ML15176A534

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 UPDATE Per Regulatory Guide 1.183, the radioiodine released from the fuel gap is assumed to be 95% particulate (Csl), 4.85% elemental, and 0.15% organic. Due to the acidic nature 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 of iodine to 99.85% elemental and 0.15% organic. In addition, and per Regulatory Guide 1.183, 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 in the available air space. All of the alkali metals released from the gap are retained in the pool. In accordance with Regulatory Guide 1.183, the chemical form of the iodines above the pool is 57% elemental and 43% organic.Per Regulatory Guide 1.183, the activity released due to an FHA is assumed to be discharged to the environment in a period of 2 hrs (or less if the ventilation system promotes a faster release rate).FHA in the FHB The radioactivity release pathways following an FHA in the FHB are established taking into consideration the following Administration Controls: During fuel movement in the FHB: a. The movable wall is put in place and secured b. No exit door is propped open c. One FHBVS exhaust fan is operating (The supply fan flow (if operating) has been confirmed by design to have less flow than the exhaust fan)Operation of the Fuel Handling Building Ventilation system (FHBVS) with a minimum of 1 exhaust fan operating and all significant openings administratively closed will ensure negative pressure in the FHB which will result in post-accident environmental release of radioactivity occurring via the Plant Vent. The activity release due to the FHA in the FHB 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 the FHBVS with a closed FHB configuration.

b. A maximum conservatively assumed outleakage of 500 cfm occurring from the closest edge of the FHB to the control room normal intake (i.e., 30 cfm outleakage 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 lambda based on the FHBVS exhaust (i.e., 8.7 hr 1) is larger than the release rate applicable to"a 2-hr release" per Regulatory Guide 1.183 (i.e., 3.45 hr 1). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.15.5-127 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE FHA in the Containment The potential radioactivity release pathways following a FHA in the containment are established taking into consideration

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

In the event the containment purge system ceased to operate (a viable scenario since it is single train and has non-vital power), the density driven convective flow out of the equipment hatch (due to the thermal gradient between inside and outside containment conditions), could be significant.

It has been determined that for the FHA in the Containment, the release rate assuming a regulatory based 2 hr release is larger than that dictated by the containment purge ventilation 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 the atmospheric dispersion factors associated with the plant vent vs the equipment hatch indicates that dose consequences due to releases via the equipment hatch will be bounding.15.5.22.2.3 Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident 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 unit is 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)15.5-128 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Credit is taken for PG&E Design Class I area radiation monitors located at the control roomcontrol 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 at the control room normal intakes as a result of an FHA.An analytical safety limit of 1 mR/hr for the gamma radiation environment at the control room normal operation air intakes has been used in the FHA analyses to initiate CRVS Mode 4. Note that the actual monitor trip setpoint is lower to include the instrument loop uncertainty.

The radiation monitor response time is primarily dependent on the type of monitor, the setpoint, 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 of 0.05 mR/hr, the response time "t" to a high alarm Setpoint (HASP < 1 mr/hr), for a step increase of radiation level DR (mR/hr) is determined by solving the following equation that represents the monitor reading approaching the final reading exponentially.

HfAS?= 0.05 OLR (I. -I-0 It is determined that a DBA FHA (i.e., occurs at 72 hrs post shutdown) will result in a radiation environment at the control room normal operation intakes that greatly exceed the analytical limit of 1 mR/hr for initiating CRVS Mode 4. This will result in an almost instantaneous generation of a radiation monitor signal to initiate CRVS Mode 4 (radiation monitor response time is estimated to be < 1 sec). For purposes of conservatism, and since the delay in isolation of the normal intake has a significant impact on the estimated dose consequences, the analysis conservatively assumes a monitor response time to the HASP of 10 secs.As discussed in Section 15.5.1.2, when crediting CRVS Mode 4, the FHA dose consequence 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 DBA FHA at both the control room normal Intakes (assumed to be instantaneous) and CRVS Mode 4 operation is estimated to be the sum total of the monitor response time (10 secs), the signal processing time (2 Secs) and the damper closure time (10 secs) for a total delay of 22 seconds.Delayed FHA: It is recognized that the response time for radiation monitors are dependent on the magnitude of the radiation level / energy spectrum of the airborne cloud at the location of the detectors, which in turn are dependent on the fuel assembly decay time. Thus an additional case is considered for each of the two FHA scenarios described above 15.5-129 Revision 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 the control room operator; i.e., a case that reflects a delayed FHA at Fuel Offload or a FHA during Reload, occurring at a time when the fuel has decayed to such an extent that the radiation environment at the control room normal intake radiation monitors is just below the setpoint; thus the control room remains in normal operation mode and CRVS Mode 4 is not initiated.

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

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

The x/Q values presented in Table 15.5-47B take into consideration the various release points-receptors applicable to the FHA to identify the bounding Z/Q values applicable to a FHA at either unit and at either location, 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 is presented 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 for an" two hourI period following the onset of the porstulated fission product release 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 not re 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 4in excess of 0.05 Sy (5 rem) total effccti':o dose equiv.alent (TEDE) forth dIuration 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 of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-47.15.5-130 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE (2) The radiation dose to 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 (during the entire period of 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 the accident 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 area for the tw.v hours immediately following the onset of the potulated fis ion product release shall not receive a total radiation dose to the whole bod-YL in exes of 6 rem or a tota! radiation dore in excess of 75 rem to the thyroid 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 fcorom the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess Of 6 rem, or a total Fadiatien dose in excess Of 75 rem to the thyroid from iodine exposue.(2) In accordance with the requirements of G.. 19, 1971, the dose to the con room operator under aGcident conditions shall not be in excess of 5 rem whole body or its equivalent to any pa~t of the body (i.e., 30 rem thyroid and beta skin, 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 the Containment aFre mitigated by containmeRt crloure. The foallowing evaluatio-n shows that in all cases the calculated exposures would be well below limits specfd in 10C FR 100.11.15.5-131 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Drrring fuel handling opertienG, containment c",lore is not the o-,nt-ainmenet ventilation purge system is oeprationa!

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

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

The most promFinent of these penetrations are the equipment hatch and the persone airlock. Closure of these penetration6 is achieved by manual mneans as discaussed i Section 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 eachu remains in the norm~al moede of operation following the FHA. Thus-, the design basis FHA does not credit charcoal filtration of the control room atmosphereintake flow or reci~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 of Safety Guide 25, Ma.rh 1972, were used as guidance w..ith the exceptions detailed below.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 those dscussed 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 all the gap activity in the damaged rods is released aRd consists of 40 of the total noble gases other than KIr 85,30 of the Kr 85, and 10 percent of the tol radioactive iodine in the rods at the time of the accidet An 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). The use of these dose conversion factors is consistent with the current guidance provided in Regulator,'

Guide 1.183, July :2000.15.5-132 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.6.22.2.2.2 Containment CIoSUre FlolloWing the postulated, accident, ativity evolves from the surface of the peol where it m.ixes with air above the pool. Airboene actiVitY is then to be discharged to the envronment via the open penetrations.

The- dluration of thereae was 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 would initiate 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 inteerco system 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 and retained within the approximately 33,600 cubic foot rectangular parallelcpipcd formed by the 25 by 70 foot pool and the 40 foot high steam gene raters. W~here the steam generators do not surround the pool, the radioactivity would actually be dispersed into a larger volume of air which would have the effect of reducing the dose. HowAev~er, for conser.atism, it was- assumed that all the radioactivity remnained within this 33,600 cubic foo0t Volume and was then transperted to the environme~nt w~ithin a two second 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 the atmopher 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 i These exposures are well within the 10 CFR 100.11 limnits.15.5.22.2.2.5 Acltion Fo-llowinf*g Containment Isolatio Following manual containment closure after the fuel handling accident, activity can be remoedwe from the coant-ain~ment atmosphere by the redundant PG&E= Design Class 11 iodine Remoeval Syte (to rins at 1 2,000 cfmn per train), which consists of-15.5-133 Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATE HE=PA,'charcoal filters. This system is deGcribed in Section 9.4.5. There are no Technical Specificaio " oquirements forF this fitration system.The contafinment can also be purged to the_ atmosphere at a controlled rate of up to 3-00 c-fm per train through the HEPNAihaFcoal filters of the hydrogen purge systemn. This sYStemn 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 hours immediately following the onset of the postulated fission product release are well 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 is exposed to the radioactiye cloud resulting fromn. the postulated fission product release (during the entire period ofits pasg) aewl elow the dose limits of 4 10 4R100.41 as shown in Table 15.5 50g.(3) In accordance with the of GDC 19, 4971, the dose to the contFroroI .o..perat under accident conditions shall not be in exGess of 5 Frem whole body or its equivalent to any pa~t of the body (i.e., 30 rem thyroid and beta skin, Reference

51) for the duration of the accGident as showni Table 15.5-50.15.5-134 Revision 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~delS and assumptions specified in pe~tinent regulator; guide,. In allw anlss h esulting potential exposures to individual mnembers of the pulic nhe eneral population have been found to be lower than the applicable guidelinies an~d HFM6t specified i 10 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 accident in 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 o 15.5.23 RADIOLOGICAL CONSEQUENCES OF A CONTROL ROD EJECTION ACCIDENT 15.5.23.1 Acceptance Criteria The radiological consequences of a CREA shall not exceed the dose limits of 10 CFR 50.67, and will meet the dose acceptance criteria of Regulatory Guide 1.183, July 2000 and outlined below: EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release 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 low population zone, who is exposed to the radioactive cloud resulting from the postulated 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 Criteria Adequate radiation protection is provided to permit access and occupancy of the control 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 exceed the dose limits of 10 CER 100. 11 as outlined below: 15.5-135 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE i. An individual IOcated at any poeint on the boundary of the eXGlusin area for the two hours imma'i ' following the onset of the potulated firsionl 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 the thyroid 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 its passage), shall not receive a total radiation dose to the whole body in exGess Of -6 rem, 9r a tetal rauiauon Ise in thyroid fom iodine expeosue.excess ei 301D reri to the (2) in accordanGe with the requirements of GD. 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem) for the duration of the accident-.

15.5.23.2 Identification of Causes and Accident Description 15.5.23.2.1 Activity Release Pathway As discussed in Section 15.4.6, this event consists of an uncontrolled withdrawal of a control rod from the reactor core. The CREA results in reactivity insertion that leads to a core power level increase, and under adverse combinations of circumstances, fuel failure, and a subsequent reactor trip. In this case, some of the activity in the fuel rod gaps 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 could leak to the environment.

Following reactor trip, and based on an assumption of a Loss of Offsite Power coincident with reactor trip, the condenser is assumed to be unavailable and reactor cooldown is achieved using steam releases from the SG MSSVs and 10% ADVs until initiation of shutdown cooling. DCPP has established that the LOL event generates the maximum primary to secondary heat transfer and the CREA assumes these same conservatively 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 the radiological consequences following a CREA.The CREA is postulated to result in 10% fuel failure resulting in the release of the associated gap activity.

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

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

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

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

ulnd4er ad'v.Frse ,mbinations Of circ-umstances, omen fuel -.addIng failureS could occur following a rod ejection In this case, some of the actiVitY in the fue! rod gaps would 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 the enafonment Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose 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 and 20 gallons per day to the secondary was assumed.Follo9Wing a postulated rod ejection accident, activity relearsed from the fuel elle laddin na uet failure of 1 nercIntII of;ll the fuel rod islU asume to be inrstantaneouslY released to the 'r mr; colant. Releases to thefprimar,'

coolant are assumed to be immediately and uniformly mnixed throughout the cooelant.The actiVity released to the containment ftrom the primP,'y coolant through the ruptured onto 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 of the elemental i;,ono leaked to the rcoolant is released to the containment atmosphere as a result of flashing Of some of the coolant water. Of the amorunt Of noble gases released to tho primary coolant, 100 prcent is as.umed to be released to the containment 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 hs release 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 UPDATE Following the release to the containment, the fission products are assumed to leak f7ro the col-ntainment at the same rates assumed for the ..., LhA, discused in Section 15.5.17. In addi-tion, the spray system is assumed to be in operation and acts to remove the i-d-inecs fro-m t-he coentainm~ent atmosphere at the same rates assumed for the -LBL-OCA.The assumptions used for breathing rates, population density, and other commonA fac-tors were also desc-ribead-in earlier sections.

Bo0th the prm; n secon~dar; coolant activities prior to the accident are given in Sectio"n 154.5.3 The gap activities 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 are give inTabe 1.5 2. hyrid doses that woul1d result from secondar; steam releases can be determ-ined from, Figues 15.5 2 and 15.5 3 for the DBA conditions and Figures 15.5 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 gie param1etrically in Figures 15.5 1 and 15.5 15. it should be noted that these figures aFe based 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 acciden From these analyses, it as n be concluded that offite exposuwes frm this accident will be 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 of potential exposures to contrio room pefrsonelis mlade in Sectionta 15.517 for cnditions following a LOCA.15.5.23.2.2 Activity Release Transport Model The CREA dose consequence analysis evaluates the following two scenarios.

Scenario 1: The failed fuel resulting from a postulated CREA is released into the RCS, which is released in its entirety into the containment via the faulted control rod drive mechanism housing, is mixed in the free volume of the containment, and then released to the environment at the containment technical specification leak rate for the first 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 the RCS which is then transmitted to the secondary side via steam generator tube leakage. 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.15.5-138 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE The 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 not controlled following a CREA, it is conservatively assumed that the iodine released via the containment leakage pathway has the same composition as the iodine released via 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 the remaining 3% is organic.Scenario 1: Transport From Containment The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously released into the containment where it mixes homogeneously in the containment 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 for the remaining 29 days after the event. Except for decay, no credit is taken for depleting the halogen or noble gas concentrations airborne in the containment.

Per Regulatory 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 is taken 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 released to the environment from the secondary coolant; i.e.; 97% elemental and 3% organic.Environmental releases due to containment leakage can occur unfiltered as a diffuse source from the containment wall, and as a point source via the containment penetration areas or the Plant Vent. The dose consequences are estimated based on the worst case atmospheric dispersion factors, i.e., an assumed environmental release via the containment penetration areas.Scenario 2: Transport from Secondary System The failed fuel activity released due to a CREA into the RCS is assumed to be instantaneously and homogeneously mixed in the reactor coolant system and transmitted to the secondary side via primary to secondary SG tube leakage. The activity 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 failed fuel release, and are therefore not included in this assessment.

DCPP Plant Technical Specification 3.4.13d limits primary to secondary SG tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate 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), has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant; therefore, the gap iodines have a partition coefficient of 100 in the SG. The gap noble gases are released freely to the 15.5-139 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE environment 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 from steam 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 System are elemental, while the remaining 3% are organic. The SG releases continue until shutdown 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 the accident) 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 any point on the boundary at the EAB, for any 2-hr period following the onset of the accident 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 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine and particulate level in the SG liquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the starting point of the worst 2 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 model are 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 is based 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 initiate a 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 for Containment High Pressure is reached in 150 seconds after the SBLOCA. As indicated earlier, releases to the containment following a CREA are through a faulted control rod drive mechanism housing. The control rod shaft diameter is 1.840 inches and the RCCA housing penetration opening is 4 inches in diameter.

Based on the above 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 to the 2 inch SBLOCA, or 300 seconds.15.5-140 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Based 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 28.2 second delay due to delays associated with diesel generator loading onto the 4kv buses. The control room dampers are fully closed 10 secs later, or at t=338.2 secs (i.e., 300 + 28.2 + 10). The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay.c. In accordance with DCPP licensing basis, the CRVS normal operation dampers of the non-accident unit are not affected by the LOOP and are isolated at t=312 secs (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 the control room normal intake of the affected unit and because the releases from the MSSVs/10%

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

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

ADVs.The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to a CREA at either unit are provided in Table 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 bounding x/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 the notes of Tables 2.3-147 and 2.3-148.The bounding control room dose following a CREA at either unit is presented in Table 15.5-52.15.5.23.3 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 of the exclusion area for any 2-hour period following the onset of the postulated fission product release is within 0.063 Sv (6.3 rem) TEDE as shown in Table 15.5-52.(2) The radiation dose to 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 (during the entire period 15.5-141 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE of 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 the accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-52 comparing the actiVity relea.es following a rod ejection accident, given in Table 15.5 51, with the activity releases calculated for a LOCA, given in Table6 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 as follows-(1) The radiation dose to the whole body and to the thyroid o-fan in~divid-ual located at any 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 dos limits 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 locateda any 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 the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown 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 o GDC 19, 1971, and discussed in Section 15.5.17.15.5.24 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A WASTE GAS DECAY TANK 15.5.24.1 Acceptance Criteria The radiological consequences of a rupture of a waste gas decay tank shall not exceed the 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 the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.15.5-142 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 15.5.27 REFERENCES

1. Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plant, N18.2, American Nuclear Society, 1972.2. Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, US Atomic Energy Commission (AEC), Rev. 1, October 1972.3. Regulatory Guide 4.2, Preparation of Environmental Reports for Nuclear Power Plants, Directorate of Regulatory Standards, AEC, March 1973.4. W. K. Burnot, et al, EMERALD (REVISION I) -A Program for the Calculation of Activity Releases and Potential Doses, Pacific Gas and Electric Company, March 1974.5. S. G. Gillespie and W. K. Brunot, EMERALD NORMAL -A Program for the Calculations of Activity Releases and Doses from Normal Operation of a Pressurized Water Plant, Program Description and User's Manual, Pacific Gas and 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 Number TID-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 the Containment Systems Experiment," Nuclear Technology, April 1971.11. G. L. Simmons, et al, ISOSHLD-I1:

Code Revision to Include Calculation of Dose 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 1970.15.5-150 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 13. 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, Plant Shutdown 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 Fluid Systems, American Nuclear Society, May 1974.17. K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criteria 19," 13th AEC Air Cleaning Conference, August 1974.18. M. L. Mooney and H. E. Cramer, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1970 (see also Appendix 2.3A in Reference 27 of Section 2.3 in this FSAR Update).19. M. L. Mooney, First Supplement, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1971 (see also Appendix 2.3C in Reference 27 of Section 2.3 in this FSAR Update).20. M. L. Mooney, Second Supplement, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1972 (see also Appendix 2.3D in Reference 27 of Section 2.3 in this FSAR Update).21. International Commission on Radiological Protection Publication 30, Limits for Intakes 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 Storage Facility 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 Tank Failure, USNRC, March 1972.15.5-151 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 25. 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 in the Wake of a Model IRndustial Complex," NUREG 0373, 178 28. 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 in 30.31.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 l1180 fromedR J Tune Exn........

+, NUREGIC,...;R 1171 US....RC, September,,,, 1080 32. R. Bhatia, J. Dodds, and J. Schulz, Building Wake ,/Qs for Post-LOCA Control Room Habitability, Bechtel Power Corporation, San Francisco, CA.33. Report on the Methodology for the Resolution on the Steam Generator Tube Uncovery 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 Product Files: Summary of Major Nuclide Data, October 1975.37. Standard Review Plan, Section 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR), NUREG-0800, USNRC, July 1981.15.5-152 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 38. Deleted in Revision 18.39. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment Spray Flow 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 Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 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,,autg D-9nGlnn;,., GC on,,monRGe of ige n Basoi;s ArG~riAnd e nts , I ; Afnr atI. Lie lNr Po9wer React9rz," M52003.Deleted 44-45. Deletedlntrntat-Ina!

Commission on PFrtection (IRP) Publicat;on

• 0n, ;mr+ c,-, ,,s .....I;,des b,,W,,k, -,, 07/!9 .7 45-A6. Diablo Canyon Units 1 and 2 Replacement Steam Generator Program -NSSS Licensing 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 NSSS Engineering 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 Power Plant, Unit Nos. 1 and 2 -Issuance of Amendment RE: Revision of Technical 15.5-153 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Specifications Section 3.9.4, Containment Penetrations (TAC Nos. MB3595 and 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 for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.56. NUREG-0017, "Calculation of Releases of Radioactive Materials in Gaseous and 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 Accident Consequence Assessments at Nuclear Power Plants", Revision 1.61. Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building 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 Computer Analyses 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, L03.15.5-154 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 66. 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 Removal And Dose Estimates.

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

72. Not Used 73. 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 of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)", Revision 1 76. NUREG-0800, SRP Section 15.2.6, "Loss of Non-Emergency AC Power to the Station Auxiliaries", Revision 1.77. USNRC Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG 0800, Section 15.6.5, Appendix B, Revision 1, "Radiological Consequences of a Design Basis LOCA: Leakage from 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 Used 80. NUREG 0800, 1988, Standard Review Plan, "Containment Spray as a Fission Product Cleanup System", Section 6.5.2, Revision 4.15.5-155 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE 81. 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

-Final Report," 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 System Pipe 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 30,1986.15.5-156 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TABLE !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"a G64-34 4-2-9 4-4 G-.26_ _ __7 14-& 2ý44 4.34 949W Ge444 O049"4.0 0-.00034 0492-9 Sf49 Q0.033 0,40 0,0029, 0,32 Sf40 0494-3 30-004-3 GO-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 UPDATE TABLE 15.5 2 (HISTtRIGAL)

RESU"LT-S OF STU'DY OF- EFFECTS OF PLUTONIUM ON ACCIDENT DOSES 30dy Thy4eid Ghal~ge 30dy WhOWeBd Ghafn*ein-Thd Chan~ge 4 2 houF Molae-Bedy Reilase fromqr gas denay taRk Loss of Feactor- prImary Goolant largle broa!kgeneator-tube-Fuptare aGGident Steam. lWne rupture 0-4 0 0-4 0-3-2-2*4 47-2-2 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 165 54 E-XPE-CTEr-D-PS CIETAMSHRCDLTO FACTORS (SEC!M4 04 §,49x41-0 4 a.49x4l-0 4-974Q9 -84D4 X4*4 24OX494 &g&5i-4 9-24 2,4&404 40W40 4-75*404 0ý A4E46*4 4 4-54-G4Q 724 OR 49*- 3,04x4G4 14_754-g-

&0x4w 2490404 4454*444 6-Q04-94 96:720 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 UPDATE TABLE 15.5-5 ATMOSPHERIC DILUITION FACTORS QRShe-9 Midpei~t QireGti9G4&

Se.+r t Ar~dFr9nm fla.mn.,RW*R QitRe .nuier, SSE IsW WSW w VwN N-W NNW-r--14,6 4,44 0-54 5,39 4,04 0-.64 048 072-6 04-8 0.40 4-760 076"-25-0722 046 04-6 044 4408 04-g-35-072-3 0-24-044 0708 0708 04-5 047-0,77 0,29 074-04-6 0708 046 Q-70 042 040 0-.2-04-5 04-3 0-706 G-.-_w-Q-.-..---...-5 04-0-44 0.48 At.o,.pheN.

Dilution Factors IQx 4 !O-seG--m 4 Downwind DistancG, meters 2000 400 7000 14000 2-g000 Di-eGtieR 800 E-G-.47-----4 GAG8 07050 0.035 0.0118 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-6 ASS UMED ONSIIT- ATMOSPHER-C DILUTION F-ACTDORS (SEC!W3}FOR THF CONTDO' RO)OM Base X14-Modifying F~al PeF~ed, is.A FEr The ,r--stien uase--8-4 8-24-9642 4-084*4-9 4,.084x40 4 4-783 7466 Ag 4-792.84.76.-2 7-2.-2 7-2 4 4 4 4-75 75 75 75 765 765 765 765~7,95404 34 -4 2,2749 B. For The Infiltration Case go.8-2-2-4-24-Q-9642 344W1-0 37-14*1-04 34.U41-04 4 7833 766g ,48 4.-9.674.-2.-2.-2 4-4-4-4-7-5.-5.-5.-5.-65.-65 7-65.-65 4-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 UPDATE TABLE 15.5-8~ltDl II ATICthI MICTDIII ITCltIKl SeGtG Midpei~t Di~eGUG9Rs SS-W-W WNW N-W NNV TePtak;Radial C pr.rOr A;Andpent )W~~ fln Qn.;nrlflaR-fln nIeS T-etal Seeto 47044 840 474 47943 0 0 4-,72-7-47966 20,334 45,666 267000 40,334 20,033 9-,70G 0 2L, 0 27000 000OG 5-00=67604 17900 22,933 21,334 21,333 57433 4-234 4-9M 7-0 4-.76 47066 49,734 4 8,666 22,267 466 7-00&4433 7-34 46,066 45,334 49,433 647 466 4--JWG 1400 634 47533 67-000&754W 44138 42,832 357904 23,066 38-173 87,609 7,034 6a9 2 33 67904 111,46090,600 7-8,800 56,600 24,730 368,944 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-9

SUMMARY

OF OFFSITE AND CONTROL ROOM DOSES FROM LOSS OF ELECTRICAL LOAD Thyroid Doses, rem Dose _(TEDE, rem)Smte Bounda.. 2 Hours DalysRequlator y Limit (TEDE, rem)10 CFR Pat 100 3W Wo Maximum 2-hour Exclusion Area 0 0.0066 Boundary Dose 1 Design basis case-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5 Spike 30-day Integrated Low Population Zone Dose-Pre-incident iodine Spike <0.1 2.5-Accident-Initiated Iodine <0.1 2.5 Spike 30-day Integrated Control Room 5 Occupancy Dose-Pre-incident iodine Spike <0.1-Accident-Initiated Iodine <0.1 Spike ExpeGted-oase 5-2-x40'Ahole Body Doses, rem St p "Il -A -Unsxr-10 CFR Pat 100 25 25 Design basic case 2. 2.444-G Fixpeeted Gase 7-~-4O44 .4 Population Doses, mnan rem Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE Design basir, ca.e Expected-ease-045 t-Note: 1. The maximum 2-hour EAB dose occurs during the following time period:-Pre-incident iodine Spike-Accident-Initiated Iodine Spike 0 -2 hours 8.73- 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-9A LOSS OF ELECTRICAL LOAD ANALYSIS ASSUMPTIONS

& KEY PARAMETER VALUES Parameter Value Power Level 3580 MWt Reactor Coolant Mass 446,486 Ibm Primary to Secondary SG tube leakage 0.75 gpm (total for all 4 SGs); leakage density 62.4 Ibm/ft 3 Failed/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 rate Duration of Accident-Initiated Iodine Spike 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Initial Secondary Coolant Iodine Concentrations 0.1 pCi/gm DE 1-131 (Table 4.2-1)Initial and Minimum SG Liquid Mass 92,301 Ibm/SG Time period of tubes uncovered insignificant Steam Releases 0-2 hrs: 651,000 Ibm 2-8 hrs: 1,023,000 Ibm 8-10.73 hrs: same release rate as that for 2-8 hrs Iodine Partition Coefficient in SGs 100 Iodine Species Released to Environment 97% elemental; 3% organic Fraction of Noble Gas Released 1.0 (Released without holdup)Termination of releases from SGs 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> Environmental Release Point MSSVs/I 0% ADVs CR emergency Ventilation

Initiation Control Room is assumed to remain on Signal/Timing normal ventilation for duration of the accident.Control Room Atmospheric Dispersion Factors Table 15.5-9B Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-9B LOSS OF ELECTRICAL LOAD Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release point and receptor 0-2hr 2-8 hr 8-10.73 hr MSSVs/10%

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

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

______________

I ___ I ___ I ___Note 1: Due to the proximity of the release from the MSSVs/1 0% ADVs, to the normal operation CR intake of the affected unit, and due to the high vertical velocity of the steam 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 the X/Q s presented reflect those applicable to the CR intake of the unaffected unit.Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TADI EC 4C: r 4n 01 IRkAfAAOV

%C: n~C:CITIrr nnCEQ 10(lA A SMALL LOS ORF C-OLANT ACCInENT NO FUEL IDAMAGE ThYroid Doese, r-em Si;+Dte Bounda 27 Lint ar 10 CER Part 1 00 De-gAbasiS Gase F~peGted Gase 300 2,04x_-W 3w0 2,7X4-4 10 Whole Body Doses, rem 2-tR OR-mmat 2nl.n H iIew 10 CFR Part 100 Design case Expeoted Gase.26 44-*4-0 25 4-.-x-4-0 4 Population Doses, man rem DesigR basis case 0.01-3 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 11

SUMMARY

OF OFFSITE DOSES FROM AN .MDEFP.EIE.Cv ACCIDEhT ThYroid Doses, rem Site O39nda 2n~n H naew[=p7= 3 10 CFR Pa;t 100 Design basic case ExpeGted Ga~e 0.024 44-4 30w 0.0066 1.2x4Q Whole Body Dozes, rema Site Ben, raw 2 Howsrc 10 FR Part 4100 Design basis ease ExpeGted Gase D)es6ig9A hn basirs- eA-FExpe~ted Ga~e-2-5 2-5 24.4-40 4 Dman rem 04-5 4.44-940 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 12

SUMMARY

OF OFFSITE DOSES FROM A SINGLE ROD CLUSTER CONTROL ASSEMBLY W!THDRA^WAL T-hyrold Deses, rem Sitp RAnundirv

2 WAi 10) CFR Part 100[Design baSis caso E*xpeGted rase 300 042 300 0.043 3-.4-94%Aflk-- 0-th- M --ar --n Z3118 159UREWY Z HOUFS 10) CFR Part 100 Decign basis case EDpeGned Gase Design ba6*6 ease FEXpeGted Gase~25 64-x-404 6,54-W4 25 6-74-40 4 6,- 4 Doses, mn Frem 4.3-40 4 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE_OW 0-4V _4 8 214 hr 241@v 06 hr _y 1424 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 -~14320RG 1 434GRG 144PAR I 132PAR I 433PAR 1424AR KF 88 Xe 133 Xe 43~m Xe 13 Xe435M Xe-42R 0.1424E-01 9.1780E 02 0.1314E 02 G44-3E-02 G-0 0-0 0-0 0.36GE 01 ().4285E01i

0. 503 E 01 0 5Q4r27 02 0.9285re00 0,6.62E-01

0.5 1289E

006,E-04 0.1556Er.02 0.2211 W-01 0.8366E-02 04-0.6227E00 n 0.4 2650 01 0.11 00E,02 041035E,02

0. 2210 E 03 0#T-4.E-04 0A670 E 05 0-4734E-02 0.0 0-0G 0 4257E,02 9.4689E-Qi 0A.2.12-E--.1 0.1 290E 01 0.-1.,,0E--3 0.6-162E-01 0 1826E-02-.7140.E4.0 0.1 490F 10 0.l384,0o 0.4313 lE 06 0,3 542F 02 0. 9077-E 12 0-0 Q.4-757E 01 a g571EF.02 0.:7297-E 01 O.1922E05 7.10Q7E01 a 73so= pa 0. 8236E 01 0. 1 WOE 2 0.1979931 0.6436E-47 047-30E--08 0.87508E 05 0.1730E 22 0-0 0-083-8 047-4-G44418&00-01 446gr 01 0.1721S 01 Revision 11 November 1996 DCPP UNITS I & 2 FSAR UPDATE TARI E,45 5 44 Af'Tl~llTV DCICACCO C:QnhAA V'A flCClflkIDAI E -AI I II A I I I 6.0 .1-6 6 10 -I I. W ..---..-..---N 40 .N ý, 2-Hr-04 6 HrF 1 433 14-34$434GRG 1140RG l424PAR 1 P33PAR 1134PAR KPWM Xe 433 Xe 435M Xe 135M Xe 138 0. 2703r E 2 0. 3085e 02 O.6207E 02 a.7063rE0O

0. 5a74 12p 02 0,72406E02 0.163 gE 03 0.08477602-

-0.1141 4E 03-O.1041E03 0.14231 R 03 0. 9280E 03 0 2922r, 04 0. 3847E 01 0.70943rE(3 Q.7402E 04 0.26064E 04 0198-03 0.217-4303 0.32866 02 0-23689 0.747193E 0438646 04 9 0426.8-05 0.far a~E0 0.41376E01 0.6 552E al1 0-0 0-0 0-0 0-0 0.65 6 lE 03 Q-2326.E02

0. I08 03.or 1 0.2015Er.00 0.690rE03 0.257-2E,04 0.7-2736 02 0- 2388&"0 0.1281 OE 0. 202gre 05 0.46,0E, 06 0 1464E-06 0-0 0-0 9.9240S ,-1 a 5263r=-03 9.28-11-906 0-344SE 04338E 04 0.657.9E 02 0.2220E 02 0-3016E-02 O.12-52E027

0. 3227E 01 0.,8557Ec 10 9 5383r=-02 0. 26656 31 9490709 9 0.082-7r,04

0.4 413E02

0.1530E 10 042333r 06 9.25566E04 041014E 02 0-0G Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE T- Ir. 4 s & 45r%THYROID DOSE 2 HOUR CONTAINMENT LEFA.KA.GE EXPCTE CARE (REM)Distanoo Fromq Release Point, meterc NUG4de 80 424DG 2000 4000 7000 40000 200900 1131 0.7456F 03 0.41792E 03 0.2636FE03 0.1970 0.506GE04 034E-04 0.1247-E 4 112 .4895 04W29 0 Q754E 0 E06 (3.2074Eo06 a4823F.0 06 a73r-07 143 0.1627E 03 0 994146EO0 0.6389004 9.22466 04 GAGNE -04 04-888,08-0 0.2546 1 24 a226ar a a 01 ~ r as~880 0 6017E06 as 22r818-a01ig 04480 0 0482250 07. 93-94 7 0-9743E-" 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 01 1382ORG Q 1169rE05 047511E 606 0 4132rE06 0.1710Ea06 0.7032rE07 0,4662r50-9490569 07 k483QRG 0.549E G4 0.331gm0E0 0.182GE1 04.97573E 05 a,4.Q8- 05 0.2112rE 05 0,94 5E0 134G4RG 0.103iE096

0. 2501ErM06 0.11425E 06 0.9967 a.16.7 a477Er.071 0A4 4EnG 148QRG 0-7820E05 0-42fiG 0-2764E)5-0.1150E05 a 0.5207r,06-0.212&r,0A-0.1480 1434PAR 0000000000000 1~322A R 0.0004- 0 0400-1 133PAR 0-0 0- ---0 0- 0-0 1 424PAR 0-000.00000000-

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

~0 09G-4 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TAB LE 1545 17 THYROID DOSE30DY ONANMNTLAKG EX-PECTED CASE86 (REM)Digto;nn From Release Pzint, moetem NUG4de 800 4200 2000 40W0 7009 4000G=00"131T 0.74668m0 0 CEO=02E0336 9 0.14007E03 0.56G 0 0,3101E0 9 0.4247E 94 1 432 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 0254E0 143 -47E44- 0426894W3&06 G-1646&0&

9480E-00 0443*9 044Z b424ORr- 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 74r 1 Q30RG 9.919 .128GE 03 0.6:797E01

.7W 9 45E0 0.7-7346 05 .74m0 0 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 0-00 1 132PAR 0 40 40 40 1422PAR 94 040 0 4 40 1434PAR 040404040 04 0 1 135PAR 0 40 40 40 TOTAL ~ ~ ~ ~ ~ 04 0.479 .9E0 O8)6 .3266Em03 0 49,0 0.9196E 04.66E0 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 4 15& 10 0.HOLE B-ODY OSE 2 HOUR CONTAINOENT LR.KAGE EXPECTED CASE (REM)Mirtanne From Release Poin;t, moeters N~e #0042400 2000 400 90 400 20000-i1i9.04240 0-495m060 0.1a7-5EO 6 0.4471E607 0.206597 0.1265E0Q7 9.089C 080 1-422 .2- e9 .442 Q .G@=0 .34Q 0- 1543F4 07 00 0456,G Q304 R09 I i22 9.404# goe9 .63 9 40im0 -64626 07 0 243F 07 042F, 0 0.-400orn g 114 014F 6 044 6 024~m0 .7110 2ir 07~i 9.-667i G 0.3084 9 I-&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 O I 422ORZ GAM~E 07- 0.389g0E0 W O.24446E07 0.8921E05 44417rEOR 0.224 G .045E 9 1 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 0.482m9 baQQ 0.1330EO6 04OM-087 0-740-7 .40706407-0.98G 9 0.557O0 Q 224 FO I 434-PAR 04 4 4040 4 1-422PAR 04G 04G 004040 04 1 133PAR 040--0 04 -0 4 -1 434PAR -40 4000 -I 43SPAR- 0-0 0 0- 040-044 KF 3M .496re06 .4246 6 06774 E070.1i300gB 07 795.O 04204E 08 Kr-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 06 K 80.424i48E 0475-04 0-3704K04 0-144E-04 G0.m4412E0 Q-4359fi4 0.,175E Xe 433 9,4269903 0.07 4 046E 4 0 4962r,04 0.8555gB 5 9.23E05 0.2408r xe-433M 0.268r 06.a1709r.5 of 9936E 0.300GB 06 0.1804F06 OR .11i05E 96 0.4447 07 Xe 435 0,47-66104 0 32960 14 0.1694B4EOI G6 0 05 .23 5 48r 5 0Q.0 Xe 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. r TTL 0.3653993 0.24SE0 0.1291E 03 067r04 0.2470r.04 0 4540F. 041 .14E0 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE T.ABLE 15r5 Sr 20 (DELtETE!D)

T.A.LE. 155 21 VWHOLEr- 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 aACOF0O 1- .2-66 0.144626 06 a -8080840 9.344E 07- 0.443E 0 094 0- 8 0 20FO Q 4420~0r, 92996 0196 066E0- 083 0-0 941742E07 9.0-0GE 00 1 44 a4R2F.or, 04i 8F5 6 .65 E 0 9711rm 7- 0.1251E 07 C).7QA;0B ga -084E 08 1 13 0.1245E06 Q-7Q- --GE-& 024W 0 2291R07 0.1766E07w G4042&8 11310R0 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 8 k433ORG G.57-36906 0,35V11BPS 0.1865BpS 9.7649007 0 2449E,07-0,2422E 07 G.436m9 M4340G a.0299F.07 0.2566Bm07 0 414E1B07 0 5870B08O 0.27G~.08 BPS O 667 J 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 40 t-435PAR 0404 0 04- 0 0 KF 8M 04356rmG6 0223E-09424 06 0.515P1B07 0.2376E 07 046 7 G65 M K 50.448ge0-08 44A=0 0.5620rm 4 .206E 084 0480Q06F as 0 fi0-Fa )245O05 6 Kr E4 744R04 047Mr 0417 aB 027=1 20r 0 1744F-05 Q.1071BP W 5 042a06 0A87807 9F 4 .59E044 9.256994 90609 4 0,408405 0.297-G G 0446E05 Kf"0-44A4fiW 4 0.4W7qF-04 0.4603E-04 0.1612B01 9-9894E-G 0409699 6 Xe 438 0 14 222. 2 0.6011Bm03 9.440046E9 aI Q~.1an0B794F 04 9.36986044 G.43836 04 Xe 432A4 0 :1376E04 9 4224rm 04 a.6170rBPs a043850-0 0 .1072BP s 046329046 42578r- 06 Xe 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 4E9 TOTAL 9 203 E02 Q I SA-02 0 549F0-03 2@. 4~803 ~0.444 03 0.446G4 25-04 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-23

SUMMARY

OF OFFSITE, CONTROL ROOM & TECHNICAL SUPPORT CENTER DOSES LOSS OF COOLANT ACCIDENT 3 SUMMA, RY OF EXPOSURE FROM CONTAINMENT Dose (TEDE, rem)Regqulatory Limit (TEDE, rem)Maximum 2-hour Exclusion Area Boundary Dose 1 30-day Integrated Low Population Zone Dose 30-day Integrated Control Room Occupancy Dose 2 30-day Integrated TSC Occupancy Dose 2 5.6 25 25 1 3.7 (0.7)4.1 (1.3)5 5 Thyroid Doses, rem EAB 2 Hours, 1rQ Par 100 Design baSic case Expee~4ed ase 300 47-4-Whole Body Doses, rem EABR 2 Hours 1 0 GFR Part 100 Design basiS case Expeeted Gase 2-5 3 7 5-.64-" 2,5 6.44 W-OPUiaTnOn Uoce& Man Fem Design basis case EX198Gted Gase-932.4 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE Note: 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 pathway is considered a part of DCPP licensing basis with respect to passive system failure. If this assumed release pathway were not included, the maximum 2 hr dose at the EAB would occur between t=0.5 hrs to t=2.5 hrs (i.e., during the post-LOCA ex-vessel release phase and would be 3.4 rem.2. The dose presented represents the operator dose due to occupancy.

Value shown in parenthesis represents that portion of the total dose reported that is the contribution of direct shine from contained sources/external cloud.dose received by the operator during transit outside the control room is not a measure 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 dose contribution to the operator during routine access to control room for the duration of the LOCA is minimal. (a) The- e values correspond to the origina! analysis.

See Table 15.5 75 for current analysis 3. (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 beta Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE-rADI 4* r ')A 0" ---rlnoulý-

Itu.u &4 -- ". e- -.--ASSUMPTO.NS USED TO CALCULATE OFFSITE EXPOSURES FROM POST LOCA CIRCULTIONH LOOP LEAKAGE IN THE AUXILIA.RY BUILDING A. 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 system 6..AC*J'ity Dgepocited in Containment RAcirQW'-eation Sump Wat9r 1. ledine (Crae inventor' base on both U 235 9 PU 239 fi4ionS)Experted 2 a 2 ofgap in*.entwy, per Table 11.1 7;i (I 127, 120, rel.131, 132,133, fat. Table 1.1 100% ef gap 3.2g. , 0 v, 1-8.5g, 9 Gi 44g4824PQ 0 Gi Sxpest!Go% of gap in.ventor,'

per Table 11. 1 7;. (1127, 129, rel. ffa~t. At 0.01 5i 1 434 433-134, 135, mel. ffaet.Tab!l 11.1 7)100% of gap iedine 30.2g,O0 i 61.5g,O1.2x 2 a 4-2 2 4 10% of gp ainventer; per 1.25; (1 127,129, rel.-fraet. of 0.30; 11431, 1 32, 133, 131,135 Fl alOf0. 10)99475% of gap odkino 4,445g,O0 i 97,8AS4O5GI07 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 gap iodine inyenten;4,445g,O0 G a. wlemental 10d'no nveotrY (1) 1-127 (2) 112-9 (3) 1 131, 132, 133, 134, 135 Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATE FADI C 4-I r ')A4 L'L.--. 4 -CC.ASSIUMPTIONSi USED TO " AILCULITE OFFSITE EXPOSURES FROM POST LOCIAýW__ _ý I-.- -ft --M_ Ir, +MM 15wHWlPl"_

b. OrglaniG iedine (4) 1 27 (2) I12 3)I131, 132, 133,4134,135

2. Noble Gases Other fkissne products G. Conbiwnmct RecercuIafieR Sump DeCay and Gleanup 1. Radieolgisal decay s~edit 2. Cleanup e~edi D. VnlumA of OWeto in %hich AGtPty is Deposited (dil!uted)

2. Raccumuela~towateF, gal.Fmpeeted sma_9 0% ef- gap Win Nlene 93,960 25,040 Expte 189, 48*0 QMA 9.25% of gap iodine-inyentor 2g, 0 11g, 0 Ci 02gA,492.0-0 DBA 0.25g; of gap idn 29, 0 Qi ll1g, 9 i 929, 42.G0 5.68,-0 7igG yes Nle~e Yes.NORe yoes 03,960 03,060 Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 1.IS5 24 ho 1 of 5 ASSUMPTIONS USED TO CALCUL.A.TE OFFSITE EXPOSURES FROM. POST LOCA.GIR1ULATION LOOP LEAK.GE IN11 TH AUXILIARY BUILDING kxpec4ed Xetd0Q8 D. Volumo of W.^tor in" Whi.',!a Activity ic Deposited (diluted) (Contd)3. Refueling V~teF etoroge tank, gal. 350,000 262,939 35 0 , 0 0 0 25,0 (Table 6.344 1. Total, gal. 4690000 2&1-T030 469,000 373,220& Conditiono of Locp Leakage Water 1. pH of loakage water 8-8 84 87 (Figuro 6.2 15)2. Tempr.tur.

of leakage "ate°, 'r 420 238 420 242 F. Loop Leakage Rate 19044 WhF q0-M 14Q GGA'* 50 gPm-8T~ .3 0) (Table 6.3 0)G. Duration of LGop Loakago I. 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, h 3. Total~n duration of loop 74-970 7.40, 0 leakage~h Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATE T.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 bdine Derantaminatien Fa-tres 1Elomontol iodinedcotoioto M liquid, ibm -b. An x I 0, V liquid, fas-c. PDoRtiion ooeficint, ,G-, 4 -4,0 d. P g."t) gcn 48x0 -PF, ((g) 2 .DQq cnt*minotocn foe'ctor, 4-. 4- 4 4-4 G--" OF, (a)fGWl, Dk (9) gas I. Auxiliar; Building Dr'c..Plateewt, and Filtcr Remonva 1.

decoy eGcdit NeRe NeRe Nene Neae-2. Plat"ot credit Nene NeAe Nee NeOre Revision 12 September 1998 DCPP UNITS 1 & 2FSAR UPDATE TAB3LE 15.5 24 ho1o 5.ASSUM1PTIONIS U'SED TO CAI ll ATE OFFRS!TE EXPOSURES FROM4 POST LOC A CIRCULATION LOOP LESAXA.GE IN THE --AU[XIL'A.RY BUILDING-Expected Expesede OBA B I.Auxii~ar; Ruilding De~ay, Plateout, and Filter Rcmeval (Contd)3.Auxii~ary bUilding Nos;e YsNon~eYe A. AzdOA filtpr AffigiznGy (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 Nene derzay Gcedi 2. AtmoephewiG dilutien Table 15.54 Tahle 4 5 64 Table-4&&-

4 Tabie 45. 4 K. Brzthing RatesTbe .Table 166 7 Table 4 5A Table 45.5 7 Revision 12 September 1998 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 26 PERCENTAGE OCCURRENCE OF WiND DIRECTION AND CGALM WINDS EXPRESSED AS PERCENTAGE OF TOTAL HOURLY OBSERVATIONS WITHIN EACH SEASON AT THE SITE (250 FOOT LEVElL)Offshoefe{b+

0hOFsOre-ý Gak:n-(,*Annual wet Transkoa 57-0 5"0 5"0'6"0 3"%4"0 42%"0&4%40k (a) Dry Season May through September Wet Sea6Gon November through March Transitional 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.hwest mneasured 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 p II I ~II II I I ~ I~JIll 0-U C z III III I +/-4 11111 C')N)m I C,)I III II I I~P? -U H m II I I III 1111 +1 CD II ~a C,)I CD CO IIIIIIIIIIIIIICOOVOC 0 CO LA I. 1.1k 0) +4 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 27 Sheet 2 of 2 CRN NNW 14 _a_A D) W 4 , A 0 w T A 0 4 0406 94086 0 457 0406 2 0-094 03473 04 42 ano9n 3 90-048 0-064 0437 0,056 4 03064 a0EM 0-44 0-049 6 0-066 0-040 0-08G 0 086 6 a-06a G-0461 0-Ogg 0-066 7 0-450 0.046 0-049 0-066 6 13044 0-03 0-046 0-072 9 0-045 0-0144 a-029 030" 40 Q 036 0-044 0-002 0-049 44 0-046 0-060 DO0 0-054 42 0425 0028 0349 0-039 43 0336 04464 0-041 90444 44 0036 0-043 0044 0345 46 0349 Q0270 0340 0-043 46 0-020 .0325 0-000 0326 44 0.322 00334 03000 00144 489 0-023 0-0133 0-046 0044 49 cl 030 0034 0346 0-046 20 0-043 owl4 -03009 24 0-003 0300 -03000 22 03047 0906 -. 03046 23 04004 03006 -cap0 240042 0042 -00 26 a0342 0.049 --26 0-042 0320 -24 0004 0304 -28 0-004 0304 -29 0.0 0-00 -30 0300 90300 9 34 0.006 0-008 9 (a) A~ -RW 0 -D~ caca (May-, thoaugh £frpt~rbe-)

T TF-M:tGoa M41z(p'z-1Otbr 364 0463 a3 0349 0494 0404 0=4 0448 042 G 0246 0-46 044 00369 0360400 0-042 0364334024033 0746 -0023 0024 03ON -OGG6 0.02i 030 -0 3.02 0407 --3i0.039 04 Q 3 004 0704 04 GA442 0302 0-424 0-247 0.224 03g 8034- 07 0444 0404 03 0422 00-44 03.= 03022 0449 0-344 30 -0.00 --04i --A 0- W 4 03944 a3900236 0426 0233 0.-24 0339 03-244 044r2041660154 0 404 0-40a 303al0494 0354 0380.374 0460466 a0Og46l0320300 0344 00434 60260 30 -0.00000QOg0000

-9033 002340334

-Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 2 8 ASSUMPTIONS USED TO CALCULATE ONSHORE CONTROLLED CONTTAI.MENT VENTING Sheet 1 of 2 DBA-Gase A. Aztovily Released to Contaiwnment Atrnespherc

-4.-4edi~e a....... nta1 c. Particulate

2. Noblol garses 3. Other fi 6ion prod-ctc B3. Decay, Cleanup, and Leakage in Contaiwnmnt-A#Resphewe

1. Radiologieal decay cr.edit 2. spray cleanup a. 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 gapinntr yes 0 Ne~e NoeR 0.05%ipot day, 25%9 o-f core iod-ine inventeor 22.75%A of Gaero iodine inVentRI 1.0% of core icdine in:entor;1.25% of core iedinc inventer;100% of core inventor;Neso Ne6e 0.05%/per day a. lodine6 h- Noble gases 4. Contafinment leak Fate Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 28 Sheet 2 of 2 ASUPIO!NS I~ ISE TO ALCULAT\ wlwwý rr~~-cE UNT-o i FOL COUNTAI NMENT A GALrJITING~

2&P4O%~b4cee DB.A.Case A G. Containment Atmcsphc."c Volume D. Purge Schedule 1. Time a#Fr LOC.A purging bogine 2. Time after LOCA purging ends 1968 hou., Chapter 6 6782 Mum, remainder of 1 Yrt.10 Q 4m, Chapter 6 67-2 Chapter6 8088 houm, e.. aindc. of 1 yr.25 cf, .hapter 65 E. Purge Fleorate F. Filter Efficency!. -4edMnes a. Elemental b. Or~ganic c. Particulate

2. Noble gases G. Atmesphenc Dispemrsi,-

1. Radioelgical decay credit 2. %iQc W. Breathing Rates (a) Although aIt , cuben..........

2 hr.... e...466.insigniflcant (Referenco 39).Q"%9"%4"0e Q"%7"0 We"e NeRe Table 15. 0 WeRe Table 46.6~valuafien shcv.ed t~hat the Design Cace eoeffid9Mn Of 34 (0Fo 2600 gpm spray header POeW) should be reduced to gpmA 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 UPDATE TABLE 15.5 29 ONSHORE CREGONTROLLED F QC GONTAI NIEA4FeNIT NlVE NTIT-l N'G EXPOSU WRE-S Thyroid eXpo..re at site 241-boundar; (800 m.eters), rem" Whole~ body exposur 0.0841-7.&*G at site boundary (800 m~eters), rem Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 30 ATM1OSHPERIC DISPERSION FACTORS FOR ONSHORE CONTROLLED CONTlAINlMENT VENTING (STABILITY CATEGORY D)QistaR~ee-m

%Zf37 .x I e--4 14 2.40-x-4-4 4- 77.884 wx40 4 7.0 3.135 x 0--4Q-0

-2Q7 6.099 *40 4 MeteGOrglogcal Input ParamneterS:

Height of releasc -70 mneterc Mixing depth 5-0m~eters ManR; wind speed -5.83 meters per second Sigma theta -10 degrees Sigma phi -3 degreeS Vertic-al expansion rate beta, 0, -0.9 Azimuth eXpansion rate alpha, et, -0.9 Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-23A LOSS OF COOLANT ACCIDENT Assumptions

& Key Parameter Values Parameter Value Core Power Level 3580 MWt (105% of the rated power of 3411 MWth)Fuel Activity Release Fractions Per Reg. Guide 1.183 (See Section 15.5.17.2.2.2)

Fuel Release Timing (gap) Onset: 30 sec Duration:

0.5 hr Fuel Release Timing (Early-In-Vessel)

Onset: 0.5 hr Duration:

1.3 hr Core Activity Table 15.5-77 Chemical Form of Iodine released from fuel to 4.85% elemental containment atmosphere 95% particulate 0.15% organic Chemical Form of Iodine Released from RCS 97% elemental and sump water 3% organic Containment Vacuum/Pressure Relief Parameters Minimum Containment Free Volume: 2.550E+06 ft 3 Primary Coolant Tech Spec Activity Table 15.5-78 Chemical Form of Iodine Released 97% elemental; 3% organic Maximum RCS flash fraction after LOCA Noble Gases 100%Halogens 40%Maximum containment pressure relief line air 218 actual cubic feet per second flow rate (acfs)Maximum duration of release via containment 13 sec pressure relief line Release Point Plant Vent Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE Containment Leakace Parameters Containment Spray Coverage -Injection Spray 82.5% (sprayed fraction)and Recirculation Spray Modes: Sprayed Volume 2.103E+06 ft 3 Unsprayed Volume 4.470E+05 ft 3 Minimum mixing flow rate from unsprayed to sprayed 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 sec End = 30 days Containment spray in injection mode Initiation time 111 sec Termination time 3798 sec Maximum delay between end of injection 12 min (based on manual operator spray and initiation of recirculation spray action)Containment spray in recirculation mode Initiation time 4518 sec Termination time 22,518 sec Long-term Sump Water pH > 7.5 Maximum allowable DF for fission product Elemental Iodine: 200 removal Others: not applicable Elemental iodine and particulate spray removal See Table 6.2-32 coefficients in sprayed region during both injection spray and recirculation spray modes Elemental iodine removal coefficients due to See Table 6.2-32 wall deposition Particulate removal coefficients in unsprayed See Table 6.2-32 region due to gravitational settling Containment Leak rate (0-24 hr) 0.1% weight fraction per day Containment Leak rate (1-30 day) 0.05% weight fraction per day Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE Containment Leakage Release Point (Unfiltered)

From the worst case release point of the following:

Diffuse source via the containment wall Via Plant Vent Via Containment Pen Area GE Via Containment Pen Areas GW& FW ESF System Environmental Leakacie Parameters Minimum post-LOCA containment water 480,015 gal.volume sources Minimum time after LOCA when recirculation is 829 sec initiated Duration of leakage 30 days Maximum ECCS fluid temperature after 259.9 OF initiation of recirculation Maximum ECCS leak rate (including safety Unfiltered via plant vent = 240 factor of 2) cc/min Unfiltered via Containment Penetration Areas GE or GW & FW= 12 cc/min RHR pump seal failure Filtered(')

via plant vent 50 gpm starting at t = 24 hrs for 30 min Iodine 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 sec Maximum ECCS / sump water back-leakage 2 gpm rate to RWST (includes safety factor of 2)RWST back-leakage iodine release fractions See Table 15.5-23C RWST back-leakage noble gas, as iodine See Table 15.5-23C daughters, release rate from the RWST vent Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE Miscellaneous Equipment Drain Tank (MEDT) Leakage Parameters MEDT inflow rate (includes safety factor of 2) 1900 cc/min MEDT leakage Iodine release fractions See Table 15.5-23D MEDT leakage noble gas, as iodine daughters See Table 15.5-23D release rate from plant vent CR Emernency Ventilation:

Initiation Signal/Timing Initiation time (signal) SI signal generated:

6 sec Non-Affected Unit NOP Intake Isolated:

18 sec Affected Unit NOP Intake Isolated and CRVS Mode 4 in full Operation:

44.2 sec Bounding Control Room Atmospheric Table 15.5-23B Dispersion Factors for LOCA Note: Releases from the RHR Pump Seal failure are filtered for CR dose evaluation and Site Boundary Dose Evaluation.

Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE 15.5-23B LOSS OF COOLANT ACCIDENT Control Room Limiting Atmospheric Dispersion Factors (sec/m 3)Release Location / Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 6-720 I hr Control Room Normal Intakes Plant Vent Release-Affected Unit Intake 1.67E-03 -- ---Non-Affected Unit Intake 9.1OE-04 Containment Penetration Areas-Affected Unit Intake 6.84E-03-Non-Affected Unit Intake 2.24E-03 Control Room Infiltration Plant Vent 1.26E-03 8.96E-04 3.44E-04 3.44E-04 2.99E-04 Containment Penetration Areas 3.22E-03 1.85E-03 7.29E-04 7.15E-04 6.64E-04 RWST Vent 1.07E-03 5.80E-04 2.18E-04 2.19E-04 1.79E-04 Control Room Pressurization Intake Plant Vent 5.65E-05 3.70E-05 1.35E-05 1.37E-05 1.11E-05 Containment Penetration Areas 6.45E-05 4.05E-05 1.65E-05 1.38E-05 1.12E-05 RWST Vent 5.25E-05 3.03E-05 1.15E-05 1.10E-05 8.83E-06 Note 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 Area Note 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 Release Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-23C LOSS OF COOLANT ACCIDENT RWST Iodine Releases Fraction and Gas Venting Rate to Atmosphere Average Interval Iodine Release Fraction Weighted Gas Space to Atmosphere Venting Rate to Atmosphere Sec Sec Fraction Ireleased I lentering Fraction Vrst / day 829 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 2592000 1.483E-06 1.702E-02 Where: Ireleased

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

= Total Iodine mass entering to the RWST during specified time interval, gm Frac. Vrt = Rate of Fractional RWST gas volume vented during specified time interval Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-23D LOSS OF COOLANT ACCIDENT MEDT Iodine Release Fraction and Gaseous Venting Rate to Atmosphere Iodine Release Average Interval Weighted From Time To Time Fraction to Gas Space Venting Rate Atmosphere to Atmosphere Sec Sec Fraction 'released

/ Fraction VMEDT / day lentering 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, gm lentedng = Total Iodine mass entering to the MEDT during specified time interval, gm Frac VMEDT = Rate of Fractional MEDT gas volume vented during specified time interval Note 1: After the MEDT overflows at t = 183,289 sec, the gas venting rates are 2640 cfm from the EDRT room, and 1760 cfm from the U1/U2 Pipe Tunnels (i.e., the exhaust ventilation rate from 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 the tank and spilled liquid after overflow occurs, should also be released instantaneously to the environment without hold-up.Note 2: The room ventilation flows addressed in Note 1 (utilized as clean in-coming air) are incorporated into the determination of the iodine equilibrium concentration in the EDRT room and 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 UPDATE TABLE 15.5-23E LOSS OF COOLANT ACCIDENT TSC Limiting Atmospheric Dispersion Factors (sec/mn 3)Release Location / Receptor 0-2 hr 2-8 hr 8-24 hr 24-96 hr 96-20 hr TSC Normal Intakes Plant Vent Release 5.52E-04 ----- ----- ----- -----Containment Penetration Areas 1.80E-03 RWST Vent 3.63E-04 ----- ----- ----- -----TSC Infiltration Plant Vent 5.43E-04 2.16E-04 9.97E-05 8.11 E-05 6.58E-05 Containment Penetration 1.83E-03 7.49E-04 3.16E-04 2.92E-04 2.41 E-04 Areas RWST Vent 3.72E-04 1.68E-04 6.64E-05 6.17E-05 5.1OE-05 CR/TSC Pressurization Intake Plant Vent 3.70E-05 1.35E-05 1.37E-05 1.11E-05 Containment Penetration Areas 4.05E-05 1.65E-05 1.38E-05 1.12E-05 RWST Vent 3.03E-05 1.15E-05 1.10E-05 8.83E-06 Note 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 Area Note 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 Release Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 31 CONTROL ROOM IN.4 F -,^TRATION A ,SUMAE D FOR RADI _,LGICAL EXPOSURE -ALl,, ,TION'S Sheet 1 of 2 L~eakaqe Pa;th.PWiedows R. Dnf s G. Pcnctratioflns

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

No lcakago: conG.erte wallS and floor poured with piping i n placae and cenRal-- rooem -he b positively pprnssurli-ze-d-.

No space between exposed conduto.s and trays sealed with RW R-01OWOOL ceramic fiber 6 inchos in depth, Yvth two coats of FLAMEMP.STIC 72A, and control roomR Will be poSitively pressr.Uized-No leakagwe:

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

O-.G 0-0 3. Conduits and trays a. EwtemFalcreal

b. Wnernal SeaRl Revision 11 November 1996 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5 31 CONTROL ROOMK INF!LTP.A.TIO.N ASSUMRED FOR P..,DIOLOG!CAL EXPOSURE CALCUL.ATIONS Sheet 2 of 2 D. Dampers where;Q -leakage, Gfm A -damper arca, square feet q- leakage per unit damnper area per in. of wateor" Ap -peresure difference aGrOcS damper i. F waor A -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#2 2. Mode damper #3 3. Mode damper#7 4, Mode damper #8 Ei.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