ML15176A534

Diablo Canyon Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term - Updated Final Safety Analysis Report Markup. Part 7 of 8

ML15176A534
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/2015
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation
References
DCL-15-069
Download: ML15176A534 (66)

Text

DCPP UNITS 1 & 2 FSAR UPDATEPer Regulatory Guide 1.183, the radioiodine released from the fuel gap is assumed tobe 95% particulate (Csl), 4.85% elemental, and 0.15% organic.

Due to the acidicnature of the water in the fuel pool (pH less than 7), the Csl is assumed to immediately disassociate and re-evolve as elemental iodine, thus changing the chemical form ofiodine to 99.85% elemental and 0.15% organic.

In addition, and per Regulatory Guide1.183, an iodine decontamination factor of 200 is assumed for the SFP / reactor cavity.Noble gases and unscrubbed iodines rise to the water surface where they are mixed inthe available air space. All of the alkali metals released from the gap are retained in thepool. In accordance with Regulatory Guide 1.183, the chemical form of the iodinesabove the pool is 57% elemental and 43% organic.Per Regulatory Guide 1.183, the activity released due to an FHA is assumed to bedischarged to the environment in a period of 2 hrs (or less if the ventilation systempromotes a faster release rate).FHA in the FHBThe radioactivity release pathways following an FHA in the FHB are established takinginto consideration the following Administration Controls:

During fuel movement in the FHB:a. The movable wall is put in place and securedb. No exit door is propped openc. One FHBVS exhaust fan is operating (The supply fan flow (if operating) has beenconfirmed by design to have less flow than the exhaust fan)Operation of the Fuel Handling Building Ventilation system (FHBVS) with a minimum of1 exhaust fan operating and all significant openings administratively closed will ensurenegative pressure in the FHB which will result in post-accident environmental release ofradioactivity occurring via the Plant Vent. The activity release due to the FHA in theFHB is assumed to be discharged to the environment as follows:a. A maximum release rate of 46,000 cfm via the Plant Vent due to operation of theFHBVS with a closed FHB configuration.

b. A maximum conservatively assumed outleakage of 500 cfm occurring from theclosest edge of the FHB to the control room normal intake (i.e., 30 cfmoutleakage is assumed for ingress/egress; 470 cfm is assumed for outleakage from miscellaneous gaps/openings in the FHB structure).

It has been determined that for the FHA in the FHB, the actual release rate lambdabased on the FHBVS exhaust (i.e., 8.7 hr1) is larger than the release rate applicable to"a 2-hr release" per Regulatory Guide 1.183 (i.e., 3.45 hr1). Thus the larger exhaustrate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500cfm outleakage is utilized in the analysis.

15.5-127Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFHA in the Containment The potential radioactivity release pathways following a FHA in the containment areestablished taking into consideration

a. Operation of the containment purge system which would result in radioactivity release via the plant ventb. Plant Technical Specification Section 3.9.4 that allows for an "opencontainment" during fuel movement in containment during offload or reload.The most significant containment opening closest to the Control room normaloperation intake is the equipment hatch. The equipment hatch is anapproximately 20-ft wide circular opening in containment.

In the event thecontainment purge system ceased to operate (a viable scenario since it issingle train and has non-vital power), the density driven convective flow out ofthe equipment hatch (due to the thermal gradient between inside and outsidecontainment conditions),

could be significant.

It has been determined that for the FHA in the Containment, the release rate assuminga regulatory based 2 hr release is larger than that dictated by the containment purgeventilation system, or convective flow out of the equipment hatch. Thus the regulatory based release rate (i.e., 3.45 hr'), is utilized for this analysis.

Review of theatmospheric dispersion factors associated with the plant vent vs the equipment hatchindicates that dose consequences due to releases via the equipment hatch will bebounding.

15.5.22.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at anypoint on the boundary at the EAB, for any 2-hr period following the onset of theaccident be reported as the EAB dose. Since the FHA is based on a 2-hour release,the worst 2-hour period for the EAB is the 0 to 2-hour period.The bounding EAB and LPZ dose following a FHA at either location and at either unitis presented in Table 15.5-47.15.5.22.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. Provided below are the critical FHA-specific assumptions associated with control room response and activity transport.

Design Basis FHA (occurs at t=72 hours after reactor shutdown) 15.5-128Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATECredit is taken for PG&E Design Class I area radiation monitors located at the controlroomcontrol room normal intakes (1-RE-25/26, 2-RE-25/26) to initiate CRVS Mode 4(filtered

/ pressurized accident ventilation) upon detection of high radiation levels atthe control room normal intakes as a result of an FHA.An analytical safety limit of 1 mR/hr for the gamma radiation environment at thecontrol room normal operation air intakes has been used in the FHA analyses toinitiate CRVS Mode 4. Note that the actual monitor trip setpoint is lower to include theinstrument loop uncertainty.

The radiation monitor response time is primarily dependent on the type of monitor, thesetpoint, the background radiation levels and the magnitude of increase in the radiation environment at the detector location.

For a monitor with an instrument time constant of "T" (2 seconds) and a background of0.05 mR/hr, the response time "t" to a high alarm Setpoint (HASP < 1 mr/hr), for a stepincrease of radiation level DR (mR/hr) is determined by solving the following equationthat represents the monitor reading approaching the final reading exponentially.

HfAS?= 0.05 OLR (I. -I-0It is determined that a DBA FHA (i.e., occurs at 72 hrs post shutdown) will result in aradiation environment at the control room normal operation intakes that greatly exceedthe analytical limit of 1 mR/hr for initiating CRVS Mode 4. This will result in an almostinstantaneous generation of a radiation monitor signal to initiate CRVS Mode 4(radiation monitor response time is estimated to be < 1 sec). For purposes ofconservatism, and since the delay in isolation of the normal intake has a significant impact on the estimated dose consequences, the analysis conservatively assumes amonitor response time to the HASP of 10 secs.As discussed in Section 15.5.1.2, when crediting CRVS Mode 4, the FHA doseconsequence analyses is not required to address the potential effects of a LOOP.Thus delays associated with diesel generator sequencing are not addressed.

Therefore, the time delay between the arrival of radioactivity released due to a DBAFHA at both the control room normal Intakes (assumed to be instantaneous) and CRVSMode 4 operation is estimated to be the sum total of the monitor response time (10secs), the signal processing time (2 Secs) and the damper closure time (10 secs) for atotal delay of 22 seconds.Delayed FHA:It is recognized that the response time for radiation monitors are dependent on themagnitude of the radiation level / energy spectrum of the airborne cloud at the locationof the detectors, which in turn are dependent on the fuel assembly decay time. Thusan additional case is considered for each of the two FHA scenarios described above15.5-129Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(i.e., a FHA in the FHB and a FHA in Containment) when determining the dose to thecontrol room operator; i.e., a case that reflects a delayed FHA at Fuel Offload or aFHA during Reload, occurring at a time when the fuel has decayed to such an extentthat the radiation environment at the control room normal intake radiation monitors isjust below the setpoint; thus the control room remains in normal operation mode andCRVS Mode 4 is not initiated.

The analyses determined that the dose consequences of a DBA FHA bound thatassociated with the delayed FHA for both the FHA in the FHB and the FHA in thecontainment.

The bounding atmospheric dispersion factors applicable to the radioactivity releasepoints / control room receptors applicable to an FHA at either location, and at eitherunit, are provided in Table 15.5-47B.

The x/Q values presented in Table 15.5-47Btake into consideration the various release points-receptors applicable to the FHA toidentify the bounding Z/Q values applicable to a FHA at either unit and at eitherlocation, and reflect the allowable adjustments

/ reductions in the values as discussed in Section 2.3.5.2.2 and summarized in the notes of Tables 2.3-147 and 2.3-148.The bounding control room dose following a FHA at either location and at either unit ispresented in Table 15.5-47.15.5.22.-.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:(1) An individu,!

located at any point on the bundan,',

of the exclusion area foran" two hourI period following the onset of the porstulated fission productrelease shall not receove a total radiatieo of 0.063 Sy (6.3 Fem)total efdemtione dose equivalent (TEpDE) as shown in Table 15.5 17.(2) An individual located at any point on the outer bonundar, of the low population zone, whxoi ,ep to the radioactive eloud feslotingth fome the postulated fsinpro~duct r_1eleas~e (during the entire period of its passage),

shall notre a total Fadiation dose in excess of 0.063 Sv (6.3 rem) total effective dose equivalent (TEDE) as shown in Table 15.5 47.(3) The dose to the conrol1 room operator under accident conditionS shal! not be 4inexcess of 0.05 Sy (5 rem) total effccti':o dose equiv.alent (TEDE) forthdIuration of the accident as sho'n in Table 15.5 17.The analysis demonstrates that the acceptance criteria are met as follows:(1) The radiation dose to an individual located at any point on the boundary ofthe exclusion area for any 2-hour period following the onset of thepostulated fission product release is within 0.063 Sv (6.3 rem) TEDE asshown in Table 15.5-47.15.5-130Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(2) The radiation dose to an individual located at any point on the outerboundary of the low population zone, who is exposed to the radioactive cloudresulting from the postulated fission product release (during the entire periodof its passage),

is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-47.(3) The radiation dose to an individual in the control room for the duration of theaccident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-47.16.5.22.2 Fuel Hadling A^-=d.nt Inside Containmcnt 15.5.22.2.1 Acceptance Crit ri(1) The FadiOgicaI consequences of a fuel handling accident inside containment shal not exceed the dose limits of 10 CFR 100. 11 as outlined below:i. An individual Iocated at any point on the boundary of the exclusion areafor the tw.v hours immediately following the onset of the potulated fis ionproduct release shall not receive a total radiation dose to the whole bod-YLin exes of 6 rem or a tota! radiation dore in excess of 75 rem to thethyroid froMmeidine exposure.

ii. An individual Iocated at any point on the outer boundaI y of the Iow-population zone, who. i exposed to the radioeative cloud resulting fcoromthe postulated fission product release (during the entire period of itspassage),

shall not receive a total radiation dose to the whole body inexcess Of 6 rem, or a total Fadiatien dose in excess Of 75 rem to thethyroid from iodine exposue.(2) In accordance with the requirements of G.. 19, 1971, the dose to the conroom operator under aGcident conditions shall not be in excess of 5 rem wholebody or its equivalent to any pa~t of the body (i.e., 30 rem thyroid and betaskin, Reference

51) for the duration of the accident.

15.5.22.2.2 of Causes and Accident Des..i;ptio The offesite radiological consequences of a postulated fuel handling accident inside theContainment aFre mitigated by containmeRt crloure.

The foallowing evaluatio-n shows thatin all cases the calculated exposures would be well below limits specfd in10C FR 100.11.15.5-131Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEDrrring fuel handling opertienG, containment c",lore is not theo-,nt-ainmenet ventilation purge system is oeprationa!

and exhau.t-air from the,ontainment two 18 inch contairmert isolation valve,. These tw, aFrecoRnecte-d-in Series. ThiS flow of a'ir from the containment is dslcharged to theenronment via the plant ,entThis exhaust strem is monitFred fr activity by monitors in the plant vent. in the eventof a porstulated fuel handling acc.ident, the plant monitors will alarm and result inthe automatic closure Of containment hetlainislation valves. This activi*ty releasemay result in offsite radiological exposures.

Containment penetrations are allowed to be open durig fuel handling operations.

Themost promFinent of these penetrations are the equipment hatch and the personeairlock.

Closure of these penetration6 is achieved by manual mneans as discaussed iSection 15.4.5. The clos6ure of these penetrations is not credited in the design basis,fuel handling accident inside containment.

The E1A analysis assum-es that the Gotrol room ventilation sy6tem Of eachuremains in the norm~al moede of operation following the FHA. Thus-, the design basisFHA does not credit charcoal filtration of the control room atmosphereintake flow orreci~rculation flow.The eva.uation of potential off+ite exposures was pe.... ,ed for a design basis case,assuming plant paramneters as limited by Technical Specifications.

The assumptions ofSafety Guide 25, Ma.rh 1972, were used as guidance w..ith the exceptions detailedbelow.1j.5.2 .2.2.1 A^ti.,ity Released to Containm+nt Atmosphere The assumptions made in g the quantity of actiity for release from....the containmlent refueling pool following the postulated accident are identical to thosedscussed in Section 15.5.22.1.

For the DBA case, these assumptions

-are ,,rcnistent with those in Safety Guides 25, March 1972, and 1.18 , July2000.

Cons~istent with the guidance of Safety Guide 25, March 1972, it was assumed that allthe gap activity in the damaged rods is released aRd consists of 40 of the totalnoble gases other than KIr 85,30 of the Kr 85, and 10 percent of the tolradioactive iodine in the rods at the time of the accidetAn effec"tive D, of 200- for the io'_iAes.

was as-sumed, for the later ;n the refue!irg cavity.This DF is consistent with the current guida.ne proided In Rega , Guide 1 413,The dos covrso factors used are froem !GRP Pubblicattion 3-0 (Refaerene 45). Theuse of these dose conversion factors is consistent with the current guidance provided inRegulator,'

Guide 1.183, July :2000.15.5-132 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.6.22.2.2.2 Containment CIoSUreFlolloWing the postulated,

accident, ativity evolves from the surface of the peolwhere it m.ixes with air above the pool. Airboene actiVitY is then to bedischarged to the envronment via the open penetrations.

The- dluration of thereaewas assumed to be within two seconds.in addition to radiation moniRtor indications, a fuel handling accident would immediatelv be known to refueling perSOnnel at the scene of the accident.

These personnel wouldinitiate GCntlRl e GI IICUSCUCUI a~tiA-.IR.

WANG R-e reui y anl EguipmenltU 16910F1 ~..uideIII1 to be in constant communication with control room personnel.

The plant inteercosystem is described in Section 9.5.2.15.5.22.2.2.3 ActiVity Released to En':ironmncn The containment refueling pool is approximately rectangu!ar in shape with approximnate dimensions of 25 by 70 feet. The pool has a surface area of about 1750 square feet.it was assumed that activity evolved from the pool was in stanta neou sly mixed andretained within the approximately 33,600 cubic foot rectangular parallelcpipcd formedby the 25 by 70 foot pool and the 40 foot high steam gene raters. W~here the steamgenerators do not surround the pool, the radioactivity would actually be dispersed into alarger volume of air which would have the effect of reducing the dose. HowAev~er, forconser.atism, it was- assumed that all the radioactivity remnained within this33,600 cubic foo0t Volume and was then transperted to the environme~nt w~ithin a twosecond time period through the ope equipment hatch.15.5.22-2-2.4 Off-site Exposures The integrated release of activity to the enviFronment and the resulting effsitc radiological exposures were calculated for the postulated fuel handling accident inside containment using the LOCGADOSE=

computer programn.

Table 15.5 48 itemizes the DBA assumptions and Runuerical values used to calculate fuel handling accident radiological exposures9.

The coalculated releases of activity to theatmopher ar lised n Tale 1.5 9. Te D A ,,q exoue euting fromn the-psuatmesph aTucise na ORi Tablaer:ne con 9.Th~nen arcA e prene --n e-i iThese exposures are well within the 10 CFR 100.11 limnits.15.5.22.2.2.5 Acltion Fo-llowinf*g Containment IsolatioFollowing manual containment closure after the fuel handling

accident, activity can beremoedwe from the coant-ain~ment atmosphere by the redundant PG&E= Design Class 11iodine Remoeval Syte (to rins at 1 2,000 cfmn per train), which consists of-15.5-133Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATEHE=PA,'charcoal filters.

This system is deGcribed in Section 9.4.5. There are noTechnical Specificaio

" oquirements forF this fitration system.The contafinment can also be purged to the_ atmosphere at a controlled rate of up to3-00 c-fm per train through the HEPNAihaFcoal filters of the hydrogen purge systemn.

ThissYStemn is decriFbed in Section 65.2.5.45.5.22.2.3 Conclusions rL... ......I~...X..

.I---&....&..

IU.A LL-L --- -!i- ----.J_ ...+He an~alysis demunGATates Mat tnc arueeptance cracria are mnet as iouews';(1) The radiation dose to the whole body and to the thyroid of an indiv'idual locsated at any point on the bound-ar; of the exclusion area for the two hoursimmediately following the onset of the postulated fission product release arewell below the dose limits of 10 CFR 100.11 as shown in Table 15.5 50.(2) The radiation dose to the whole body and to the thyroid of an indiv:idual located at any point on the outer boundar,'

of the low population zone, who isexposed to the radioactiye cloud resulting fromn. the postulated fission productrelease (during the entire period ofits pasg) aewl elow the doselimits of 4 10 4R100.41 as shown in Table 15.5 50g.(3) In accordance with the of GDC 19, 4971, the dose to thecontFroroI

.o..perat under accident conditions shall not be in exGess of 5Frem whole body or its equivalent to any pa~t of the body (i.e., 30 rem thyroidand beta skin, Reference

51) for the duration of the accGident as showniTable 15.5-50.15.5-134Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE* WI IP r' rIwi ....I !---- r ..... I I I--.----I1!

.... * ! JI .... ,I.C mgit 1.. [i 1!1-i Io.e

• r!uci '. HII nan ti ff'ici e [fi 1-in the perceding sections the potential off-site exposures from mnajor fuel handlin~g accdetshave been cluae.The anRalyses have been carried out using the mo~delSand assumptions specified in pe~tinent regulator; guide,. In allw anlss h esultingpotential exposures to individual mnembers of the pulic nhe eneral population havebeen found to be lower than the applicable guidelinies an~d HFM6t specified i10 CFR 1 00.1 1. (F=HA OR Containment) anid- 10 C-F=R 50.67 (FHA in FHB).On- this basis,06 it cGa be conchuded that the occurrence of a major fuel accidentin a DCPP unit w~ould not Gonstitute an undue risk to the health and safet of the public.Adtoalit can; be concluded that the ES L eiedfF teMitigation o15.5.23 RADIOLOGICAL CONSEQUENCES OF A CONTROL ROD EJECTIONACCIDENT15.5.23.1 Acceptance CriteriaThe radiological consequences of a CREA shall not exceed the dose limits of 10 CFR50.67, and will meet the dose acceptance criteria of Regulatory Guide 1.183, July 2000and outlined below:EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area forany 2-hour period following the onset of the postulated fission productrelease shall not receive a radiation dose in excess of 0.063 Sv (6.3 rem)TEDE.(2) An individual located at any point on the outer boundary of the lowpopulation zone, who is exposed to the radioactive cloud resulting from thepostulated fission product release (during the entire period of its passage),

shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem)TEDE.Control Room Dose CriteriaAdequate radiation protection is provided to permit access and occupancy of thecontrol room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

(1) The a rod ejection accident shall not exceedthe dose limits of 10 CER 100. 11 as outlined below:15.5-135Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEi. An individual IOcated at any poeint on the boundary of the eXGlusin areafor the two hours imma'i ' following the onset of the potulated firsion l rease shall not a total radiation dose to the whole in excoss ef 25 rFer or a total radiation dos'6e in excess of 300F rem to thethyroid froM iodine exposure.

i.An individual located at any point on the outer boundary of the low-population zone, who is exposed to the radioactive cloud resulting from-the postulated fission product release the entire perFid of itspassage),

shall not receive a total radiation dose to the whole body inexGess Of -6 rem, 9r a tetal rauiauon Ise inthyroid fom iodine expeosue.

excess ei 301D reri to the(2) in accordanGe with the requirements of GD. 19, 1971, the dose to thecontrol room operator under accident conditions shall not be in excess of 5rem whole body or its equivalent to any part of the body (i.e., 30 rem) for theduration of the accident-.

15.5.23.2 Identification of Causes and Accident Description 15.5.23.2.1 Activity Release PathwayAs discussed in Section 15.4.6, this event consists of an uncontrolled withdrawal of acontrol rod from the reactor core. The CREA results in reactivity insertion that leads toa core power level increase, and under adverse combinations of circumstances, fuelfailure, and a subsequent reactor trip. In this case, some of the activity in the fuel rodgaps would be released to the coolant and in turn to the inside of the containment building.

As a result of pressurization of the containment, some of this activity couldleak to the environment.

Following reactor trip, and based on an assumption of a Loss of Offsite Powercoincident with reactor trip, the condenser is assumed to be unavailable and reactorcooldown is achieved using steam releases from the SG MSSVs and 10% ADVs untilinitiation of shutdown cooling.

DCPP has established that the LOL event generates themaximum primary to secondary heat transfer and the CREA assumes these sameconservatively bounding secondary steam releases.

Regulatory requirements provided for the CREA in pertinent sections of Regulatory Guide 1.183 including Appendix H is used to develop the dose consequence model.Table 15.5-52A lists the key assumptions

/ parameters utilized to develop theradiological consequences following a CREA.The CREA is postulated to result in 10% fuel failure resulting in the release of theassociated gap activity.

Per Regulatory Guide 1.183, the core gap activity is assumedto be comprised of 10% of the core noble gases and halogens.

A radial peaking factorof 1.65 is applied to the activity release from the fuel gap.15.5-136Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEIn accordance with the requirements provided in Regulatory Guide 1.183, twoindependent release paths to the environment are analyzed:

first, via containment leakage of the fission products released due to the event from the primary system tocontainment, assuming that the containment pathway is the only one available; andsecond, via releases from the secondary system, outside containment, following primary-to-secondary leakage in the steam generators, assuming that the latter pathwayis the only one available.

The actual doses resulting from a postulated CREA would be a composite of dosesresulting from portions of the release going out via the containment building and,portions via the secondary system. If regulatory compliance to dose limits can bedemonstrated for each of the scenarios, the dose consequence of a scenario that is acombination of the two will be encompassed by the more restrictive of the two analyzedscenarios.

ulnd4er ad'v.Frse

,mbinations Of circ-umstances, omen fuel -.addIng failureS could occurfollowing a rod ejection In this case, some of the actiVitY in the fue! rod gapswould be released to the coolant and in turn to the inside of the containm~ent building.

As a result Of pressurization of the containment, some of this acti',,tY could leak to theenafonment Computer code RADTRAD 3.03, is used to calculate the control room and site boundarydose due to airborne radioactivity releases following a CREA.ruiF uIF -eir 616 Gce, Was wu s6UW~eu --i I~ -.ait HUu bee.. _pe-...'continuously With 1 pecn ulcadn efects and 1 gpmn pri~nar, to secoenda:y-leakage.

Fer the expected ca-e calculation, operation at 0.2 percent defects and20 gallons per day to the secondary was assumed.Follo9Wing a postulated rod ejection

accident, activity relearsed from the fuelelle laddin na uet failure of 1 nercIntII of;ll the fuel rod islU asume to beinrstantaneouslY released to the 'r mr; colant. Releases to thefprimar,'

coolant areassumed to be immediately and uniformly mnixed throughout the cooelant.

The actiVity released to the containment ftrom the primP,'y coolant through the rupturedonto od mechani pressue housing is assumed to be mixed instantaneousl-;

throughout the containmepnt and is available for leakage to the atmo-sphere.

It has been assumed for both the design basis and expected cases that 10 percent ofthe elemental i;,ono leaked to the rcoolant is released to the containment atmosphere asa result of flashing Of some of the coolant water. Of the amorunt Of noblegases released to tho primary coolant, 100 prcent is as.umed to be released to thecontainment atmosphere at the time of the accident.

it is assumed that the amount of"4;" edneO cheMical fcrms that a.e net affe"te, by the sp,,.y sy rem ~ elgbe hsrelease fractions arc used for both the de5ign basis case and the expected case.15.5-137 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFollowing the release to the containment, the fission products are assumed to leak f7rothe col-ntainment at the same rates assumed for the ..., LhA, discused inSection 15.5.17.

In addi-tion, the spray system is assumed to be in operation and actsto remove the i-d-inecs fro-m t-he coentainm~ent atmosphere at the same rates assumed forthe -LBL-OCA.

The assumptions used for breathing rates, population

density, and othercommonA fac-tors were also desc-ribead-in earlier sections.

Bo0th the prm; nsecon~dar; coolant activities prior to the accident are given in Sectio"n 154.5.3 The gapactivities are listed in Table 11 .1 7.All of the data and assumptions listed abov.e were used with the EMERALD comFputer progaram to calculate the actiVity

releaser, and potential do-ses following the accident.

The ca!GUlated activity releases are listed in Table 15.5 51, and the potential doses aregive inTabe 1.5 2. hyrid doses that woul1d result from secondar; steam releasescan be determ-ined from, Figues 15.5 2 and 15.5 3 for the DBA conditions and Figures15.5 4 and 15.5 5 for the expec-ted con~ditions.

If atmospheric steamn releases, Aoccnur feollowing this accident, there will be soe exposures via this pathway.

The detailed assumptions used in estimatiRg moede of exposure are described in Secation 15.5 21. The resuts ar gieparam1etrically in Figures 15.5 1 and 15.5 15. it should be noted that these figures aFebased on the assumptions of a full plant cooeldown with no condenser capacity available, a condition that would not be expected to Mour follwing a rod ejection accidenFrom these analyses, it as n be concluded that offite exposuwes frm this accident willbe well below the guidelie levels specified ir 10 puR 100.e11, and that the occurrence of s hch acidents would not result in undue risk to the pubcii. A detailed evaluation ofpotential exposures to contrio room pefrsonelis mlade in Sectionta 15.517 for cnditions following a LOCA.15.5.23.2.2 Activity Release Transport ModelThe CREA dose consequence analysis evaluates the following two scenarios.

Scenario 1: The failed fuel resulting from a postulated CREA is released into theRCS, which is released in its entirety into the containment via the faulted control roddrive mechanism

housing, is mixed in the free volume of the containment, and thenreleased to the environment at the containment technical specification leak rate for thefirst 24 hrs and at half that value for the remaining 29 days.Scenario 2: The failed fuel resulting from a postulated CREA is released into theRCS which is then transmitted to the secondary side via steam generator tubeleakage.

The condenser is assumed to be unavailable due to a loss of offsite power.Environmental releases occur from the steam generators via the MSSVs and 10%ADVs.15.5-138Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe chemical composition of the iodine in the gap is assumed to be 95% particulate (Csl), 4.85% elemental and 0.15% organic.

However, because the sump pH is notcontrolled following a CREA, it is conservatively assumed that the iodine released viathe containment leakage pathway has the same composition as the iodine releasedvia the secondary system release pathway; i.e.; it is assumed that for both scenarios, 97% of all halogens available for release to the environment are elemental, while theremaining 3% is organic.Scenario 1: Transport From Containment The failed fuel activity released due to a CREA into the RCS is assumed to beinstantaneously released into the containment where it mixes homogeneously in thecontainment free volume. The containment is assumed to leak at the technical specification leak rate of 0.10% per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at half that value forthe remaining 29 days after the event. Except for decay, no credit is taken fordepleting the halogen or noble gas concentrations airborne in the containment.

PerRegulatory Guide 1.183, the chemical composition of the iodine in the gap fuel is 95%particulate (Csl), 4.85% elemental and 0.15% organic.

However, since no credit istaken for the actuation of sprays or pH control, the iodine released via containment leakage pathway is assumed to have the same composition as iodine activity releasedto the environment from the secondary coolant; i.e.; 97% elemental and 3% organic.Environmental releases due to containment leakage can occur unfiltered as a diffusesource from the containment wall, and as a point source via the containment penetration areas or the Plant Vent. The dose consequences are estimated based onthe worst case atmospheric dispersion

factors, i.e., an assumed environmental release via the containment penetration areas.Scenario 2: Transport from Secondary SystemThe failed fuel activity released due to a CREA into the RCS is assumed to beinstantaneously and homogeneously mixed in the reactor coolant system andtransmitted to the secondary side via primary to secondary SG tube leakage.

Theactivity associated with the release of the initial inventory in secondary steam/liquid, and primary to secondary leakage of normal operation RCS, (both at Technical Specification levels) via the MSSVs/10%

ADVs are insignificant compared to the failedfuel release, and are therefore not included in this assessment.

DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tubeleakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. Toaccommodate any potential accident induced leakage, the CREA dose consequence analysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events),

hasbeen evaluated for potential impact on dose consequences as part of a WOGProgram and demonstrated to be insignificant; therefore, the gap iodines have apartition coefficient of 100 in the SG. The gap noble gases are released freely to the15.5-139Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEenvironment without retention in the SG.The condenser is assumed unavailable due to the loss of offsite power. Consequently, the radioactivity release resulting from a CREA is discharged to the environment fromsteam generators via the MSSVs and the 10% ADVs. Per Regulatory Guide 1.183,97% of all halogens available for release to the environment via the Secondary Systemare elemental, while the remaining 3% are organic.

The SG releases continue untilshutdown cooling is initiated via operation of the RHR system (10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> after theaccident) and environmental releases are terminated.

15.5.23.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at anypoint on the boundary at the EAB, for any 2-hr period following the onset of theaccident be reported as the EAB dose. For Scenario 1 (release via Containment leakage),

the worst case 2-hour period occurs during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). For Scenario 2(release via secondary side), the worst two hour period can occur either during the 0-2hr period when the noble gas release rate is the highest, or during the t=8.73 hr to10.73 hr period when the iodine and particulate level in the SG liquid peaks (SGreleases are terminated at T=10.73 hrs). Regardless of the starting point of the worst2 hr window, the 0-2 hr EAB X/Q is utilized.

The bounding EAB and LPZ dose following a CREA at either unit for both scenarios are presented in Table 15.5-52.15.5.23.2.4 Control Room Dose Assessment The parameter values utilized for the control room in the accident dose transport modelare discussed in Section 15.5.9. Provided below are the critical CREA-specific assumptions associated with control room response and activity transport.

Timing for Initiation of CRVS Mode 4:The time to generate a signal to switch CRVS operation from Mode 1 to Mode 4 isbased on the containment pressure response following a 2 inch small-break LOCA(SBLOCA),

and the fact that at DCPP, a Containment High Pressure signal will initiatea SIS which will automatically initiate CRVS Mode 4 pressurization.

The containment pressure response analysis for a 2 inch SBLOCA shows that the 3 psig setpoint forContainment High Pressure is reached in 150 seconds after the SBLOCA. Asindicated

earlier, releases to the containment following a CREA are through a faultedcontrol rod drive mechanism housing.

The control rod shaft diameter is 1.840 inchesand the RCCA housing penetration opening is 4 inches in diameter.

Based on theabove and for the purposes of conservatism, the time to generate the Containment High Pressure SIS following a CREA is assumed to be double the value applicable tothe 2 inch SBLOCA, or 300 seconds.15.5-140Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEBased on the above, following a CREA,a. An SIS will be generated at t = 300 sec following a CREA.b. The CRVS normal intake dampers of the accident unit start to close after a 28.2second delay due to delays associated with diesel generator loading onto the 4kvbuses. The control room dampers are fully closed 10 secs later, or at t=338.2secs (i.e., 300 + 28.2 + 10). The 2 second SIS processing time occurs in parallelwith diesel generator sequencing and is therefore not included as part of thedelay.c. In accordance with DCPP licensing basis, the CRVS normal operation dampersof the non-accident unit are not affected by the LOOP and are isolated at t=312secs (i.e., 300 + 2 secs signal processing time + 10 sec damper closure time).Control Room Atmospheric Dispersion Factors:As noted in Section 2.3.5.2.2, because of the proximity of the MSSV/10%

ADVs to thecontrol room normal intake of the affected unit and because the releases from theMSSVs/10%

ADVs have a vertically upward discharge, it is expected that theconcentrations near the normal operation control room intake of the faulted unit(closest to the release point) will be insignificant.

Therefore, prior to switchover toCRVS Mode 4 pressurization, only the unaffected unit's control room normal intake isassumed to be contaminated by a release from the MSSVs/10%

ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity releasepoints / control room receptors applicable to a CREA at either unit are provided inTable 15.5-52B.

The X/Q values presented in Table 15.5-52B take into consideration the various release points-receptors applicable to the CREA to identify the boundingx/Q values applicable to a CREA at either unit, and reflect the allowable adjustments

/reductions in the values as discussed in Chapter 2.3.5.2.2 and summarized in thenotes of Tables 2.3-147 and 2.3-148.The bounding control room dose following a CREA at either unit is presented in Table15.5-52.15.5.23.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:(1) The radiation dose to an individual located at any point on the boundary ofthe exclusion area for any 2-hour period following the onset of thepostulated fission product release is within 0.063 Sv (6.3 rem) TEDE asshown in Table 15.5-52.(2) The radiation dose to an individual located at any point on the outerboundary of the low population zone, who is exposed to the radioactive cloudresulting from the postulated fission product release (during the entire period15.5-141Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEof its passage),

is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-52.The radiation dose to an individual in the control room for the duration of theaccident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-52comparing the actiVity relea.es following a rod ejection

accident, given inTable 15.5 51, with the activity releases calculated for a LOCA, given inTable6 15.5 13 and 15.5 14, it can be conclIuded that an" controlI room exposurFes following a rod ejection accident will be wel! below the GIDC 19, 1971, criterion level.Additionally, the analysis demonstrates that the acceptance criteria are Met asfollows-(1) The radiation dose to the whole body and to the thyroid o-fan in~divid-ual located atany point on the boun~dary of the eXclusion area for the two hours immediately folloWing the onset of the postulated fissionR product release are well below the doslimits of 10 CFR 100.11 as shown in Table 15.5 52.(2) The radiation dose to the whole body and to the thyroid of an individual locatedaany point on the outer boundary of the low population zone wh is exposed to the-radioac~tive cloud resulting from the postulated fissionR product release (during theentire period of its passage),

are well below the dose limits of 10 CFR 100.11 asshown in Table 15.5 522.(3) Since the activity releases from the rod ejection accident giwven in Table-15.5 51 are less than those from a LBLOCA (see Table 154.5 And 1.5.5 14), any-control room dose which mnight occur would be well below the established criteri oGDC 19, 1971, and discussed in Section 15.5.17.15.5.24 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A WASTE GASDECAY TANK15.5.24.1 Acceptance CriteriaThe radiological consequences of a rupture of a waste gas decay tank shall not exceedthe dose limits of 10 CFR 100.11 as outlined below:(1) An individual located at any point on the boundary of the exclusion area for thetwo hours immediately following the onset of the postulated fission productrelease shall not receive a total radiation dose to the whole body in excess of25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodineexposure.

15.5-142Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.27 REFERENCES

1. Nuclear Safety Criteria for the Design of Stationary Pressurized Water ReactorPlant, N18.2, American Nuclear Society, 1972.2. Regulatory Guide 1.70, Standard Format and Content of Safety AnalysisReports for Nuclear Power Plants, US Atomic Energy Commission (AEC),Rev. 1, October 1972.3. Regulatory Guide 4.2, Preparation of Environmental Reports for Nuclear PowerPlants, Directorate of Regulatory Standards, AEC, March 1973.4. W. K. Burnot, et al, EMERALD (REVISION I) -A Program for the Calculation ofActivity Releases and Potential Doses, Pacific Gas and Electric Company,March 1974.5. S. G. Gillespie and W. K. Brunot, EMERALD NORMAL -A Program for theCalculations of Activity Releases and Doses from Normal Operation of aPressurized Water Plant, Program Description and User's Manual, Pacific Gasand Electric

Company, March 1973.6. Regulatory Guide Number 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors, AEC, Rev. 1, June 1973.7. D. H. Slade, ed., Meteorology and Atomic Energy 1968, AEC Report NumberTID-24190, July 1968.8. International Commission on Radiological Protection (ICRP) Publication 2,Report of Committee II, Permissible Dose for Internal Radiation, 1959.9. DeletedR.

L. Engel, et a!, I8SO4SL A Compute r-,de f, .Gene...

I Io .I---io,,tpShielding Analygi,,

BNVVI, 236, UC 31, Physics, Pacific North.e.t Laboratory,

Richland,

,, June 1N66.10. R. K. Hilliard, et al, "Removal of Iodine and Particles by Sprays in theContainment Systems Experiment,"

Nuclear Technology, April 1971.11. G. L. Simmons, et al, ISOSHLD-I1:

Code Revision to Include Calculation ofDose Rate from Shielded Bremsstrahlung

Sources, BNWL-236-SUP1, UC-34,Physics, Pacific Northwest Laboratory,

Richland, WA, March 1967.12. L. F. Parsly, Calculation of Iodine -Water Partition Coefficients, ORNL-TM-2412, Part IV, January 1970.15.5-150Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE13. Westinghouse, Radiological Consequences of a Fuel Handling

Accident, December 1971.14. Deleted.

F. J. Bruts.hy, etal, B-ehA,-iOr Of Ioine in Reactor Dun, PlantShutdown and- Sftartuia, General Electric Co. Atomic Power Equipment-Depatrtment Report, NEFDO 10585, August 197-2.15. Deleted in Revision 16.16. Proposed Addendum to ANS Standard N18.2, Single Failure Criteria for FluidSystems, American Nuclear Society, May 1974.17. K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control RoomVentilation System Design for Meeting General Design Criteria 19," 13th AECAir Cleaning Conference, August 1974.18. M. L. Mooney and H. E. Cramer, Meteorological Study of the Diablo CanyonNuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1970 (see also Appendix 2.3A in Reference 27 of Section 2.3 in thisFSAR Update).19. M. L. Mooney, First Supplement, Meteorological Study of the Diablo CanyonNuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1971 (see also Appendix 2.3C in Reference 27 of Section 2.3 in thisFSAR Update).20. M. L. Mooney, Second Supplement, Meteorological Study of the Diablo CanyonNuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1972 (see also Appendix 2.3D in Reference 27 of Section 2.3 in thisFSAR Update).21. International Commission on Radiological Protection Publication 30, Limits forIntakes of Radionuclides by Workers, 1979.22. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2,Appendix A to License Nos. DPR-80 and DPR 82, as amended.23. Safety Guide 25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and StorageFacility for Boiling and Pressurized Water Reactors, USNRC, March 1972.24. Safety Guide 24, Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage TankFailure, USNRC, March 1972.15.5-151Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE25. Deleted.Sta"t, G. &, J. H. Cate, C. R. Dickson, N. R. Ricks, G. H. Ackerman, and- et. F.. Saged, "Ranch. Seco Bui.ding Wake Effect. On Atmospheric Difusioen, NA~A T-echnical Me-morandum.

ERL .ARL 60,17.26. DeletedWalker, D. H., R. N. Na.sa.o, M. A. Capo, "Control Room V ,ntilation Intake Selecti-n for the Floating PWer Plant," 14 th ERDA-Air

-lA2 nin-Con.fere.nc, 1976.27. DeletedHatcher, R. N., R. N. Mer.ney, J. A. Peterka, K. Kothari, "Disper.ion inthe Wake of a Model IRndustial Complex,"

NUREG 0373, 17828. DeletedMerO.ey, R. N., and B. T. Tunnel Stud' on Gaseous due to Va iourn, Star Hei-nL Unho ,nrd Rinmani;nn Pate,a Abv knn', Isol,4ated StFrUGWn~

FDDL Report CER 7-1 72 RNMI TY16, Clr,,ado State University, 1971.29. DeletedR.

P. Ak- , I. " nnO; R o.n ;n th A \/,'i,; , i ldin; " D e...i. S Onf.SeGOnd- Conference on Industrial M~eteorology, New Orlean~s, LA, March 21 28,1980, pp. 02 107, American Meteorological

Society, Boston, Mass. Also in30.31.Pw ion, D. Randerso, ed., USnE.Deleted[).

j. Wilson, Cont-amination of Air Intakes from Roof E~xhausgt Vents,ASHRAE Trans-. 82, Part 1, pp. 1021 1038, 197-6.Deleted B ,um,+.t,.

M. 5tr R. N. Me ,,.y An A!ftem Winn Tunne +r-sfane.

I tr.,.Ja.

tkI *a, I4,aRG, Spteei l1180fromedR J Tune Exn........

+, NUREGIC,...;R 1171 US....RC, September,,,,

108032. R. Bhatia, J. Dodds, and J. Schulz, Building Wake ,/Qs for Post-LOCA ControlRoom Habitability, Bechtel Power Corporation, San Francisco, CA.33. Report on the Methodology for the Resolution on the Steam Generator TubeUncovery Issue, WCAP-13247, March 1992.34. Deleted in Revision 18.35.

Guide 1.4, Assumptions Used for Evaluating tho Potontial 0-RI ; 1,,.ir.sI

'al G nseuAe ao a 0Crf a, 1=9ape f Geelin t n* A 0 t- fqF a Drasufg *l.Aater Reactorfs, AEG, Revision 2, June 107-4.,36. T. R. England and R. E. Schenter, ENDF-223, ENDF/B-IV Fission ProductFiles: Summary of Major Nuclide Data, October 1975.37. Standard Review Plan, Section 15.6.3, Radiological Consequences of SteamGenerator Tube Failure (PWR), NUREG-0800, USNRC, July 1981.15.5-152Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE38. Deleted in Revision 18.39. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment SprayFlow Rate Reduction, February 7, 1991.40. Westinghouse Letter PGE-93-652 dated October 5, 1993, transmitting NSAL-93-016, Revision 1.41. K. F. Eckerman et. al., Limiting Values of Radionuclide Intake and AirConcentration and Dose Conversion Factors for Inhalation, Submersion, andIngestion, Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.42. K. F. Eckerman and J. C. Ryman, External Exposure to Radionuclides in Air,Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.43. Deleted in Revision 12.44. Regulato,,'

Guide 1.195, "Methd. and A^,umtin-for, =E,,autgD-9nGlnn;,.,

GC on,,monRGe of ige n Basoi;s ArG~riAnd e nts , I ; Afnr atI. Lie lNrPo9wer React9rz,"

M52003.Deleted 44-45. Deletedlntrntat-Ina!

Commission on PFrtection (IRP) Publicat;on

• 0n, ;mr+ c,-, ,,s .....I;,des b,,W,,k,

-,, 07/!9 .745-A6. Diablo Canyon Units 1 and 2 Replacement Steam Generator Program -NSSSLicensing Report, WCAP-16638 (Proprietary),

September 2007.46A7. LOCADOSE-NE319, A Computer Code System for Multi-Region Radioactive Transport and Dose Calculation, Release 6, Bechtel Corporation.

4-7-48. PG&E Calculation N-166, Small Break LOCA Doses, Revision 0,October 31, 1994.48-.49. Diablo Canyon Units 1 and 2 Tavq and Tfeed Ranges Program NSSSEngineering Report, WCAP-16985 (Proprietary),

April 2009.49.50. DeletedA,.

G. ,,

A 0A .Revised and Uodated reor. ,f ,, ,-, a -Ridqgezg I -tp Gnrtion and Deplto CoeQRNL=

5621, Oak Ridge-National Laborator,',

July 1980.50751. NRC Letter, License Amendment No. 155/155, "Diablo Canyon Nuclear PowerPlant, Unit Nos. 1 and 2 -Issuance of Amendment RE: Revision of Technical 15.5-153Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATESpecifications Section 3.9.4, Containment Penetrations (TAC Nos. MB3595and MB3596),

Accession No. ML021010606, October 21, 2002.52. "RADTRAD:

A Model for Radionuc.ide Tr.anspo.t, Removal and Dose",NUREGIR 6604, SANDg8 0272, April, 1998.Deleted

53. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites",1962.54. Not Used.55. Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms forEvaluating Design Basis Accidents at Nuclear Power Reactors",

July 2000.56. NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseousand Liquid Effluents from Pressurized Water Reactors",

Revision 1.57. NUREG 0737, Clarification of TMI Action Plan Requirements, November 1980.58. NUREG 0737, Supplement 1, Clarification of TMI Action Plan Requirements, January 1983.59. NUREG-0800, Standard Review Plan 15.0.1, "Radiological Consequence Analyses using Alternative Source Terms," Revision 0.60. Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential AccidentConsequence Assessments at Nuclear Power Plants",

Revision 1.61. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations inBuilding Wakes". Prepared by Pacific Northwest Laboratory for the U.S.Nuclear Regulatory Commission, PNL-1 0521, NUREG/CR-6331, Revision 1,May 1997.62. NUREG/CR 5009, Assessment of the Use of Extended Burn Fuel in LWRs,January 1988.63. RADTRAD 3.03 (GUI Mode Version),

A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, NUREG/CR-6604, Users' Guide-Supplement 2, October 2002.64. SCALE 4.3, "Modular Code System for Performing Standardized ComputerAnalyses for Licensing Evaluation for Workstations And Personal Computers,"

Control Module SAS2.65. ACTIVITY2, "Fission Products in a Nuclear Reactor"

-CB&I Proprietary Computer Code NU-014, V01, L03.15.5-154Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE66. IONEXCHANGER,

-CB&I Proprietary Computer Code NU-009, Ver. 01, Lev.03.67. SWNAUA, "Aerosol Behavior in Condensing Atmosphere",

CB&I Proprietary Computer Code NU-185, V02, LO.68. RADTRAD 3.03 "A Simplified Model for RADionuclide Transport and RemovalAnd Dose Estimates.

69. PERC2, "Passive Evolutionary Regulatory Consequence Code" -CB&IProprietary Computer Code, NU-226, VOO, L02.70. SW-QADCGGP, "A Combinatorial Geometry Version of QAD-5A" -CB&IProprietary Computer Code, NU-222, VOO, L02.53.71. GOTHIC, "Generation of Thermal-Hydraulic Information for Containments".

72. Not Used73. EN-1 13, "Atmospheric Dispersion Factors"

-CB&I Computer Code EN-1 13,V06, L08.74. ARCON96, "Atmospheric Relative Concentrations in Building Wakes.75. NUREG-0800, Standard Review Plan (SRP) Sections 15.2.1-15.2.5, "Loss ofExternal Load; Turbine Trip; Loss of Condenser Vacuum; Closure of MainSteam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)",

Revision 176. NUREG-0800, SRP Section 15.2.6, "Loss of Non-Emergency AC Power to theStation Auxiliaries",

Revision 1.77. USNRC Standard Review Plan (SRP) for the Review of Safety AnalysisReports for Nuclear Power Plants, NUREG 0800, Section 15.6.5, Appendix B,Revision 1, "Radiological Consequences of a Design Basis LOCA: Leakagefrom engineered Safety Feature Components outside Containment".

78. NRC Generic Letter No. 99-02, Laboratory Testing of Nuclear Grade Activated

Charcoal, June 3, 1999.79. Not Used80. NUREG 0800, 1988, Standard Review Plan, "Containment Spray as a FissionProduct Cleanup System",

Section 6.5.2, Revision 4.15.5-155Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE81. NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays",

June 1993.82. NUREG/CR-5732, "Iodine Chemical Forms in LWR Severe Accidents

-FinalReport,"

April 1992.83. ANSI/ANS 6.1.1-1977, "Neutron and Gamma-ray Flux-to-Dose Rate Factors"84. ANSI/ANS 6.1.1-1991, "Neutron and Gamma-ray Fluence-to-dose Factors".

85. NRC Information Notice 91-56, September 19, 1991, "Potential Radioactive Leakage to Tank Vented to Atmosphere".

86. NUREG 0800, Standard Review Plan 15.2.8, Revision 2, "Feedwater SystemPipe Break Inside and Outside Containment (PWR)"5487. NRC SER Related to Amendment No. 8 and 6 to Facility Operating License No.DPR-80 and. DPR-82, PG&E, Diablo Canyon Power Station, Units 1 and 2,dated May 30,1986.15.5-156Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATETABLE !5.5 !RE-ACTOR COOLA.N NT FWSSION l A'AND CORROSION PRODUCT ACTIVIT-ES DUIRING STEADY STATE OPERAT!ON AND PLANT SHUTDOWN

OPERPTIO, Boafara QhlffdAw,4 ShW tdovw i 4 tht Boforo Shutdown, ShutdoW n A ctiVit,44.9 2-45 4--Xe43W 447.40 &.e2"0"aG64-34 4-2-9 4-4 G-.26_ _ __7 14-& 2ý44 4.34 949WGe444 O049"4.0 0-.00034 0492-9Sf49 Q0.033 0,40 0,0029, 0,32Sf40 0494-3 30-004-3GO-58 90 0-026 14,4(a) Activity reduce-Pd kfro Ateady state level by approximately 1 day of system dcgastficatvOn prior tO plant GhUtdGYW.

Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 2 (HISTtRIGAL)

RESU"LT-S OF STU'DY OF- EFFECTS OF PLUTONIUM ON ACCIDENT DOSES30dyThy4eidGhal~ge30dyWhOWeBdGhafn*ein

-ThdChan~ge 42 houFMolae-Bedy Reilase fromqr gas denaytaRkLoss of Feactor-prImaryGoolant largle broa!kgeneator-tube-Fuptare aGGidentSteam. lWne rupture0-400-40-3-2-2*447-2-2Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 165 54E-XPE-CTEr-D-PS CIETAMSHRCDLTO FACTORS (SEC!M404 §,49x41-0 4 a.49x4l-0 4-974Q9 -84D4 X4*4 24OX494 &g&5i-49-24 2,4&404 40W40 4-75*404 0ý A4E46*44 4-54-G4Q724 OR 49*- 3,04x4G4 14_754-g-

&0x4w 2490404 4454*444 6-Q04-9496:720 41-4Q &20ý4 3- 4040 44-44 §,2g*4-9 3.4G-4Q9 4-24X04~(a) Monimum egGtUzien a~e boundar; FadiUc ic 0.5 milea (appmoxmately SOO mn). Radiuc f low populatien zen is 6.2 Miles (appcGAimately 10,000 mn)Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-5ATMOSPHERIC DILUITION FACTORSQRShe-9Midpei~tQireGti9G4&

Se.+r t Ar~dFr9nm fla.mn.,RW*R QitRe .nuier,SSEIsWWSWwVwNN-WNNW-r--14,64,440-545,394,040-.64048072-604-80.404-760076"-25-072204604-6044440804-g-35-072-30-24-0440708070804-5047-0,770,29074-04-60708046Q-700420400-.2-04-504-30-706G-.-_w-Q-.-..---...-5 04-0-440.48At.o,.pheN.

Dilution FactorsIQx 4 !O-seG--m 4Downwind

DistancG, meters2000 400 7000 14000 2-g000Di-eGtieR 800E-G-.47-----4GAG8 07050 0.0350.0118Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-6ASS UMED ONSIIT- ATMOSPHER-C DILUTION F-ACTDORS (SEC!W3}FOR THF CONTDO' RO)OMBase X14-Modifying F~alPeF~ed, is.A FEr The ,r--stien uase--8-48-24-96424-084*4-9 4,.084x40 44-7837466Ag4-792.84.76.-27-2.-27-24444-75757575765765765765~7,9540434 -42,2749B. For The Infiltration Casego.8-2-2-4-24-Q-9642344W1-037-14*1-04 34.U41-04 47833766g,484.-9.674.-2.-2.-24-4-4-4-7-5.-5.-5.-5.-65.-657-65.-654-k -04&44(a) The k4Q calculated above do not account for credit forF dual pressurization inlet and occu1pancY faGtGFS.Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-8~ltDl II ATICthI MICTDIII ITCltIKlSeGtGMidpei~tDi~eGUG9Rs SS-W-WWNWN-WNNVTePtak;Radial C pr.rOr A;Andpent

)W~~ fln Qn.;nrlflaR-fln nIeS T-etalSeeto4704484047447943004-,72-7-4796620,33445,66626700040,33420,0339-,70G02L, 027000000OG5-00=676041790022,93321,33421,333574334-2344-9M7-04-.764706649,7344 8,66622,2674667-00&44337-3446,06645,33449,4336474664--JWG14006344753367-000&754W4413842,83235790423,06638-17387,6097,0346a9 2 3367904 111,46090,600 7-8,800 56,600 24,730 368,944Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-9SUMMARY OF OFFSITE AND CONTROL ROOM DOSES FROMLOSS OF ELECTRICAL LOADThyroid Doses, remDose _(TEDE, rem)Smte Bounda..

2 Hours DalysRequlator y Limit(TEDE, rem)10 CFR Pat 100 3W WoMaximum 2-hour Exclusion Area 0 0.0066Boundary Dose1Design basis case-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5Spike30-day Integrated LowPopulation Zone Dose-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5Spike30-day Integrated Control Room 5Occupancy Dose-Pre-incident iodine Spike <0.1-Accident-Initiated Iodine <0.1SpikeExpeGted-oase 5-2-x40'Ahole Body Doses, remSt p "Il -A -Unsxr-10 CFR Pat 100 25 25Design basic case 2. 2.444-GFixpeeted Gase 7-~-4O44

.4Population Doses, mnan remRevision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEDesign basir, ca.eExpected-ease-045 t-Note:1. The maximum 2-hour EAB dose occurs during the following time period:-Pre-incident iodine Spike-Accident-Initiated Iodine Spike0 -2 hours8.73- 10.73 hour8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />sRevision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-9ALOSS OF ELECTRICAL LOADANALYSIS ASSUMPTIONS

& KEY PARAMETER VALUESParameter ValuePower Level 3580 MWtReactor Coolant Mass 446,486 IbmPrimary to Secondary SG tube leakage 0.75 gpm (total for all 4 SGs); leakagedensity 62.4 Ibm/ft3Failed/Melted Fuel Percentage 0%RCS Technical Specification Iodine Levels Table 15.5-78(1 pCi/gm DE 1-131)RCS Technical Specification Noble Gas Levels Table 15.5-78(270 pCi/gm DE Xe-1 33)RCS Equilibrium Iodine Appearance Rates Table 15.5-79(1 pCi/gm DE 1-131)Pre-Accident Iodine Spike Concentration Table 15.5-79(60 pCi/gm DE 1-131)Accident-Initiated Iodine Spike Appearance Rate 500 times TS equilibrium appearance rateDuration of Accident-Initiated Iodine Spike 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />sInitial Secondary Coolant Iodine Concentrations 0.1 pCi/gm DE 1-131 (Table 4.2-1)Initial and Minimum SG Liquid Mass 92,301 Ibm/SGTime period of tubes uncovered insignificant Steam Releases 0-2 hrs: 651,000 Ibm2-8 hrs: 1,023,000 Ibm8-10.73 hrs: same release rate as that for 2-8 hrsIodine Partition Coefficient in SGs 100Iodine Species Released to Environment 97% elemental; 3% organicFraction of Noble Gas Released 1.0 (Released without holdup)Termination of releases from SGs 10.73 hour8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />sEnvironmental Release Point MSSVs/I 0% ADVsCR emergency Ventilation

Initiation Control Room is assumed to remain onSignal/Timing normal ventilation for duration of theaccident.

Control Room Atmospheric Dispersion Factors Table 15.5-9BRevision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-9BLOSS OF ELECTRICAL LOADControl Room Limiting Atmospheric Dispersion Factors (sec/m3)Release point and receptor 0-2hr 2-8 hr 8-10.73 hrMSSVs/10%

ADVs to CR NOP Intake (Note 1) 8.60E-04 5.58E-04 5.58E-04MSSVs/10%

ADVs to CR Inleakage (CR 2.78E-03 1.63E-03 1.63E-03Centerline)

______________

I ___ I ___ I ___Note 1: Due to the proximity of the release from the MSSVs/1 0% ADVs, to the normaloperation CR intake of the affected unit, and due to the high vertical velocity of thesteam discharge from the MSSVs/10%

ADVs, the resultant plume from the MSSVs/10%

ADVs will not contaminate the normal operation CR intake of the affected unit. Thus theX/Q s presented reflect those applicable to the CR intake of the unaffected unit.Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATETADI EC 4C: r 4n01 IRkAfAAOV

%C: n~C:CITIrr nnCEQ 10(lAA SMALL LOS ORF C-OLANT ACCInENTNO FUEL IDAMAGEThYroid Doese, r-emSi;+Dte Bounda 27 Lint ar10 CER Part 1 00De-gAbasiS GaseF~peGted Gase3002,04x_-W3w02,7X4-410Whole Body Doses, rem2-tR OR-mmat 2nl.n H iIew10 CFR Part 100Design caseExpeoted Gase.2644-*4-0254-.-x-4-0 4Population Doses, man remDesigR basis case0.01-3Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 11SUMMARY OF OFFSITE DOSES FROMAN .MDEFP.EIE.Cv ACCIDEhTThYroid Doses, remSite O39nda 2n~n H naew[=p7= 310 CFR Pa;t 100Design basic caseExpeGted Ga~e0.02444-430w0.00661.2x4QWhole Body Dozes, remaSite Ben, raw 2 Howsrc10 FR Part 4100Design basis easeExpeGted GaseD)es6ig9A hn basirs- eA-FExpe~ted Ga~e-2-52-524.4-404Dman rem04-54.44-940Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 12SUMMARY OF OFFSITE DOSES FROM ASINGLE ROD CLUSTER CONTROL ASSEMBLY W!THDRA^WAL T-hyrold Deses, remSitp RAnundirv

2 WAi10) CFR Part 100[Design baSis casoE*xpeGted rase3000423000.0433-.4-94%Aflk-- 0-th- M --ar --nZ3118 159UREWY Z HOUFS10) CFR Part 100Decign basis caseEDpeGned GaseDesign ba6*6 easeFEXpeGted Gase~2564-x-4046,54-W4256-74-4046,- 4 Doses, mn Frem4.3-404Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE_OW 0-4V _4 8 214 hr 241@v 06 hr _y1424 0.4195E 04 0-0 0-0 00-1420 9.4i S 0,0 -142 a taw- 02 0-. -000 -1 435 0462Q&" -000 00 -~14320RG1 434GRG144PARI 132PARI 433PAR1424ARKF 88Xe 133Xe 43~mXe 13Xe435MXe-42R0.1424E-01 9.1780E 020.1314E 02G44-3E-02 G-00-00-00.36GE 01().4285E01i

0. 503 E 010 5Q4r27 020.9285re00 0,6.62E-01 0.5 1289E 006,E-040.1556Er.02 0.2211 W-010.8366E-02 04-0.6227E00 n0.4 2650 010.11 00E,02041035E,02

0. 2210 E 030#T-4.E-04 0A670 E 050-4734E-02 0.00-0G0 4257E,029.4689E-Qi 0A.2.12-E--.1 0.1 290E 010.-1.,,0E--3 0.6-162E-010 1826E-02-.7140.E4.0 0.1 490F 100.l384,0o 0.4313 lE 060,3 542F 020. 9077-E 120-0Q.4-757E 01a g571EF.02 0.:7297-E 01O.1922E05 7.10Q7E01 a 73so= pa0. 8236E 010. 1 WOE 20.1979931 0.6436E-47 047-30E--08 0.87508E 050.1730E 220-00-083-8047-4-G44418&00-01 446gr 010.1721S 01Revision 11 November 1996 DCPP UNITS I & 2 FSAR UPDATETARI E,45 5 44Af'Tl~llTV DCICACCO C:QnhAA V'A flCClflkIDAI E -AI I II A I II6.0 .1-6 6 10 -I I. W ..---..-..---N 40 .N ý, 2-Hr-04 6 HrF1 43314-34$434GRG1140RGl424PAR1 P33PAR1134PARKPWMXe 433Xe 435MXe 135MXe 1380. 2703r E 20. 3085e 02O.6207E 02a.7063rE0O

0. 5a74 12p 020,72406E02 0.163 gE 030.08477602-

-0.1141 4E 03-O.1041E03 0.14231 R 030. 9280E 030 2922r, 040. 3847E 010.70943rE(3 Q.7402E 040.26064E 040198-030.217-4303 0.32866 020-23689 0.747193E 0438646 049 0426.8-05 0.far a~E00.41376E01 0.6 552E al10-00-00-00-00.65 6 lE 03Q-2326.E02

0. I08 03.or 10.2015Er.00 0.690rE03 0.257-2E,04 0.7-2736 020- 2388&"00.1281 OE0. 202gre 050.46,0E, 060 1464E-060-00-09.9240S ,-1a 5263r=-03 9.28-11-906 0-344SE 04338E 040.657.9E 020.2220E 020-3016E-02 O.12-52E027

0. 3227E 010.,8557Ec 109 5383r=-02

0. 26656 319490709 90.082-7r,04 0.4 413E020.1530E 10042333r 069.25566E04 041014E 020-0GRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATET- Ir. 4 s & 45r%THYROID DOSE 2 HOUR CONTAINMENT LEFA.KA.GE EXPCTE CARE (REM)Distanoo Fromq Release Point, metercNUG4de 80 424DG 2000 4000 7000 40000 2009001131 0.7456F 03 0.41792E 03 0.2636FE03 0.1970 0.506GE04 034E-04 0.1247-E 4112 .4895 04W29 0 Q754E 0 E06 (3.2074Eo06 a4823F.0 06 a73r-07143 0.1627E 03 0 994146EO0 0.6389004 9.22466 04 GAGNE -04 04-888,08-0 0.25461 24 a226ar a a 01 ~ r as~880 0 6017E06 as 22r818-a01ig 04480 0 0482250 07. 93-94 70-9743E-"

04646 0544682-01 4024r.as 0-A444-E-06 1134O4RG 0:2687E0 32 0416636E03 0.446ri 9 0.3805r!5 04.a1756Er.

0 a1076E01D-0.432BE 011382ORG Q 1169rE05 047511E 606 0 4132rE06 0.1710Ea06 0.7032rE07 0,4662r50-9490569 07k483QRG 0.549E G4 0.331gm0E0 0.182GE1 04.97573E 05 a,4.Q8- 05 0.2112rE 05 0,94 5E0134G4RG 0.103iE096

0. 2501ErM06 0.11425E 06 0.9967 a.16.7 a477Er.071 0A4 4EnG148QRG 0-7820E05 0-42fiG 0-2764E)5-0.1150E05 a 0.5207r,06-0.212&r,0A-0.14801434PAR 0000000000000 1~322A R 0.0004- 0 0400-1 133PAR 0-0 0- ---0 0- 0-01 424PAR 0-000.00000000-

!35R1AR 0-0 00- -000000TOTAL 0 :40R-02 20380r.0 044~46r 0 1R3r 803 0.-047E0 0.516E,04

~0 09G-4Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETAB LE 1545 17THYROID DOSE30DY ONANMNTLAKG EX-PECTED CASE86 (REM)Digto;nn From Release Pzint, moetemNUG4de 800 4200 2000 40W0 7009 4000G=00"131T 0.74668m0 0 CEO=02E0336 9 0.14007E03 0.56G 0 0,3101E0 9 0.4247E 941 432 9.4383E-05 a a~. i 04496, 0 0 6445FE06 0 24r,9 0.~48296 9-3E0"-33 0.4527SE03 0 0484F, 4 0.538E 94 9.2246E-04 94366 04 0.6580 9 0254E0143 -47E44- 0426894W3&06 G-1646&0&

9480E-00 0443*9 044Zb424ORr-0.14252E0G2 0 75426E03 0.1587E a4 -R03 0.22r-9 0.41452E-04 0.47762P04 4 32ORG 0,2601138E 4AW05 9.-82.63 27-E06,O 9.4 87F 6 09460 74r1 Q30RG 9.919 .128GE 03 0.6:797E01

.7W 9 45E0 0.7-7346 05 .74m00 30G Q494,C .3456E 96 0.1736E06 0.7222Em07 0.33E97 0242,0 0.8215608 1-3O4- 026E0- 0-4473E-04 0.705E0 0-3264E 0.1105E05&

0202r.06 0.3670E06 1 !31 PAR 04 0404 0.4 04 0-001 132PAR 0 40 40 401422PAR 94 040 0 4 401434PAR 040404040 04 01 135PAR 0 40 40 40TOTAL ~ ~ ~ ~ ~ 04 0.479 .9E0 O8)6 .3266Em03 0 49,0 0.9196E 04.66E0Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 4 15& 100.HOLE B-ODY OSE 2 HOUR CONTAINOENT LR.KAGE EXPECTED CASE (REM)Mirtanne From Release Poin;t, moetersN~e #0042400 2000 400 90 400 20000-i1i9.04240 0-495m060 0.1a7-5EO 6 0.4471E607 0.206597 0.1265E0Q7 9.089C 0801-422 .2- e9 .442 Q .G@=0 .34Q 0- 1543F4 07 00 0456,G Q304 R09I i22 9.404# goe9 .63 9 40im0 -64626 07 0 243F 07 042F, 0 0.-400orn g114 014F 6 044 6 024~m0 .7110 2ir 07~i 9.-667i G 0.3084 9I-&044EW 04724E-06 0-4694E00 0424-0 0-2884&07 0-4760t 0-7402E-08 I 434C)RC O.1055r-06 0.7-4r 07 ~ G.7-i 97 06o 7 a7163E-oa 0 42r O a4766 OI 422ORZ GAM~E 07- 0.389g0E0 W O.24446E07 0.8921E05 44417rEOR 0.224 G .045E 91 433ORG O.43E9 .90826E7 9.9690 74 O.0S 7 G.59i0 .5876E08 0.2361E089 1 4240RG, O.327-7Er07 0.2106907 aO.1158E.07 a 404r OR .2221E089 0.4363E0go 0.482m9baQQ 0.1330EO6 04OM-087 0-740-7 .40706407-0.98G 9 0.557O0 Q 224 FOI 434-PAR 04 4 4040 41-422PAR 04G 04G 004040 041 133PAR 040--0 04 -0 4 -1 434PAR -40 4000 -I 43SPAR- 0-0 0 0- 040-044KF 3M .496re06

.4246 6 06774 E070.1i300gB 07 795.O 04204E 08Kr-8& 046-91;04 0429;; (A 0.51926 0 0.2WE 9 0 0067r00 R a64 gor, 0 0.457F506 Kr ASM Q 00.0 .74 5 039 s 0.142246,05 GAS 6 0.343& 9 0.1506Em06 KF8- 0.4 7 04 0.3083rm 04 O106E1 AGE 4 0.0006e 05 a2266.054 9~49955 0.9026E 06K 80.424i48E 0475-04 0-3704K04 0-144E-04 G0.m4412E0 Q-4359fi4 0.,175EXe 433 9,4269903 0.07 4 046E 4 0 4962r,04 0.8555gB 5 9.23E05 0.2408rxe-433M 0.268r 06.a1709r.5 of 9936E 0.300GB 06 0.1804F06 OR .11i05E 96 0.4447 07Xe 435 0,47-66104 0 32960 14 0.1694B4EOI G6 0 05 .23 5 48r 5 0Q.0Xe 4350 0070.0 )5605r.0s6

( .3083BpCAaS22R 0.1252BPS9 0.367 7 01158E0-7 Xe-428 0.45 6 0.609E05

.3 E 4 05 0.396 05 9.437i 0,3466 06 n R6. rTTL 0.3653993 0.24SE0 0.1291E 03 067r04 0.2470r.04 0 4540F. 041 .14E0Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATET.ABLE 15r5 Sr 20 (DELtETE!D)

T.A.LE. 155 21VWHOLEr-BODY DOSE 30 DAY CONTAINAAENT L--..KA.GE-EXPECTED CASE (REM)D-t;icbnp From Release Raint, motors__kl 9w0 4200 2000 40W0 7000 40000 20000-I1l49.92 0 a466.a 9.1i0766E06 0A444 97 a2065F. 07 400809 aACOF0O1- .2-66 0.144626 06 a -8080840 9.344E 07- 0.443E 0 094 0- 8 0 20FOQ 4420~0r, 92996 0196 066E0- 083 0-0 941742E07 9.0-0GE 001 44 a4R2F.or, 04i 8F5 6 .65 E 0 9711rm 7- 0.1251E 07 C).7QA;0B ga -084E 081 13 0.1245E06 Q-7Q- --GE-& 024W 0 2291R07 0.1766E07w G4042&811310R0 0.510G 06 .98S6 06iS 0.647-4E07 0ý247-~ 0 0.181E097 0.71487E08 0.2RG 1 467r.6 as 9463E097 9.400060W 0.1607-E07 0.7822E 08 0.4796F508 0.10 27- n0 8k433ORG G.57-36906 0,35V11BPS 0.1865BpS 9.7649007 0 2449E,07-0,2422E 07 G.436m9M4340G a.0299F.07 0.2566Bm07 0 414E1B07 0 5870B08O 0.27G~.08 BPS O 667J 044028&W 02522B.pa 04-162946 0.5590B07 0.2566B07 0.1576ý 0.6301BF-0 k434AR 0 40 40 40-432RAR 0040040044 I-433PAR 04004004000 I 424PAR 0 40 40 40t-435PAR 0404 0 04- 0 0KF 8M 04356rmG6 0223E-09424 06 0.515P1B07 0.2376E 07 046 7 G65 MK 50.448ge0-08 44A=0 0.5620rm 4 .206E 084 0480Q06F as 0 fi0-Fa )245O05 6Kr E4 744R04 047Mr 0417 aB 027=1 20r 0 1744F-05 Q.1071BP W 5 042a060A87807 9F 4 .59E044 9.256994 90609 4 0,408405 0.297-G G 0446E05Kf"0-44A4fiW 4 0.4W7qF-04 0.4603E-04 0.1612B01 9-9894E-G 0409699 6Xe 438 0 14 222. 2 0.6011Bm03 9.440046E9 aI Q~.1an0B794F 04 9.36986044 G.43836 04Xe 432A4 0 :1376E04 9 4224rm 04 a.6170rBPs a043850-0 0 .1072BP s 046329046 42578r- 06Xe 135 0.2ii3E0-08 28r 3 77g- 04729E 4 0.24 604 0764 5 0.34065-0W Xe 135M 08-063r 68,620- .3060 .28E9 0 5947BF07 P.2644FB07 80 444r-Xe 138 0.04GO8-05 P.120 .33626B5PS 9OM0 .44E03 025r59 4E9TOTAL 9 203 E02 Q I SA-02 0 549F0-03 2@. 4~803 ~0.444 03 0.446G4 25-04Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-23SUMMARY OF OFFSITE, CONTROL ROOM & TECHNICAL SUPPORT CENTER DOSESLOSS OF COOLANT ACCIDENT3SUMMA, RY OF EXPOSURE FROM CONTAINMENT Dose(TEDE, rem)Regqulatory Limit(TEDE, rem)Maximum 2-hour Exclusion AreaBoundary Dose130-day Integrated LowPopulation Zone Dose30-day Integrated Control RoomOccupancy Dose230-day Integrated TSCOccupancy Dose25.6252513.7 (0.7)4.1 (1.3)55Thyroid Doses, remEAB 2 Hours,1rQ Par 100Design baSic caseExpee~4ed ase30047-4-Whole Body Doses, remEABR 2 Hours1 0 GFR Part 100Design basiS caseExpeeted Gase2-5375-.64-"2,56.44W-OPUiaTnOn Uoce& Man FemDesign basis caseEX198Gted Gase-932.4Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATENote:1. The maximum 2 hr EAB dose is based on the assumed RHR pump seal failure resulting in a 50 gpm leak of sump water occurring at t=24 hr for 30 mins. This release pathwayis considered a part of DCPP licensing basis with respect to passive system failure.

Ifthis assumed release pathway were not included, the maximum 2 hr dose at the EABwould occur between t=0.5 hrs to t=2.5 hrs (i.e., during the post-LOCA ex-vessel releasephase and would be 3.4 rem.2. The dose presented represents the operator dose due to occupancy.

Value shown inparenthesis represents that portion of the total dose reported that is the contribution ofdirect shine from contained sources/external cloud.dose received by the operator during transit outside the control room is not ameasure of the "habitability" of the control room which is defined by the radiation protection provided to the operator by the control room shielding and ventilation system design. Thus, the estimated dose to the operator during routine post-LOCA access to the control room is addressed separately from the control room occupancy dose and is not included with the control room occupancy dose for the demonstration of control room habitability.

As demonstrated in Section 15.5.17.2.4, the dosecontribution to the operator during routine access to control room for the duration ofthe LOCA is minimal.

(a) The- e values correspond to the origina!

analysis.

SeeTable 15.5 75 for current analysis3. (b) The EAB Whole Body dose of 5.61 rem is 3.69 rem gamma and 1.92 remn beta&3. (G) The L=PZ= Whole Body dose of 0.57 rems is 0.33 rem gamma and 0.24 rem betaRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE-rADI 4* r ')A0" ---rlnoulý-

Itu.u &4 -- ". e- -.--ASSUMPTO.NS USED TO CALCULATE OFFSITE EXPOSURES FROM POST LOCACIRCULTIONH LOOP LEAKAGE IN THE AUXILIA.RY BUILDINGA. EGGS, Contaim~ent F~an Cooler,Containment Spray System Operation

1. E-C-CS tFRan fun~tiqning

2. Contaiwnent fan coolcRz AtIneliOning

3. Containment spray system6..AC*J'ity Dgepocited in Containment RAcirQW'-eation Sump Wat9r1. ledine(Crae inventor' base on both U 2359 PU 239 fi4ionS)Experted2a2ofgapin*.entwy, perTable 11.1 7;i (I127, 120, rel.131, 132,133,fat. Table 1.1100% ef gap3.2g. , 0 v,1-8.5g, 9 Gi44g4824PQ 0GiSxpest!Go% of gapin.ventor,'

per Table11. 1 7;. (1127, 129,rel. ffa~t. At 0.01 5i1 434 433-134, 135, mel. ffaet.Tab!l 11.1 7)100% of gap iedine30.2g,O0 i61.5g,O1.2x 2a4-22410% of gpainventer; per1.25; (1 127,129, rel.-fraet. of 0.30; 11431,1 32, 133, 131,135Fl alOf0. 10)99475% of gap odkino4,445g,O0 i97,8AS4O5GI07 490% ef ga?1.25i 0 127, 129, Fel.f-aet. I 131, 132,123, 134, 135, rel.o:a, cf 0.10)99.75% of gapiodine inyenten; 4,445g,O0 Ga. wlemental 10d'no nveotrY(1) 1-127(2) 112-9(3) 1 131, 132, 133, 134, 135Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATEFADI C 4-I r ')A4L'L.--. 4 -CC.ASSIUMPTIONSi USED TO " AILCULITE OFFSITEEXPOSURES FROM POST LOCIAýW__ _ý I-.- -ft --M_ Ir, +MM 15wHWlPl"_

b. OrglaniG iedine(4) 1 27(2) I123)I131, 132, 133,4134,135

2. Noble GasesOther fkissne productsG. Conbiwnmct RecercuIafieR Sump DeCayand Gleanup1. Radieolgisal decay s~edit2. Cleanup e~ediD. VnlumA of OWeto in %hich AGtPty is Deposited (dil!uted)

2. Raccumuela~towateF, gal.Fmpeetedsma_90% ef- gap WinNlene93,96025,040Expte189, 48*0QMA9.25% of gap iodine-inyentor2g, 011g, 0 Ci02gA,492.

0-0DBA0.25g; of gap idn29, 0 Qill1g, 9 i929, 42.G05.68,-0 7igGyesNle~eYes.NOReyoes03,96003,060Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATETABLE 1.IS5 24 ho 1 of 5ASSUMPTIONS USED TO CALCUL.A.TE OFFSITE EXPOSURES FROM. POST LOCA.GIR1ULATION LOOP LEAK.GE IN11 TH AUXILIARY BUILDINGkxpec4ed Xetd0Q8D. Volumo of W.^tor in" Whi.',!a Activity icDeposited (diluted)

(Contd)3. Refueling V~teF etoroge tank, gal. 350,000 262,939 350,00025,0(Table 6.3441. Total, gal. 4690000 2&1-T030 469,000 373,220& Conditiono of Locp Leakage Water1. pH of loakage water 8-8 84 87(Figuro 6.2 15)2. Tempr.tur.

of leakage "ate°, 'r 420 238 420 242F. Loop Leakage Rate 19044 WhF q0-M 14Q GGA'* 50 gPm-8T~ .3 0) (Table 6.3 0)G. Duration of LGop LoakagoI. Time Altar LOCI \ leakage beginR, hr Q 27 0,207 0,209 24(Tbe6.3 5)2. Timo after LOCA leakage 7-20 0.837 720 24-ends, h3. Total~n duration of loop 74-970 7.40, 0leakage~h Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATET.ABL-R1E

16. 5 214 Sheet 1 of 5.A.SSU.MPTIONS USED TO CALCJ'..TE OFFSITE EXPOSUPES FROM POST LOCA LOOP LEAr"AC AUIUII ARY UDINGIIlkt3 H. Auxilioar Building bdineDerantaminatien Fa-tres1Elomontol iodinedcotoioto M liquid, ibm -b. An x I 0,V liquid, fas-c. PDoRtiion ooeficint, ,G-, 4 -4,0d. P g."t) gcn 48x0 -PF, ((g) 2 .DQq cnt*minotocn foe'ctor, 4-. 4- 4 4-4 G--"OF, (a)fGWl, Dk(9) gasI. Auxiliar; Building Dr'c..Plateewt, and FiltcrRemonva1.

decoy eGcdit NeRe NeRe Nene Neae-2. Plat"ot credit Nene NeAe Nee NeOreRevision 12 September 1998 DCPP UNITS 1 & 2FSAR UPDATETAB3LE 15.5 24 ho1o 5.ASSUM1PTIONIS U'SED TO CAI ll ATE OFFRS!TE EXPOSURES FROM4 POST LOC ACIRCULATION LOOP LESAXA.GE IN THE --AU[XIL'A.RY BUILDING-Expected Expesede OBA BI.Auxii~ar; Ruilding De~ay,Plateout, and FilterRcmeval (Contd)3.Auxii~ary bUilding Nos;e YsNon~eYe A. AzdOA filtprAffigiznGy (1) EmlemonbI joGdin, PA, (2) Orgqr.ic edinc, %0 8 0-0(3) Pzartisubate oie%0090009-b- Nome gase~c0 ~ ~j. Atm~epheki Dipepmian

1. DownYA '.Ad Fadlalagiza 01Gie NO~e Nleme Nenederzay Gcedi2. AtmoephewiG dilutien Table 15.54 Tahle 4 5 64 Table-4&&-

4 Tabie 45. 4K. Brzthing RatesTbe

.Table 166 7 Table 4 5A Table 45.5 7Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 26PERCENTAGE OCCURRENCE OF WiND DIRECTION AND CGALM WINDSEXPRESSED AS PERCENTAGE OF TOTAL HOURLY OBSERVATIONS WITHINEACH SEASON AT THE SITE (250 FOOT LEVElL)Offshoefe{b+

0hOFsOre-ý Gak:n-(,*

AnnualwetTranskoa57-05"05"0'6"03"%4"042%"0&4%40k(a) Dry Season May through September Wet Sea6Gon November through MarchTransitional April and Octobcr(b) Off-shore Wind directfionAsare defined as wind directions fromn nO~hwcst through east southeast measured cock:-ise.

(G) Onshoere wid directions are defined-a wind directions from southeast through weSt no.hwestmneasured clockwise.

(d) Calm wind directions are defined-as w.Ainds with speeds, less than one 1 mnph.Revision 11 November 1996 I II )I I pII I ~II II I I ~ I~JIll 0-UCzIII III I +/-411111 C')N)mI C,)I III II I I~P? -UHmII II III 1111 +1CDII ~aC,)I CDCO IIIIIIIIIIIIIICOOVOC 0CO LA I. 1.1k0) +4 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 27 Sheet 2 of 2CRN NNW 14 _a_A D) W 4 , A 0 w T A 0 4 0406 94086 0 457 04062 0-094 03473 04 42 ano9n3 90-048 0-064 0437 0,0564 03064 a0EM 0-44 0-0496 0-066 0-040 0-08G 0 0866 a-06a G-0461 0-Ogg 0-0667 0-450 0.046 0-049 0-0666 13044 0-03 0-046 0-0729 0-045 0-0144 a-029 030"40 Q 036 0-044 0-002 0-04944 0-046 0-060 DO0 0-05442 0425 0028 0349 0-03943 0336 04464 0-041 9044444 0036 0-043 0044 034546 0349 Q0270 0340 0-04346 0-020 .0325 0-000 032644 0.322 00334 03000 00144489 0-023 0-0133 0-046 004449 cl 030 0034 0346 0-04620 0-043 owl4 -0300924 0-003 0300 -0300022 03047 0906 -. 0304623 04004 03006 -cap0240042 0042 -0026 a0342 0.049 --26 0-042 0320 -24 0004 0304 -28 0-004 0304 -29 0.0 0-00 -30 0300 90300 934 0.006 0-008 9(a) A~ -RW0 -D~ caca (May-, thoaugh £frpt~rbe-)

T TF-M:tGoa M41z(p'z-1Otbr 364 0463 a3 03490494 0404 0=4 0448042 G 0246 0-46 04400369 0360400 0-0420364334024033 0746 -0023 002403ON -OGG6 0.02i030 -0 3.020407 --3i0.039 04 Q 3004 0704 04 GA4420302 0-424 0-247 0.22403g 8034- 07 04440404 03 0422 00-44 03.= 03022 04490-344 30 -0.00 --04i --A 0- W 403944 a3900236 04260233 0.-24 0339 03-244044r2041660154 0 4040-40a 303al0494 03540380.374 0460466a0Og46l0320300 034400434 60260 30 -0.00000QOg0000

-9033 002340334

-Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 2 8ASSUMPTIONS USED TO CALCULATE ONSHORE CONTROLLED CONTTAI.MENT VENTINGSheet 1 of 2DBA-GaseA. Aztovily Released to Contaiwnment Atrnespherc

-4.-4edi~e a.......

nta1c. Particulate

2. Noblol garses3. Other fi 6ion prod-ctcB3. Decay, Cleanup, and Leakage in Contaiwnmnt

-A#Resphewe

1. Radiologieal decay cr.edit2. spray cleanupa. ElRmental

b. Organic'G. Particulate

3. Filtr cneanue onf containment atmeoshere 25% of gap iedine invento.y 24.95% of gap iodine invento.'

0.05% of gap iodine invento,"

0% of gap iedine inventoe.-;

1090% of gapinntryes0Ne~eNoeR0.05%ipot day,25%9 o-f core iod-ine inventeor 22.75%A of Gaero iodine inVentRI1.0% of core icdine in:entor; 1.25% of core iedinc inventer; 100% of core inventor; NesoNe6e0.05%/per daya. lodine6h- Noble gases4. Contafinment leak FateRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 28 Sheet 2 of 2ASUPIO!NS I~ ISE TO ALCULAT\

wlwwý rr~~-cE UNT-o i FOL COUNTAI NMENT A GALrJITING~

2&P4O%~b4cee DB.A.Case AG. Containment Atmcsphc."c VolumeD. Purge Schedule1. Time a#Fr LOC.A purging bogine2. Time after LOCA purging ends1968 hou., Chapter 66782 Mum, remainder of 1 Yrt.10 Q 4m, Chapter 667-2 Chapter68088 houm, e.. aindc. of 1 yr.25 cf, .hapter 65E. Purge FleorateF. Filter Efficency

!. -4edMnesa. Elemental

b. Or~ganicc. Particulate

2. Noble gasesG. Atmesphenc Dispemrsi,-

1. Radioelgical decay credit2. %iQcW. Breathing Rates(a) Although aIt , cuben..........

2 hr.... e...466.insigniflcant (Referenco 39).Q"%9"%4"0eQ"%7"0We"eNeReTable 15. 0WeReTable 46.6~valuafien shcv.ed t~hat the Design Cace eoeffid9Mn Of 34 (0Fo 2600 gpm spray header POeW) should be reduced togpmA spray, headr fleW), the potential effcilo doe increase due to thic Change is exteremely small and Gan be cencidered Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 29ONSHORE CREGONTROLLED F QC GONTAI NIEA4FeNIT NlVE NTIT-l N'G EXPOSU WRE-SThyroid eXpo..re at site 241-boundar; (800 m.eters),

rem"Whole~ body exposur 0.0841-7.&*G at site boundary (800 m~eters),

remRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 30ATM1OSHPERIC DISPERSION FACTORS FOR ONSHORE CONTROLLED CONTlAINlMENT VENTING (STABILITY CATEGORY D)QistaR~ee-m

%Zf37 .x I e--414 2.40-x 44- 77.884 wx4047.0 3.135 x 0--4Q-0

-2Q7 6.099 *404MeteGOrglogcal Input ParamneterS:

Height of releasc -70 mnetercMixing depth 5-0m~eters ManR; wind speed -5.83 meters per secondSigma theta -10 degreesSigma phi -3 degreeSVertic-al expansion rate beta, 0, -0.9Azimuth eXpansion rate alpha, et, -0.9Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-23ALOSS OF COOLANT ACCIDENTAssumptions

& Key Parameter ValuesParameter ValueCore Power Level 3580 MWt(105% of the rated power of 3411 MWth)Fuel Activity Release Fractions Per Reg. Guide 1.183 (See Section15.5.17.2.2.2)

Fuel Release Timing (gap) Onset: 30 secDuration:

0.5 hrFuel Release Timing (Early-In-Vessel)

Onset: 0.5 hrDuration:

1.3 hrCore Activity Table 15.5-77Chemical Form of Iodine released from fuel to 4.85% elemental containment atmosphere 95% particulate 0.15% organicChemical Form of Iodine Released from RCS 97% elemental and sump water 3% organicContainment Vacuum/Pressure Relief Parameters Minimum Containment Free Volume: 2.550E+06 ft3Primary Coolant Tech Spec Activity Table 15.5-78Chemical Form of Iodine Released 97% elemental; 3% organicMaximum RCS flash fraction after LOCANoble Gases 100%Halogens 40%Maximum containment pressure relief line air 218 actual cubic feet per secondflow rate (acfs)Maximum duration of release via containment 13 secpressure relief lineRelease Point Plant VentRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATEContainment Leakace Parameters Containment Spray Coverage

-Injection Spray 82.5% (sprayed fraction) and Recirculation Spray Modes:Sprayed Volume 2.103E+06 ft3Unsprayed Volume 4.470E+05 ft3Minimum mixing flow rate from unsprayed tosprayed region:Before actuation of CFCUs 2 unsprayed regions/hr After actuation of CFCUs 9.13 unsprayed regions/hr Minimum duration of mixing via CFCUs Start = 86 secEnd = 30 daysContainment spray in injection modeInitiation time 111 secTermination time 3798 secMaximum delay between end of injection 12 min (based on manual operatorspray and initiation of recirculation spray action)Containment spray in recirculation modeInitiation time 4518 secTermination time 22,518 secLong-term Sump Water pH > 7.5Maximum allowable DF for fission product Elemental Iodine: 200removal Others: not applicable Elemental iodine and particulate spray removal See Table 6.2-32coefficients in sprayed region during bothinjection spray and recirculation spray modesElemental iodine removal coefficients due to See Table 6.2-32wall deposition Particulate removal coefficients in unsprayed See Table 6.2-32region due to gravitational settlingContainment Leak rate (0-24 hr) 0.1% weight fraction per dayContainment Leak rate (1-30 day) 0.05% weight fraction per dayRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATEContainment Leakage Release Point(Unfiltered)

From the worst case release point ofthe following:

Diffuse source via thecontainment wallVia Plant VentVia Containment Pen Area GEVia Containment Pen Areas GW& FWESF System Environmental Leakacie Parameters Minimum post-LOCA containment water 480,015 gal.volume sourcesMinimum time after LOCA when recirculation is 829 secinitiated Duration of leakage 30 daysMaximum ECCS fluid temperature after 259.9 OFinitiation of recirculation Maximum ECCS leak rate (including safety Unfiltered via plant vent = 240factor of 2) cc/minUnfiltered via Containment Penetration Areas GE or GW & FW= 12 cc/minRHR pump seal failure Filtered(')

via plant vent 50 gpmstarting at t = 24 hrs for 30 minIodine Airborne Release Fraction 10%Auxiliary Building ESF Ventilation System filter Elemental iodine: 88%efficiency Organic iodine: 88%Refuelingq Water Storagqe Tank (RWST) Back-Leakage Parameters Earliest initiation time of RWST back-leakage 829 secMaximum ECCS / sump water back-leakage 2 gpmrate to RWST (includes safety factor of 2)RWST back-leakage iodine release fractions See Table 15.5-23CRWST back-leakage noble gas, as iodine See Table 15.5-23Cdaughters, release rate from the RWST ventRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATEMiscellaneous Equipment Drain Tank (MEDT) Leakage Parameters MEDT inflow rate (includes safety factor of 2) 1900 cc/minMEDT leakage Iodine release fractions See Table 15.5-23DMEDT leakage noble gas, as iodine daughters See Table 15.5-23Drelease rate from plant ventCR Emernency Ventilation:

Initiation Signal/Timing Initiation time (signal)

SI signal generated:

6 secNon-Affected Unit NOP IntakeIsolated:

18 secAffected Unit NOP Intake Isolatedand CRVS Mode 4 in full Operation:

44.2 secBounding Control Room Atmospheric Table 15.5-23BDispersion Factors for LOCANote:Releases from the RHR Pump Seal failure are filtered for CR dose evaluation and SiteBoundary Dose Evaluation.

Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE15.5-23BLOSS OF COOLANT ACCIDENTControl Room Limiting Atmospheric Dispersion Factors (sec/m3)Release Location

/ Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 6-720I hrControl Room Normal IntakesPlant Vent Release-Affected Unit Intake 1.67E-03

-- ---Non-Affected Unit Intake 9.1OE-04Containment Penetration Areas-Affected Unit Intake 6.84E-03-Non-Affected Unit Intake 2.24E-03Control Room Infiltration Plant Vent 1.26E-03 8.96E-04 3.44E-04 3.44E-04 2.99E-04Containment Penetration Areas 3.22E-03 1.85E-03 7.29E-04 7.15E-04 6.64E-04RWST Vent 1.07E-03 5.80E-04 2.18E-04 2.19E-04 1.79E-04Control Room Pressurization IntakePlant Vent 5.65E-05 3.70E-05 1.35E-05 1.37E-05 1.11E-05Containment Penetration Areas 6.45E-05 4.05E-05 1.65E-05 1.38E-05 1.12E-05RWST Vent 5.25E-05 3.03E-05 1.15E-05 1.10E-05 8.83E-06Note 1: Release from the Containment penetration areas (i.e., areas GE or GW & FW):applicable to containment leakage and ESF system leakage that occurs in the Containment Penetration AreaNote 2: Release from Plant Vent: applicable to ESF system leakage that occurs in the Auxiliary

building, MEDT releases, RHR Pump Seal Failure Release and Containment Vacuum/Pressure Relief Line ReleaseRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-23CLOSS OF COOLANT ACCIDENTRWST Iodine Releases Fraction and Gas Venting Rate to Atmosphere Average IntervalIodine Release Fraction Weighted Gas Spaceto Atmosphere Venting Rate toAtmosphere Sec Sec Fraction Ireleased I lentering Fraction Vrst / day829 7200 9.451E-05 2.610E+00 7200 28,800 6.357E-05 7.291E-01 28,800 86,400 8.796E-06 7.375E-02 86,400 345,600 4.560E-07 9.955E-03 345,600 471,600 6.347E-07 1.311E-02 471,600 1,011,600 8.231 E-07 1.489E-02 1,011,600 2,048,400 1.114E-06 1.547E-02 2,048,400 25920001.483E-06 1.702E-02 Where:Ireleased

= Total Iodine mass released to atmosphere during specified time interval, gmlentering

= Total Iodine mass entering to the RWST during specified time interval, gmFrac. Vrt = Rate of Fractional RWST gas volume vented during specified time intervalRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-23DLOSS OF COOLANT ACCIDENTMEDT Iodine Release Fraction and Gaseous Venting Rate to Atmosphere Iodine Release Average Interval WeightedFrom Time To Time Fraction to Gas Space Venting RateAtmosphere to Atmosphere Sec Sec Fraction

'released

/ Fraction VMEDT / daylentering 829 7,200 4.521E-07 5.024E+00 7,200 28,800 1.386E-08 3.024E-02 28,800 86,400 2.362E-07 3.324E-01 86,400 183,289 3.950E-07 6.497E+00 183,289 345,600 1.236E-02 (Note 2) (Note 1)345,600 752,400 2.028E-02 (Note 2) (Note 1)752,400 1,530,000 2.390E-02 (Note 2) (Note 1)1,530,000 2,592,000 2.166E-02 (Note 2) (Note 1)Where:Ireleasd

= Total Iodine mass released to atmosphere during specified time interval, gmlentedng

= Total Iodine mass entering to the MEDT during specified time interval, gmFrac VMEDT = Rate of Fractional MEDT gas volume vented during specified time intervalNote 1: After the MEDT overflows at t = 183,289 sec, the gas venting rates are 2640 cfm fromthe EDRT room, and 1760 cfm from the U1/U2 Pipe Tunnels (i.e., the exhaust ventilation ratefrom the respective rooms + 10%). To be consistent with the methodology used to determine the iodine release fractions after spillover, the noble gases generated by decay of iodines in thetank and spilled liquid after overflow occurs, should also be released instantaneously to theenvironment without hold-up.Note 2: The room ventilation flows addressed in Note 1 (utilized as clean in-coming air) areincorporated into the determination of the iodine equilibrium concentration in the EDRT roomand U1I/U2 Pipe Tunnels air space, respectively.

The bounding iodine release fractions presented above after spillover assume instantaneous release of iodines to the environment without hold-up in the room.Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5-23ELOSS OF COOLANT ACCIDENTTSC Limiting Atmospheric Dispersion Factors (sec/mn3)Release Location

/ Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-20hrTSC Normal IntakesPlant Vent Release 5.52E-04

---

----- ----- -----Containment Penetration Areas 1.80E-03RWST Vent 3.63E-04

---

----- ----- -----TSC Infiltration Plant Vent 5.43E-04 2.16E-04 9.97E-05 8.11 E-05 6.58E-05Containment Penetration 1.83E-03 7.49E-04 3.16E-04 2.92E-04 2.41 E-04AreasRWST Vent 3.72E-04 1.68E-04 6.64E-05 6.17E-05 5.1OE-05CR/TSC Pressurization IntakePlant Vent 3.70E-05 1.35E-05 1.37E-05 1.11E-05Containment Penetration Areas 4.05E-05 1.65E-05 1.38E-05 1.12E-05RWST Vent 3.03E-05 1.15E-05 1.10E-05 8.83E-06Note 1: Release from the Containment penetration areas (i.e., areas GE or GW & FW):applicable to containment leakage and ESF system leakage that occurs in the Containment Penetration AreaNote 2: Release from Plant Vent: applicable to ESF system leakage that occurs in the Auxiliary

building, MEDT releases, RHR Pump Seal Failure Release and Containment Vacuum/Pressure Relief Line ReleaseRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 31CONTROL ROOM IN.4 F -,^TRATION A ,SUMAE D FOR RADI _,LGICAL EXPOSURE

-ALl,, ,TION'SSheet 1 of 2L~eakaqe Pa;th.PWiedowsR. Dnf sG. Pcnctratioflns

1. Ducting (extcm~al real)No leakage, no windows.No leakage; durting penetrations caulled ful depthand exterior su'faees scealed with FL-,MEMA~STlC 7!Aand- con.trol room will be poitively presritwzed.

No lcakago:

conG.erte wallS and floor poured with pipingi n placae and cenRal--

rooem -he b positively pprnssurli-ze-d-.

No space between exposed conduto.s and trayssealed with RW R-01OWOOL ceramic fiber 6 inchos indepth, Yvth two coats of FLAMEMP.STIC 72A, and controlroomR Will be poSitively pressr.Uized-No leakagwe:

cnduits arc sealed with THIXOT-ROPIC silicnsrubber com.pound, with a F minimum depth of one diameter, and control room will! bo positivel, preessurized.

O-.G0-03. Conduits and traysa. EwtemFalcreal

b. Wnernal SeaRlRevision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATETABLE 15.5 31CONTROL ROOMK INF!LTP.A.TIO.N ASSUMRED FOR P..,DIOLOG!CAL EXPOSURE CALCUL.ATIONS Sheet 2 of 2D. Damperswhere;Q -leakage, GfmA -damper arca, square feetq- leakage per unit damnper area per in. of wateor"Ap -peresure difference aGrOcS damper i. F waorA -6.n00- f?,q-0-0,0

,1 h,,, in. and Ap -6.0 n W.G........ q -and Ap -6.0 in.G.A- -60 42-,-"q -nn-00n*t/-

in. and Ap -6.0 .W.G.a -nd 8 q -n I A-,, a, ,p -6.0 in. "G.I. Mode damper#22. Mode damper #33. Mode damper#74, Mode damper #8Ei.Total(a) FromF mpanufacturc (b) Assume Geoserwa(G) 10 cfmR is canser':.FS puB1snsu 0013.tively large value of 6 inches of water; damperc will never see a p~essure differential this large.afively assumed in the ana!l'cis-Revision 11 November 1996