP Assumed Leakage 50 gpm for 30 Minutes Starting 24 Hours After Accident Sump Liquid Volume Post-LOCf.t' 45,902 Fraction of Recirculation Loop Leakage Flashing to Steam (%)

10

=

IJ.S. Nuclear Regulatory Commission Attachment D l

. 3 F 1297-19 Page 14 TABLE 3 Summary ofinput Parameters Used for Iodine Spray Removal Analysis

~

Input Parameter

'alue

~

3 Total Containment Free Volume 2.0 x 10' R Sprayed Containment Free Volume 65.2% (1.304 x 10'

$R)

Unsprayed Containment Free Volume 34.8% (6.96 x 10' fP)

Spray Nozzle Type SPRAYCO Model 1713 A Spray Distribution See Table 2 Number of Drop Sizes See Table 2 Mean Spray Fall lleight-One lleader Model 110.5 f)

Spray Flow Rate. One lleader Model 1112 gpm Collection Drop Efliciency 1

Elemental Iodine Partition CoeDicient Standard Review Plan 6.5.2 Normi.! Temperature at Which Spray Water is Stored (40 100)"F Maximum Post -Accident Sump Temperature 275*F 2

Laminar Boundary Layer Surface Area 4084 R 2

Turbulent 11oundary Layer Surface Area-One lleader Model 57,708 R Water Wall Flow Fraction 0.1 A T Across Wall! Gas Boundary 1.0 F 3

Liquid Volume of Containment Sump 45,902 R Containment Wall Surface Area impacted by Sprays-One 37,900 R*

licader i,1odel Containment Radius 65 A I

--.~_________m.___

b.S. Nuclear Regulatory Commission Attaclime.nt D 3F1297-19 Page 15 I

TABLE 4 Spray Distribution for SPILWCO MODEL 1713A Nor21e Data Point No.

Drop Size (cm)

Relative Frequency (fraction) 1 3 75-3

.011 2

6.25 3

.027 3

8.75-3

.056 4

1.125-2

.105 5

1.375-2

.095 6

1.625-2 0E0 7

1.875 2

.070 8

21252

.051 9

2.375-2

.066 10 2.625-2

.044 11 2.875-2

.026 12 3.125-2

.022 13 3.375-2

.017

.020 14 3.625-2 15 3.875-2

.023 16 4.125-2

.011 17 4.375-2

.011 18 4.625-2

.015 19 4.875-2

.012 20 5.125-2

.013 21 5.375-2

.011 22 5 625-2

.016 23 5.875-2

.012 24 6.125-2

.009, 25 6.375-2

.008 26 6 625-2

.007 27 6.875-2

.01I 28

_7.125-2

.009 29 7.375-2

.011 30 7.625-2

.009 31 7.875 2

.008

.007 32 8.125 2

.006 33 8.375-2 34 8.625-2

.006 35 8.875-2

.008 36 9.125-5

.006 37 9.375-2

.005

(J S Nuclear Regulatory Commission Attachment D 31:1297 19

age 16 TAllt.E 4 (contint.ed)

Spray I)istribution for SPitAYCO SiODEl,1713A Nonle

~

llata Point No.

Drop Slic (cm)

Relative Frequency (fraction) 38 9 625-2

.005 39 98752 005 Ib 1.013 l_

.004 41 1 038-1 005 42 1.063-1

.004 l

43 1.088 1

.005 44 1.I13 1

.005 45 1.138-1 005 46 1.163 1

.004 4~

1.188 '

005 48 1213-1

,005 49 1238 1

.007 50 1.28M 1 005 51 13131

.002

~

52 l338-1

. 00 2_ __

53 1.413 1

.001 54 1438 1

.001 55 1.613 1

.001 Sc 1738 1

.002

IJ S. Nuclear Regulatory Commission Attachment D 3F129719 Page 17 TABl.E 5 1.ist of Assumptions and Parameters Used to Model the Control Room for the Controllloom Dose liabitability Analysis Parameter Value hiode of Operation Zone isolation With Filtered Recirculated Air Afbr 30 hiinutes

=

__liabitability I!sivelope Free Volume (if) 364,922 3

Control Room Free Volume (fl 1 88,000 Unlittered Infiltration Rate (SCFht) 0 8 hrs 153 0 _

8 24 hrs 247.0 1-4 days 377.0 4 30 days 8320 Filtered Recirculation Flow Rete (SCFht) 37,800 Recirculation Carbon Filter lled Depth (Inch) 2 Filter lifBciency for lodines(%)

95 Control Room X/Q values (sec/m')

0-8 hrs 9.00 x 10" 8 24 hrs 5.31 x 10" 4

l-4 days 2.03 x 107 4 30 days 5.94 x 10 Thy roid Dose Conversion Factors ICRP-30 CR lireathing Rate m'/sec 3 47 x 10"

IJ.S. Nuclear Regul: tory Commission mtachment D 3F129719 Page 18 Table 6 SGTR INPUT Parameter l

Value l Comrnents

[

Thirty Four Minute bolation Time Source Term (34 min isolation Analyis)

Reactor Coolant Pressure _

2200 psia Aveinge Temperature of the reactor 579 F coolant

~

Volume of the unsprayed region l 11 '

Assumption for instantaneously release to atmosphere Volunie of the sprayed region i ff Assumption for instantaneously release to atmosphere.

~

Projected Lontainment area of wind wake 1852 m' or 19,933 2 ff lilementnl lodine Fraction 0 91 Particulate lodine Fraction 0.05 Organic lodine Fraction 0.04 Control Room Volume 364,922 ff

[_

Purge llow rate to atmosphere 100 ff/ min Assumption for an instantaneous release to the atmosphere.

Control Room ilreathing Rate

_ 3.47E 04 m'Isec intake (c /Q') 0-2 hour 9 OE-04 sec/m'

/

0 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control room effective wind 1.2 nt sec speed 8 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> control room effective wind 2.034 m/sec speed _

l-4 day control room effective wind speed 5 320nt sec

/

4 30 day control room efTective wind 18.182 m/sec speed occupancy factor 1.0 Incorporated into the EfTective Wind Speeds Untiltered leakage into the control room 123_sfm CA!culatedrILCClllidiffetcMial pressurepf 0 201' wg.

Control room makeup air flow 5335 ff Assumed design goal of 5700 cfm less the unfiltered leakage Recirculation of air in the control room 37,800 cfm lodine Pai1ition Factor (0 - 34 minutes) 10

Release factor through the steam reliefvalves lodine Dose ConIersion Factors ICRP30 il OSTDilA Default Values Gamma Correction Factor for Control 0.0 Room Dose

U.S. Nuclear Regulatory Commission Attachment D 3F1297-19 Page 19 I

Primary to Secondary Leakage through 435 gpm affected Steam Generator I

Primary to Secondary Leakage through I spn.

unalrected Steam Generator Recirculation Filtr~ Flliciency 95% for iodine species Eight flour Isolation Analysit: Uses the above input and assumptions unless same variable is 1

shown below.

liight hour isolation source term lodine Partition Factor (0 - 34 minutes) 10

Release factor through the steam reliefvalves lodine Partition Factor (34 minutes - 8 10" Release factor through the condenser j

hours)

Table 7 3

Steam Generator Tube Rupture Source Term

)

(Doth Analyses) isotope Concentration mci /mi Kr 85m 1.54 Kr-85 8.94 Kr 81, 0.84 Kr 88 2.69 3

Xe-131m 2.40 Xc-133m 2.79 Xc 133 250.0 Xe-135m 0 93 i

Xc 135 5.96 Xe-138 0.51 l 131 3.17 l-132 4.81 J133 3.75 1 134 0 499 l 135 1.92 F

---

r---

-m-r.

,r

,-,.,-m r.-,

---w n

~-w.-

y-r

-rr=

hS. Nuclear Regulatory Commission Attachment D 3F129719 Page 20 Table 8 2

Steam Generator Tube Rupture Activity Released isotope Activity Released _

Activity Released (Ci)

(Ci) i (34 min Isolation)

(8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Isolation)

Kr-85 5.02E+02 7.08E403 Kr 85m 8.64E+01 1.22E+03 i

Kt 87 -

4.71 E+01 6.66E+02 Kr 88 1.51 E+02 2.13E403 Xe 131m 1.35E+02 1.90E+03 Xe 133m 1.57E+02

-2 21E+03 t

Xe 133,

1.40E404 1.98E+05 Xe-135m 5.22E401 7.37E+02 Xe-135 3.34 E+02 4.72E403 Xe 138 2.86E+01 4.04E+02 1-131 1.78E+02 2.51 E+03 1132 2.70E+02 3.81 E+03 1133 110E+02 2.97E+03 l

1-134 2.80E+01 3.95E+02 1-135 1.08E+02 1.52E403 4

Table 9 SGTR ACCIDENT CONTROL ROOh! DOSE (REht)

'P (without CREVS isolation)

Ilnwid Dos lyholchedy Thirty-four hiinute Steam 121E-1 8.9 LE-1 Generator Isolation Eight Ilour Steam 8.14 E-1

_l,03 E-1 Generator Isolation (with CREVS isolation initiated by Rht-AS)

Ibymid_ Dos Whqls.hedy Thirty-four hiinute Steam U 0 E-2 7.31 E-3 Generator isolation Eight ilour Steam 3.33 E-2 8.78E-2 Generstor Isolation

-,-~,-e.

u.

r

- + +

e

..w-,-rr,--

,r-----,-.-,#

r,-

-t v*v

Fl.ORIDA POWER CORPORATION CRYSTAL. RIVER UNIT 3 DOCKET NUM11ER 50-302 / LICENSE NUMllER DPR-72 I lCENSE AMENDMENT REQUEST (LAR) #222, REVISION 0 CONTROL, ROOM EMERGENCY AND EMERGENCY FILTERS ATTACllMENT E COMPARISON OF INPUTS TO CONTROL ROOM IIAlllTAllit.lTY ANAIMSES d

-e--

a

-m

-u m'

m

-u m.--.

.--m m-

13 S. Nucle r Regulatory Commission Attcchment E l

31'1297-19 Page1 ATTACllMENT E COMPARISON OF INPUTS TO CONTROL ROOM ll ABITAlllLITY ANALYSES Parameter Value in Value in Comments 6/30/87 Current Submittal Analysis Reactor 71 seconds 124 seconds Resision 3 to Calculation 186-0003 (dated 7/6/93) used a Building two minute RB spray delay time based on request from Spray FPC. Since then 186 00'.3 has been reviscd several times Actuation and uses 124 seconds as a conservative Ril actuation Time time. This value is obtained by using 120 seconds for RB spray actua' ion plus 4 seconds for RB pr.,sure to go from 0 psig to 30 psig aller a LOCA.

More realistic values for RB Spray initiation time are found in Calculation M94-0004 Rev. 0 (dated I/26/94 s, which determined the full RB spray actuation time fro n initiation, to diesel start, including block loading, pump starting, header till time md time to scach full tiow.

Calculation shows RB spray A reaching full flow in 81.1 seconds and B train reaching full flow in 86.1 seconds.

This calculation modeled the spray system completely and included all the maximum expected delay times.

Reactor 1500 gpm i112 gpm in the 6/30/871-labitability Evaluati 3,lB spr

~ v" Building described as full flow (3000 gp n). half flow (150,pm)

Spray Flow No differentiation was made between initial injection Rate and recirculation llow ntes. Reviewing OP-405 Rev.

31, RD Spray System, which was in etTect in 1987, has recirculation spray flow set at 1:50 ppm to 1250 gpm.

Calculation 190-0022 Rev. l\\ 3/12/91, determined that with Ri1 spray controller set at 1500 gpm (duriag initial injection), the actual RB spray Cow could be as low as 1397 gpm considening instrumentation error, in recirculation with RB spray controller set at 1200 gpm, the spray Cow could be as low as 1112 gpm.

Calculation 186 0002 Rev. 5,1/i6/96, determined containment spray removal constants using the riew instiument error coirected flow values of 1397 gpm (injection phase) and 1112 gpm (recirculation phase).

j Spray constants associated with the lower value_of 1112

o d.S. Nuclear Regulatory Commission Attachment E 3F1297-19 Page 2 P

gpm is used in revised dose calculations.

The instrument loop uncertainties for spray Cow indication and control were being reviewed concurrent with performing the revised dose calculations. As a contingency, the revised dose calculation looked at a containment spray now rate of 1000 gpm and found that it was essentially the same as the 1112 gpm case. The calculation concludes that containment spray rate of i

1000 gpm can be tolerated.

Balliet to Widell htter NOE97-2311 dated i1/11/97, shows that when spray is being supplied from the RB Sump, the actual now may be 121 gpm below the indicated flow of 1200 gpm. Thus, the lowest value may i

be 1079 gpm

~

Reactor 490,182 gal 343,347 gal The habitability submittal assumes the liquid sump lluilding (65,532 ft')

(45,902 n')

volume as 490,182 gallons (7.48 gal /A' or 65,532 353 Sump n'). Calculation 186 0003 Rev.1,5/2/91, referenced Volume GCI calculation DC-5515-084-1-ME, Rev. O, dated 3/26/90 that calculated new Ril sump volumes based on eliminating NaOli tanks and switching to TSP baskets (MAR 88 05 01-01). New volumes were based on cubic feet and were referenced to 130* F. New volume was determined to be 500,71S.7 gal or 66,941 A'.

Calculation 186-0003 Rev. 6,3/30/95, then switched to 45,902 n' or 343,347 gallons This figure was the output from Calculation M95-0007. An imponant design reference for Calculation M95 0007 was Calculation M95-0005, Minimum 13WST Level to Prevent Vortexing Rev. O. EOP-8 swaps from BWST to RB sump staning at 15' An instrument error of 1.2' was used in BWST level calculations. EOP 8 requires swapping over when BWST is less than 15' and has to be complete by 7' to prevent BWST vortexing. (5.5' from Calculation M95 0005) These low level considerations reduced the amount of BWST water going into the RB st:mp significantly.

Reactor 8.5 7-76 The 1987 habitability evaluation report contained spray Building solution pil Table 4.1-1, Results of Drawdown Analysis Sump for a Minimum of 6.0 wt % Sodium Hydroxide in the Additive / pil Storage Tank. This table listed Ave RB spray cases with initial spray pH and time post-LOCA for spray pH to reach 8.5. The iodine removal constants were calculated using SRP 6 5.2 Rev.1.

d S. Nuclear Regulatory Commission Attachment F 3F129719 Page 3 i

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BAW-2044,"EliminatioIof Containment Spray l

Additive", was a B & W study to determine how to convert to from NaOII storage tank to TSP. With TSP, the M ;al RB spray pil will be around 4-5 becau:e that t

is thu pil of the BWST water. After the water mises with the TSP in the RB flooded level and RB spray is swapped to recirculation, then the RB spray water pli increases to the range of 7-7.6. FPC installed the TSP baskets by hi AR 88-05-0101.

GCI revised Calculation 186 0002, Containment Spray Removal Constants (Iodine Removal) to Rev. 2 and calculated the CR-3 specific iodine removal constants using SRP 6.5 2 Rev. 2 methodology in 1991. 186-0002 Rev. 5,1/16/96, recalculated the total containment spray iodine removal constants for 1397 (1500 gpm with largest maximum negative error) and i 112 gpm (1200 gpm with largest maximum negative error). These constants are considered to refket current plant design and configuration, and are used in revised dose cales.

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hillA Source Ba*cd on Hased on The higher power rating was incorporated based on Terms TID 14844 TID 14844 recent licensing activities regarding a CR #3 power and a power and a power uprate. This action has not oeen completed, but the level of level of post accident source temi associated with the higher 2595 h1Wth 2610 htWth power rating has been incorporated into dose calculations. Since the source term is deterrnined based on a per megawatt basis per TID-14844, the use of the larger htWth rating results in e source term slightly higher than that which would be predicted with the low::r power rating. This is clearly a consers atism (not a USQ) given that the plant is still licensed to the lower value.

Auxiliary 0% cilicient 0% eflicient By letter dated September 13,1989 (3F0989 01), FPC Building in LOOP submitted a revised licensing basis for the CR-3 Loss of Filtration events,75%

Coolant Accident (LOCA) and the hiakeup System etlicient in Letdown Line Failure Accident (LLFA) ofTsite events for radiological consequences to eliminate the credit for the which Auxiliary Building Ventilation System (ABV) due to power is lack of safety grade power. FPC re-evaluated the offsite assumed to radiological consequencts of a LOCA using the same be methodology for fission product release as that used to maintained.

evaluate the CR 3 control room habitability in its June 30,1987 habitability report (3F0687-16), i.e, no credit for Auxilia:y Building filters.

+

,a.

111 Nuclear Regulatory Commission Attachment E 3F124719 Page 4 During calculational verification efforts relative to the Reactor Building (RB) flooding issue, FPC identified that the control room habitability dose is adversely effected by the change in RB ficod volume. This affect war documented in FPC letter to the NRC dated June 4, 1990 (3F0690-04). The habitability report posculates a gross failure of a passive component which causes a 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It was considered that since CR-3 does have a filtration system associated with the areas containing the Enginected Safeguards (ESF) systems and passive failures such as that postuiated to cause the 50 gpm leak have not been considered as past of the CR-3 licensing basis, the gross faihire of a passive component would not be postulated in the CR-3 control room habitability dose analyses Discussion with the NRC regarding the RB flooding issue and the aaverse effect on control room habitability dose resulted in the FPC analyse 3 including the postulated gross failure of a passive component causing a 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the ABV system in service with 75% efficient carbon filters for iodine removal (3F0690-06 aml 3F069013). The NRC documented acceptance of this in its letter to FPC dated June 21,1990 (3N0690-15) as an interim measure until the RB flooding issue was permanently resolved.

Subsequent to replacement of Sodium llydroxide spray additive solution with TSP baskets, calculctions were performed which demonstrated acceptable dose sequences without the ABVS filters and credit for their operation was discontinued.

In revised dose analyses, the ABVS filters are assumed to be operating for any event which assumes that the Auxiliary Bui! ding is at a high negative pressure.

Under these condiions, the ABVS supply fans are assumed to be t

tripped and the exhaust fans discharging through the carbon filtration system and out the stack. Differentia! pressures across the CCliE on the order of 0.20" wg would be expected, which would result in leakages considerably higher than that associated with MllA/ LOOP. Ilowever, given that the ABVS is assumed to be operating throughout the event, per SRP 15.6.5 no 50 gpm leak would be postulated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> into the event, and 'hormal"ECCS bakage would be subject to filtration Tl us, this scenario is well bounded by the MilA/ LOOP scenario with respect to Control Room ilabitability.

' O.S. Nuclear Regulatory Commission Attachment E 3F1297-19 Page 5 In the event of a MilA w/1.OOP, no credit is taken for ADVS filtration for the duration of the 30 day accidst period.

CREVS Flow 43,500 cfm 37J00 cfm 43,500 cfm i.) the original design flow rate of the Rate CREVS, and is the value used to determine IPF in the 6/30/87 habitability report. Dose consequences were (recirc mode) later evaluated at 43,500 - 10% (39,150 cfrn) corresponaing to the allowable range of oneration found in ITS Section 5.6.2.12 relative to the fiiter test program.

During a system readiness raicw it was reco;nized that previous modifications had been made which,1!oced system Ilcw rate without adequatcly assessing.

, Oct on CREVS Revised dose analyses incorporat< a value of 37,800 c6n, based on consideration of current system capabilitie$ uader dirty filter conditions CREVS 95%

95%

Filtration eflicicney has not changed, but filter te., ting has Filtration been upgraded to utilize more challenging criteria.

Elliciency Previous carbon testing was performed a'. 80 C at 30%

Ril, test program has been revised to evaluate carbon at 30 C and 90% Ril. Criteria for inplace filter te ting is nenetration and system bypass of(0.05%

TCllE/CR 355,311 (P /

364,922 ff /

Original volumes wer' based on an internal memo from Volume 85,573 ff 88,000 ff Gilbert. CCllE voimne was estimated by calcubiting the volume of the entire envelope, then subtracting 10% for internal walls and contents. Updated volumes were calculated based on a room by room survey performed by S&L for use in Control Room heat up evaluations.

CREVS/

As described As modified Figures C-1 and C-2 provide a schematie of tlie~ pre-and CCllE in the post i.;odification con 6gurations. Note that except as Configuration habitability otherwise stated, pairs of dampers replacing a single report damper receive the same control signals and act in unison, such that system logic is not changed.

Damper AllD-99, which brings supply air to the Ventilation Equipment Room is being removed and a permanent blank installed. New sapply and return registers are installed in the ductwork (164' elevation) which will now serve as the ventilation for this area.

This will eliminate AllD-99 as a potential source ofinleakage.

Existing damper AllD 12, located in the supply duct to the CA, has been removed and replaced f

with two new bubble tirbt dampers, AllD-12 and

O

' d S N'iclear Regulatory Commission Attachment E 3Fl297-19 Page 6 AllD 12D.

Existing damper AllD 2, located in the exhaust duct to the outside, has been locked open and abandoned in place.

Two new bubble tight dampers, AllD-2C and AllD-2E. were installed in series in the exhaust path. AllD 2C will be normally closed.

The position of recirculation air damper AHD 3 will be established during the process of balancing the system for the normd operating mode.

Dampers AHD 1 and AllD-lD, located in the air t

intake duct, are being disabled ud abandoned in placc. Two new bubble tight dagers, AllD lC and AllD-1E, have been installed in series on the inlet duct.

Dampers AliD-lC, AllD-2C and AllD-3 will retain positioners which provide a manual ove: ride feature.

This feature allows operators to position these dampers to modulate the outside airilow as required for purging smoke or other contaminants from the CCllE.

Mechanical Equipment Room Ventilation Air llandling Fans, AllF-21 A/ll and associated dampers AllD-24, AllD-25, AllD-26, and AllD-27 have been spared in place and the associated CCllE penetration sealed This ponion of the system originally exhausted air from the Mechanical Equipment Rcom, Elevator Equipment Room, lavatory, kitchen and toilet.

This climinates another potential source of inleakage into the CCllE.

New supply and return registers have been installed in the ductwork in the Mechanical Equipment Room. This will provide ventilation to this portion of the CCllE during both normal and recirculation modes.

A skid mounted air handling unit consisting of a fan and a charcoal filtration unit will be installed

'o ventilate the Elevator Equipment Room, lavatory, kitchen and toilet. This system is non-safety and non seismic and will vent approximately 1,000 cfm by way of a field connection to a non-safety related (NSR) duct.

I This system will not create a penetration to the outside environment.

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e

' ' L.S. Nuclear Regul-tory Commission Attachment E i

l 3F129719 Page 7 Small bore drain pipes penetrating the CCllE have been fitted with loop seals to prevent inleakage though the lines.

These have been added to a queued work request in the work controis system which maintains CCilE citain line loop seals-Vestibules have been installed over all CCllE boundaiy doors, and have been scaled to provide maQnum leaktightness Tliese vestiboles provide c. means to test individual CCilE boundary door leaktightness, as well as ieducing inleakage associated with CCllE access / egress.

In addition to the above modifications, an extensive effort was undertaken to survey CCilE penetrations and seal as required to minimize inleakage. As a result of this work, it is concluded that conduit penetrations do not pose a significant liability to CCilE integrity.

Penetrations associated with electrical cable banks were inspected and scald to the extent feasible with existing procedures and materials, but some leakage paths remain through the interstitial spaces between individual cables.

Additional work is being planned to improve the scaling of penetratione with the most significant leakage

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Estimated Measured See detailed discussion peitaining to inleakage in

'JCilE inicakage on the basis by tracer gas Attachment D.

of testing and summation analytically

' leakage past corrected to CCilE predict boundary inleakage elements per under SRP 6.4 postulated post-accident conditions Dose ICRP2 ICRP30 The NRC Safety Evaluation of FPC's control room Conversion habitability is based on the " Control Room liabitability Factors Evaluation Report" submitted to the NRC on June 30, 1987. At that time, ICRP-2 methodology was used for internal dose calculations Revised methods for calculating organ dose and relating organ dose to whole body dose were published in ICRP-30, r.nd endorsed for I

use in this country by the Environmental Protection

e 1

d.S Nuclear Regul: tory Commission Attachment E 3F1297-19 Page 8

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Agency (EPA)in Federal Guidance Report #11. For the radionuclides of concern, use ofICRP-30 / Federal Guidance Report # 11 dose cor. version factors results in the accident thyroid dose to be -30% lower than previously calculated. CR-3 Improved lechnical Specifications (ITS) include specific activity limits for primary and secondary coolant, which is measured and reported as DOSE EQUlVALENT l 131. The ITS dernition of DCSE EQUIVALENT l-131 specifies that the thyroid dose conversion factors used ror this calculation sball be those from ICRP-30 2

Sonware Accident Analysis Software (POSTDBA)

Computer program POSTDBA is Sargent & Lundy proprietary.collware which performs radiological dose calculations and related analyses for the LOCA in a PWR or a BWR. POSTDBA was originally developed to calculate PWR control room (CR) and offsite doses in accordance with requirements and secommendations of Regulatory Guide (RG) 1.4, Standard Review Plan (SRP) Section 6.4, and SRP 5 5.2., and was revised and revalidated most recently in 1994.

POSTDBA is constructed to allow the user to select the time steps and to control variable parameters for each time step The variables ir.clude containment ; pray iodine removal rates; post accident source re* case rates (iodme and noble pases) and any iodine filtration, x/Q changes; CR parameters (makeup, inleakage, iodine removal, breathing rates, and occupancy factors), plus the fractions of elemental, particulate, and organic iodine released to the environmei.t. The first and the following time steps can be used to vary most of the variables, and if needed, the firrt time step can be used to model a delayed release. This degree of user control allows other types of accidents to be analyzed.

Similar to POSTDB A, Computer program AX1 DENT is NUS SCIENTECil proprietary sottware which performs radiological dose calculations and related analyses.

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