ML15176A531

Diablo Canyon Units 1 and 2 - License Amendment Request 15-03 Application of Alternative Source Term - Updated Final Safety Analysis Report Markup. Part 5 of 8

ML15176A531
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: ML15176A531 (51)

Text

DCPP UNITS 1 & 2 FSAR UPDATE15.5.8.8 ACTIVITY2ACTIVITY2 (Reference 65) calculates the concentration of fission products in the fuel,coolant, waste gas decay tanks, ion exchangers, miscellaneous tanks, and release linesto the atmosphere for a pressurized water reactor system. The program uses a libraryof properties of more than 100 significant fission products and may be modified toinclude as many as 200 nuclides. The program output presents the activity and energyspectrum at the selected part of the system for any specified operating timeACTIVITY2 is used to develop the reactor coolant activity inventory (design and aslimited by the plant Technical Specifications) utilized to assess the design basisaccidents excluding the tank ruptures.15.5.8.9 IONEXCHANGERION EXCHANGER (Reference 66) calculates the activity of nuclides in an ionexchanger or tank of a nuclear reactor plant by solving the appropriate growth-decay-purification equations. Based on a known feed rate of primary coolant or other fluid withknown radionuclide activities, it calculates the activity of each nuclide and its products inthe ion exchanger or tank at some later time. The program also calculates the specificgamma activity for each of the seven fixed energy groups.IONEXCHANGER is used to develop the secondary coolant activity inventory (designand as limited by the plant Technical Specifications) utilized to assess the design basisaccidents excluding the tank ruptures.15.5.8.10 EN 113, Atmospheric Dispersion FactorsEN-1 13 Atmospheric Dispersion Factors (Reference 73) calculates x/Q values at theEAB and LPZ following the methodology and logic outlined in Regulatory Guide 1.145,Revision 1. The program can handle single or multiple release points for a specifiedtime period and set of site-specific and plant-specific parameters. A release point canbe identified as either of two types of release (i.e., ground or elevated), time periods forwhich sliding averages are calculated (i.e., 1 to 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> and/or annual average),applicable short-term building wake effect, meandering plume, long-term building heightwake effect, and a wind speed value to be assigned to calm conditions. Downwinddistances can be assigned for each of the sixteen 22.5-degree sectors for two irregularboundaries and for ten additional concentric boundaries used only in the annualaverage calculation. EN-1 13 performs the same calculations as the NRC PAVAN codeexcept that EN-1 13 calculates X/Q values for the various averaging periods directlyusing hourly meteorological data whereas PAVAN uses a joint frequency distribution ofwind speed, wind direction, and stability class.EN-1 13 is used to develop the DCPP site boundary atmospheric dispersion factorsutilized to assess the design basis accidents excluding the tank ruptures.15.5-31Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.8.11 ARCON96ARCON96 (Reference 74) was developed by Pacific Northwest National Laboratory(PNNL) for the NRC to calculate relative concentrations in plumes from nuclear powerplants at control room air intakes in the vicinity of the release point. ARCON96 has theability to evaluate ground-level, vent, and elevated stack releases; it implements astraight-line Gaussian dispersion model with dispersion coefficients that are modified toaccount for low wind meander and building wake effects. The methodology is also ableto evaluate diffuse and area source releases using the virtual point source technique,wherein initial values of the dispersion coefficients are assigned based on the size ofthe diffuse or area source. Hourly, normalized concentrations (x/Q) are calculated fromhourly meteorological data. The hourly values are averaged to form X/Qs for periodsranging from 2 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> in duration. The calculated values for each period are usedto form cumulative frequency distributions.ARCON96 is used to develop the control room and TSC atmospheric dispersion factorsutilized to assess the design basis accidents excluding the tank ruptures.15.5.8.12 SWNAUASWNAUA (Reference 67) is a derivative of industry computer code NAUA/Mod 4 whichwas originally developed in Germany and was based on experimental data. NAUA/Mod4 addressed particulate aerosol transport and removal following a LOCA at an LWR. Itdeveloped removal coefficients to address physical phenomena such as gravitationalsettling (also called gravitational sedimentation), diffusion, particle growth due toagglomeration, etc using time-dependent airborne aerosol mass. NAUA4 (included inthe NRC Source Term Code Package) was used by NRC during the initial evaluationsof post-TMI data. NAUA/Mod 4 was modified to include spray removal anddiffusiophoretic effects suitable for design basis accident analyses. A version ofSWNAUA (SWNAUA-HYGRO) was proven to be the most reliable of more than a dozeninternational entries, in making predictions of aerosol removal for the LWR AerosolContainment Experiments (LACE) series.SWNAUA is used to develop the time dependent post LOCA particulate aerosolremoval coefficients in the sprayed and unsprayed regions of containment.15.5.8.13 RADTRAD 3.03RADTRAD 3.03 (Reference 68) is a NRC sponsored program, developed by SandiaNational Labs (SNL). It can be used to calculate radiological doses to the public, plantoperators and emergency personnel due to environmental releases that resulting frompostulated design basis accidents at light water reactor (LWR) power plants. TheRADTRAD 3.03 (GUI Interface Mode) includes models for a variety of processes thatcan attenuate and/or transport radionuclides. It can model sprays and natural depositionthat reduce the quantity of radionuclides suspended in the containment or other15.5-32Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEcompartments. It can model the flow of radionuclides between compartments within abuilding, from buildings into the environment, and from the environment into a controlroom). These flows can be through filters, piping, or simply due to air leakage.RADTRAD 3.03 can also model radioactive decay and in-growth of daughters.Ultimately the program calculates the Thyroid and TEDE dose (rem) to the publiclocated offsite and to onsite personnel located in the control room due to inhalation andsubmersion in airborne radioactivity based on user specified, fuel inventory, nucleardata, dispersion coefficients, and dose conversion factors.RADTRAD is used to develop the TEDE dose to the public located offsite and to onsitepersonnel located in the control room due to inhalation and submersion in airborneradioactivity following design basis accidents excluding tank ruptures15.5.8.14 PERC2PERC2 (Reference 69) is a multi-region activity transport and radiological doseconsequence program. It includes the following major features:(1) Provision of time-dependent releases from the reactor coolant system to thecontainment atmosphere.(2) Provision for airborne radionuclides for both TID and AST releaseassumptions, including daughter in growth.(3) Provision for calculating the CEDE to individual organs as well as EDE frominhalation, DDE and beta from submersion, and TEDE.(4) Provisions for tracking time-dependent inventories of all radionuclides in allcontrol regions of the plant model.(5) Provision for calculating instantaneous and integrated gamma radiationsource strengths as well as activities for the inventoried radionuclides topermit direct assessment of the dose from contained / or external sourcesfor equipment qualification, vital area access and control room and EABdirect shine dose estimates.PERC2 is used to calculate the accident energy release rates and integrated gammaenergy releases versus time for the various post-LOCA external and contained radiationsources. This source term information is input into SWQADCGGP to develop thedirect shine dose to the control room. PERC2 is also used to develop the decay heat inthe RWST and MEDT and develop the TEDE dose to personnel located in the TSC dueto inhalation and submersion in airborne radioactivity following LOCA.15.5-33Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.8.15 SW-QADCGGPSW-QADCGGP (Reference 70) is a variant of the QAD point kernel shielding programoriginally written at the Los Alamos Scientific Laboratory by R. E. Malenfant. TheQADCGGP version implements combinatorial geometry and the geometric progressionbuild-up factor algorithm. The SW-QADCGGP implements a graphical indication of thestatus of the computation process.SW-QADCGGP is used to develop the direct shine dose to the operator in the controlroom, TSC and EAB.15.5.8.16 GOTHICGOTHIC (Reference 71) is developed and maintained by Numerical ApplicationsIncorporated (NAI) and an integrated, general purpose thermal-hydraulics softwarepackage for design, licensing, safety and operating analysis of nuclear power plantcontainments and other confinement buildings. GOTHIC solves the conservationequations for mass, momentum and energy for multicomponent, multi-phase flow inlumped parameter and/or multi-dimensional geometries. The phase balance equationsare coupled by mechanistic models for interface mass, energy and momentum transferthat cover the entire flow regime from bubbly flow to film/drop flow, as well as singlephase flows. The interface models allow for the possibility of thermal non equilibriumbetween phases and unequal phase velocities, including countercurrent flow. Otherphenomena include models for commonly available safety equipment, heat transfer tostructures, hydrogen burn and isotope transport.GOTHIC is used to estimate the containment and sump pressure and temperatureresponse with recirculation spray, the temperature transient in the RWST / MEDT gasand liquid due to incoming sump water leakage / inflow / decay heat from the RWST /MEDT fission product inventory, and the volumetric release fraction transient from theRWST / MEDT gas space to the environment.15.5.9 CONTROL ROOM DESIGN AND TRANSPORT MODELThe control room serves both units and is located at El 140' of the Auxiliary Building.The walls facing the Unit 1 and Unit 2 containments (i.e., the north and south walls) aremade of 3'-0" concrete, whereas as the control room east and west walls are made upof 2'-0" concrete. The floor and ceiling thickness / material reflect a minimum of 2'-0"and 3'-4" of concrete, respectively. The control room Mechanical Equipment and HVACroom is located adjacent to the control room (east side), at El 154'-6".The control room has a normal intake per unit (each located on opposite sides theauxiliary building; i.e. north and south), and a pressurization flow intake per unit (eachlocated on either side of the turbine building; i.e. north and south). The control room15.5-34Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEpressurization air intakes have dual ventilation outside air intake design as defined byRegulatory Position C.3.3.2 of Regulatory Guide 1.194, June 2003 (refer to Section2.3.5.2.2)During normal operation (CRVS Mode 1), both control room normal intakes areoperational. Redundant PG&E Design Class I radiation monitors located at each controlroom normal intake have the capability of isolating the control room normal intakes ondetection of high radiation and switching the control room ventilation system (CRVS) toMode 4 operation (i.e., control room filtered intake and pressurization).CRVS Mode 4 operation utilizes redundant PG&E Design Class I radiation monitorslocated at each control room pressurization air intake and the provisions of acceptablecontrol logic to automatically select the least contaminated inlet at the beginning of theaccident, and manually select the least contaminated inlet during the course of theaccident in accordance with Regulatory Guide 1.194, June 2003. Thus, during Mode 4operation the dose consequence analyses can utilize the X/Q values for the morefavorable pressurization air intake reduced by a factor of 4 to credit the "dual intake"design (refer to Section 2.3.5.2.2).Other signals that initiate CRVS Mode 4 operation include the safety injection signal(SIS) and Containment Isolation Phase A. The SIS does not directly initiate CRVSMode 4, however, it initiates Containment Isolation Phase A which initiates Mode 4operation.During normal operations, unfiltered air is drawn into the control room envelope (refer toTable 15.5-81) from the Unit 1 and Unit 2 normal intakes. In response to a control roomradiation monitor or SIS, the control room switches to CRVS Mode 4 operation, andcontrol logic ensures that the CRVS pressurization fan of the non-accident unit isinitiated and air is taken from the less contaminated of the Unit 1 or Unit 2 control roompressurization air intakes. The control room pressurization flowrate used in the doseconsequence analyses is selected to maximize the estimated dose in the control room.With the exception of 100 cfm which is unfiltered due to backdraft damper leakage, allpressurization flow is filtered.The allowable methyl iodide penetration and filter bypass for the CRVS Mode 4Charcoal Filter is controlled by Technical Specifications and the VFTP, and is 2.5% and<1%, respectively. In accordance with Generic Letter 99-02, June 1999 a safety factorof 2 is used in determining the charcoal filter efficiency for use in safety analyses (referto Section 9.4.1 and Table 9.4-2. Thus the control room charcoal filter efficiency forelemental and organic iodine used in the DCPP safety analyses is 100% -[(2.5% + 1%)x 2] = 93%. The acceptance criteria for the in-place test of the high efficiencyparticulate air (HEPA) filters in Technical Specifications is a "penetration plus systembypass" < 1.0%. Similar to the charcoal filters, the HEPA filter efficiency for particulatesused in the DCPP safety analyses is 100% -[(1%) x 2] = 98%.During Mode 4 operation, the control room air is also recirculated and a portion of the15.5-35Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATErecirculation flow filtered through the same filtration unit as the pressurization flow.Refer to Table 15.5-81 for a summary of recirculation flow rates.Unfiltered inleakage into the control room during Mode 1 and Mode 4 is provided inTable 15.5-86 and includes ingress/egress inleakage based on the guidance provided inSRP 6.4.For purposes of estimating the post-accident dose consequences, the control room ismodeled as a single region. When in CRVS Mode 4, the Mode 1 intakes are isolatedand outside air is a) drawn into the control room through the filtered emergencyintakes; b) enters the control room as infiltration, c) enters the control room duringoperator egress/ingress, and d) enters the control room as unfiltered leakage via theemergency intake back draft dampers. The direction of flow uncertainty on the CRVSventilation intake flowrates (normal as well as accident), are selected to maximize thedose consequence in the control room.The dose consequence analyses for the LOCA, MSLB, SGTR and the CREA, assumea LOOP concurrent with reactor trip.In addition, and as noted in Section 15.5.1.2, in accordance with current licensingbasis the non-accident unit is assumed unaffected by the LOOP. Thus, to address theeffect of a LOOP, and taking into consideration the fact that the time of receipt of thesignal to switchover from CRVS Mode 1 to Mode 4 is accident specific:a. Automatic isolation of the control room normal intake of the "non-accident" unit,is delayed by 12 seconds from receipt of the signal, to switch to CRVS Mode 4.This delay takes into account a 2 second SIS processing time and a 10 seconddamper closure time.b. Automatic isolation of the control room normal intake of the accident unit, andcredit for CRVS Mode 4 operation is delayed by 38.2 seconds from receipt ofthe signal to switch to CRVS Mode 4. This delay takes into account a) 28.2seconds for the diesel generator to become fully operational includingsequencing delays, and b) 10 seconds for the control room ventilation dampersto re-align. The 2 second SIS processing time occurs in parallel with dieselgenerator sequencing and is therefore not included as part of the delay. Inaddition, and as discussed earlier, the CRVS system design ensures that uponreceipt of a signal to switch to Mode 4, the control room pressurization fans ofthe non-accident unit is initiated; thus fan ramp-up is assumed to occur wellwithin the 38.2 seconds delay discussed above, unhampered by a LOOP.The dose consequence analyses for the LRA and the LOL event assume that thecontrol room remains in normal operation mode and do not credit CRVS Mode 4operation.Table 15.5-81 lists key assumptions / parameters associated with control room design.15.5-36Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe i;nfnrmaton in thi;- 1ec-tion has been moved to Sect*on 155.9.4-.15.5.10 RADIOLOGICAL CONSEQUENCES OF CONDITION II FAULTS15.5.10.1 Acceptance CriteriaThe radiological consequences of accidents analyzed in Section 15.2 (or from otherevents involving insignificant core damage, but requiring atmospheric steam releases)shall not exceed the dose limits of 10 CFR 10Q._-1-50.67, and will meet the doseacceptance criteria of Regulatory Guide 1.183, July 2000 as outlined below:EAB and LPZ Dose CriteriaRegulatory Guide 1.183 does not specifically address Condition II scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.(1) An individual located at any point on the boundary of the exclusion area for the-two hours immcdiately any 2-hour period following the onset of the postulatedfission product release shall not receive a total-radiation dose to the whole bodyin xces o 25 rem or a total Fadoatien dose in excess Of 300 rem to the thyroidfrom iodine exp,.u.ein excess of 0.025 Sv (2.5 rem) TEDE.(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose to the whole body in excess of 25 rem, or a totalradiation dose in eXcess Of 300 remn to the thyroid from iodine expoure. inexcess of 0.025 Sv (2.5 rem) TEDE.Control Room Dose Criteria(3) Adequate radiation protection is provided to permit access and occupancy ofthe control room under accident conditions without personnel receivingradiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of theaccident.15.5-37Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.10.2 Identification of Causes and Accident Description15.5.10.2.1 Activity Release PathwaysAs reported in Section 15.2, Condition II faults are not expected to cause breach of anyof the fission product barriers, thus preventing fission product release from the core orplant. Under some conditions, however, small amounts of radioactive isotopes could bereleased to the atmosphere following Condition II events as a result of atmosphericsteam dumps required for plant cooldown. The particular Condition II events that areexpected to result in some atmospheric steam release are:(1) Loss of electrical load and/or turbine trip(2) Loss of normal feedwater(3) Loss of offsite power to the station auxiliaries(4) Accidental depressurization of the main steam systemThe amount of steam released following these events depends on the time relief valvesremain open and the availability of condenser bypass cooling capacity.The mass of environmental steam releases for the Loss of Load Event bound allCondition II events.A LOL event is different from the Loss of Alternating Current (AC) power condition, inthat offsite AC power remains available to support station auxiliaries (e.g., reactorcoolant pumps). The Loss of AC power condition results in the condenser beingunavailable and reactor cooldown being achieved using steam releases from the SGMSSVs and 10% ADVs until initiation of shutdown cooling.In-keeping with the concept of developing steam releases that bound all Condition IIevents and encompass the LRA and CREA, the analysis performed to determine themass of steam released following a LOL event incorporates the assumption of Loss ofoffsite power to the station auxiliaries.Although Regulatory Guide 1.183 does not provide specific guidance with respect toscenarios to be assumed to determine radiological dose consequences from ConditionII events, the scenario outlined below for the LOL analysis is based on the conservativeassumptions outlined in Regulatory Guide 1.183 for the MSLB, and was analyzed tobound all Condition II events that result in environmental releases.Table 15.5-9A lists the key assumptions / parameters utilized to develop theradiological consequences following a LOL event. The conservative assumptionsutilized to assess the dose consequences ensure that it represents the LimitingCondition II event.15.5-38Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEComputer code RADTRAD 3.03, is used to calculate the control room and siteboundary dose due to airborne radioactivity releases following a LOL event.15.5.10.2.2 Activity Release Transport ModelNo melt or clad breach is postulated for the LOL (refer to Section 15.2.7). Thus, andin accordance with Regulatory Guide 1.183, Appendix E, item 2, the activity releasedis based on the maximum coolant activity allowed by the plant TechnicalSpecifications, which focus on the noble gases and iodines. In accordance withRegulatory Guide 1.183, two scenarios are addressed, i.e., a) a pre-accident iodinespike and b) an accident-initiated iodine spike.a. Pre-accident Iodine Spike -the initial primary coolant iodine activity isassumed to be 60 gCi/gm of DE 1-131 which is the transient TechnicalSpecification limit for full power operation. The initial primary coolant noblegas activity is assumed to be at Technical Specification levels.b. Accident-Initiated Iodine Spike -the initial primary coolant iodine activity isassumed to be at Technical Specification of 1 jiCi/gm DE 1-131 (equilibriumTechnical Specification limit for full power operation). Immediately followingthe accident the iodine appearance rate from the fuel to the primary coolant isassumed to increase to 500 times the equilibrium appearance ratecorresponding to the 1 giCi/gm DE 1-131 coolant concentration. The durationof the assumed spike is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The initial primary coolant noble gas activityis assumed to be at Technical Specification levels.The initial secondary coolant iodine activity is the Technical Specification limit of0.1 gCi/gm DE 1-131.Plant Technical Specification limits primary to secondary steam generator (SG) tubeleakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. Toaccommodate any potential accident induced leakage, the LOL dose consequenceanalysis addresses a limit of 0.75 gpm from all 4 SGs (or a total of 1080 gpd).The entire primary-to-secondary tube leakage of 0.75 gpm (maximum leak rate at STPconditions; total for all 4 SGs) is leaked into an effective SG. In accordance withRegulatory Guide 1.183, the pre-existing iodine activity in the secondary coolant andiodine activity due to reactor coolant leakage into the 4 SGs is assumed to behomogeneously mixed in the bulk secondary coolant. The effect of SG tube uncoveryin intact SGs (for SGTR and non-SGTR events) has been evaluated for potentialimpact on dose consequences as part of a WOG Program and demonstrated to beinsignificant. Therefore, per Regulatory Guide 1.183, the iodines are released to theenvironment via the via the main steam safety valves (MSSVs) and 10% atmosphericdump valves (ADVs) in proportion to the steaming rate and the inverse of a partitioncoefficient of 100. The iodine releases from the SG are assumed to be 97% elemental15.5-39Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEand 3% organic. The noble gases are released freely to the environment withoutretention in the SG.The condenser is assumed unavailable due to a coincident loss of offsite power.Consequently, the radioactivity release resulting from a LOL event is discharged to theenvironment from the steam generators via the MSSVs / 10% ADVs. The SGreleases continue for 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />, at which time shutdown cooling is initiated viaoperation of the Residual Heat Removal (RHR) system, and environmental releasesare terminated.15.5.10.2.3 Offsite Dose AssessmentAST methodology requires that the worst case dose to an individual located at anypoint on the boundary at the EAB, for any 2-hr period following the onset of theaccident be reported as the EAB dose. For the LOL event, the worst two hour periodcan occur either during the 0-2 hr period when the noble gas release rate is thehighest, or during the t=8.73 hr to 10.73 hr period when the iodine level in the SGliquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the startingpoint of the worst 2 hr window, the 0-2 hr EAB X/Q is utilized.The bounding EAB and LPZ dose following a LOL event at either unit is presented inTable 15.5-9.15.5.10.2.4 Control Room Dose AssessmentThe parameter values utilized for the control room in the accident dose transportmodel are discussed in Section 15.5.9. A summary of the critical assumptionsassociated with control room response and activity transport for the LOL event isprovided below:Control Room VentilationThe LOL event does not initiate any signal which could automatically start the controlroom pressurization air ventilation. Thus the dose consequence analysis for the LOLevent assumes that the control room remains in normal operation mode.Control Room Atmospheric Dispersion FactorsDue to the proximity of the MSSVs/10% ADVs to the control room normal intake of theaffected unit, and because the releases from the MSSVs/10% ADVs have a verticallyupward discharge, it is expected that the concentrations near the normal operationcontrol room intake of the affected unit (closest to the release point) will beinsignificant. Therefore, only the unaffected unit's control room normal intake isassumed to be contaminated by releases from the MSSVs/1 0% ADVs (refer toSection 2.3.5.2.2 for detail).15.5-40Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe bounding atmospheric dispersion factors applicable to the radioactivity releasepoints / control room receptors applicable to an LOL event at either unit are providedin Table 15.5-9B. The x/Q values presented in Table 15.5-9B take into considerationthe various release points-receptors applicable to the LOL to identify the bounding X/Qvalues applicable to a LOL event at either unit, and reflect the allowable adjustments /reductions in the values as discussed in Section 2.3.5.2.2 and summarized in thenotes of Tables 2.3-147 and 2.3-148.The bounding Control Room dose following a LOL event at either unit is presented inTable 15.5-9.The amount Of radioactive iodine released depends on the amount of steam releasedand the iodine concentration in the steam generator water prior to the accident. Ananalysis of potential thyroid doses has been made over the full range of possible valuesof these two key parameters'; the results are presented in Figures 15.5 2 through;15.5 5. As shown on the figures, the potentia! thyroid doses arc higher with incresinsteam releasee and iodine oncentrations. Figures 15.5 2 and 15.5 3 are results thatassume Regulator- ' Guide 1.4, Revision 1, assumptions for post acrident meteerologyand- b-reathing rates (Design Rasis; C-ase Ass6umptions) .As shown in Figure 15.5-2-,approximately 1.6 x1 Imostaishemaximumn steam release expected for a fullcooldow n w.ithout any cneerailbitand a steam release of approximatelyI x 1 4'5 1Ibm would result fromn releasing only the contents Of one steam generator due toa safet valve rele-ase or steam "Rne break With condenser cooling available.Figures 15.5 2 through 15.5,5 illustrate the range of possible thyroid doses fromCondition I! ev.... The highest anticipated doses would result from an event suc.h asloss of electFrial load, and the potential thyroid and- whole body dose.s .hisrparticular ev.ent have been anaye usin the EMERALD program. For both the designbasis ae an;d the expected case.. , itwasassumed that 656,000 Ibm of steam would beto the atmosphere during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and an additional 1,035,000 Ibmwould be released the followi ng 6 houFr for a limiting total release of about14.7F+06 ibm (see Table 6.4.2 1 of Reference 49 for a summar' of RSG and RSGCondition 11 event steam releases). The assumptions used forF meteorology, breathingrates, population density, and other common factors were describePd in e~arlie~r_paragraphs. Note that the preceding steam release quantities are associated with theoriginal steam generator (OSG) los6 of load (LOL) analysis which pr.ovides the basis forthe dose analysis of recr.Gd. These values are greater than the replacement steamgenerator (RSG) LO L with Tayg and T-feed Range analysis releases (651,000 !bm and1,023,000 Ibmn, respectively) and are therefore bounding since total dose is proportionalto total steamn release.For the design basis case, it was- assumed that the plant had been operating-continuously With 1 percent fuel caladding defects and I gpm primar,' to secondaryleakage. For the expected case c~alc~ulation, operation at 0.2 percent defects and-20 gallons per day to the secondary was assumed. in both cases, leakage of waterfromq primnary to secondary was assumed to continue during cooldown at 75 percent ofthe pre accident rate duFrig the 46rt 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at 50 percent of the pre accident rate15.5-41 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEduring the next 6 hourS. These values were derived frmpia, to secon-dar;Jpressre,,U4 differen~tials during cooldownit was also conse~vatively assumed for bo0th c-alses that the iodine partition factor on thesteam generatorS rcleasing steam was 0.01, on a mass basis. in addition, to accountfo4r the effect of iodino spiking, fuel eScape rate coefficients forF iodines of 30 times thenorm~al operation values given in Table 11.1 8 were used for a period of 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s-folloAWing the start of the accnide-nt. Other detailed and less significant modelinassumptions are presented in Reference 4.The rcsuling potential exposures from this type of accident are summarized inRgurcs 15.5 2 through 15.5 5.15.5.10.3 ConclusionsIt can be concluded from the results discussed that the occurrence of any of the eventsanalyzed in Section 15.2 (or from other events involving insignificant core damage, butrequiring atmospheric steam releases) will result in insignificant radiation exposures andare bounded by the LOL event.Additionally, the analysis demonstrates that the acceptance criteria are met as follows:(1) The radiation dose to the whole body and to the thyroid of an individuallocated at any point on the boundary of the exclusion area for the twe-heWsany 2-hour period immediately-following the onset of the postulatedfission product release is within 0.025 Sv (2.5 rem) TEDE a.e asshown in Table 15.5-9.(2) The radiation dose to the whole body and to the, hyfrid of an individuallocated at any point on the outer boundary of the low population zone, who isexposed to the radioactive cloud resulting from the postulated fission productrelease (during the entire period of its passage), is within 0.025 Sv (2.5 rem)TEDE shown in Table 15.5-9.(2Q(3) The radiation dose to an individual in the control room for the duration ofthe accident is within 0.05 Sv (5 rem) TEDE as shown in Table 15.5-9.15.5-42Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.11 RADIOLOGICAL CONSEQUENCES OF A SMALL-BREAK LOCA15.5.11.1 Acceptance CriteriaThe radiological consequences of a small-break loss-of-coolant-accident (SBLOCA)shall not exceed the dose limits of 10 CFR 100-.4 50.67, and will meet the doseacceptance criteria of Regulatory Guide 1.183, July 2000 as outlined below:Regulatory Guide 1.183 does not specifically address Condition III scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area for any2-hour period following the onset of the postulated fission product release shallnot receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose -in excess of 0.025 Sv (2.5 rem) TEDE.15.5-43Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEI.An individual located at any point on the boundar,' of the eXclusion area fort-he t'::oQ hours immediately followi ngte onset Of the postulated ftissinproduct rFelP-a4se shall not reGeive a total radiation dose to the whole body ineXcess of 25 rem Or al ttal radiation dose iR eXcess Of 300 rem to the thyroidfrom iodine exposure-.ii. An individual located at any point on the outer bounda,,' of the lowpopulation zone, who is exposed to the radio.actie cloud resulting from thepostulated fission product release (during the entire period of its passage),shall not receive a total radiation dose to the whole body in eXcess bf 2ran, + +r i

• f.,I rnl,,ta rlne ~. ~f ')f An n , +n +ka +t,,r .4 fran 1;nA.nf.J At.I L.L~f* %J *I~.%S %JIf l 5 IT .fA* .*rwrri, k7rif vct rd ca Uri WOU ri VM;Fvý-- V rum ty V irr" 17M V r1wexpes4uFeT01+in accordanc~e with the requirementsof GDC 19, 1971, the dose to the contrFoiuui! ope~I.2LU uinutI auuiueii GGR11UItIUIS S.IIll 1not be III eXes of aJ Fe.! WHuleNOV OF RsII

• I,,,% requivalent to any padt of the body (i.e., 30rAp rthymeid a~dbetaU PPsk;in, R'e;ewiie 0 l101 TO LiI GUUIatieUi- t01 ULLaGiUUIL.15.5.11.2-Identification of Causes and Accident DescriptionAs discussed in Section 15.3.1, a SBLOCA (defined in UFSAR Chapter 15.3.1 as abreak that is large enough to actuate the emergency core cooling system), is notexpected to cause fuel cladding failure. For this reason, the only activity release to thecontainment will be the dissolved noble gases and iodine in the reactor coolant waterexpelled from the pipe rupture. Some of this activity could be released to thecontainment atmosphere as the water flashes, and some of this amount could leak fromthe containment as a result of a rise in containment pressure.The possible radiological consequence of this event is expected to be bounded by the"containment release" scenario of the CREA discussed in Section 15.5.23.The dose consequences following a SBLOCA will be significantly less than a CREAsince the CREA is postulated to result in 10% fuel damage, whereas the SBLOCA hasno fuel damage.As demonstrated in Table 15.5-52, the dose consequences at the EAB and LPZfollowing a CREA is within the acceptance criteria applicable to the SBLOCA.The detailed des-ription of the models used in calI-Ulatng the potential exposures fromna smnall LOCA is contained in Refe-re-nce, 4, and a genera! description is contained inSection 15.5.17 of this repodt. The specific assumptions used in the analysis are asfig-IIsi(!1) The fission ,oduIctinventGFrie, mneteorological data, breathing rates, aRddata are described in SeG;ons 15.5.3, 15.5.5, 15.5.6, and15.5.7, respectiVely. Other commone assumptions are deScribed in the15.5-44Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEprvius sections of 15.5.41 4. C I., A 11 L, 4." it, t aI " IV I. to t- L I-to-%9 ,f , tt-tInI t L-- --- --ILor'."Wactivitics associated with 1 percent defective cladding were used; andfothe expected pace, the reactor coolant actiVities associated with 0.2percent defiective clad-ding were used. These activities and concentratinare listed in Tables 11.1-11 and- 11.1 12, and all models and assumptionsEused in determininig these values are deScribed in SectionF 11 .1.Of uthen amounts UT Roble gaseG containca in the primay Gg'eolant,inn fAn man4in AIC fi -Irar Ia A~ +aisa 4k% 4., an +l n+~nn ~, a. ir;k, 4; ; +k~ ;. 4 r.C 4k1; A;lA 01 + ! in +f +k 4; 1 4 , A; +k i +I. --i

• AL -k.nn~ir 4i +m I km !a 1k~i4. ; 01s; 14;n IIn~,n, A +k4 + k+ aa fna ;Anf n l4; i'I4Anm k'f -ia na n+k + i. 4m nI*%+ 5I* IU**UL4 tf +l, IAA 4I KlL 41- -k*Ie sfin n i r m na %7 V r1V In TI, 0n rain- MW 0 Wr- V'4 ny 10r opal( r fi lcH,j'ac; t anifltnosrtl .eIA4 in addiation to aGGOUnt OFme etf ctorit e;wine SnIKIng all ofthme aGtiviP..x *ireleased from the fuel up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident is assumed to ereleased to the contafinment. Of the amounts of noble gases released to-the containment, 100 percen~t is assumed to be released to thecontainment atmosphere. For the iodines, it is assumed that onlY10 percent of the iodines released to the containment are released tothcontainment atmosphere,(5) The spray removal rates forF the SIBLOCA arc assumed to be the same asthose applicable for the large break LOCA as described inSection 15.5417.(6) The containment leakage rates in this analysis are also assumed to be thesame as for the large break LOCA and are discussed in Section 15.5.1-7.The resulting potential exposures are listed in Table 15.5 10 and demonstrate that allcalculated doses are well below the guideline values specified in 10 CFR 100.11. Sincethe activity releases from this, type of event will be significantly lower than those fromn alarge break LOCA, anY control room exposure Whicah might occGur would be well mwithinthe established criteria discussed in Section 15.5.17. in addifion, because of ig -n4fcnllo-wer fission product releases to the sump and the absence of any Zirconium waterreaction, the amounts of free hydrogen produced by GumAp radiolysis foel owing a smallI Ct'A .,m .A inlI~ a aa.,aC15.5.11.3 ConclusionsT~L... ~ ~ AL..A AL------. -- C-1i ------I I1e iliiiV1i de~nRiiitiue6 trat th LIIC eLJgjlle Lli~e~t1 ia te uiw15.5-4515.5-45Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(1) The r-adiation dose to the whole body and to the thyroid of an indiv.idualloc~ated at any point on the boundar,' of the eXclusion area for the txo hoursimmediately foelewin g the onset of the postulated fission product release arcwell withhin the dose limits of 10 CFR 100.11 as sho1Wn in Table 15.5 10.(2) The radeiaion dose to the w..hole body and to the thyGrod of an ind;d'u"'A"leuateu at any peint en the euteF D0Unuary 9+ tne iow pepulatten Z9Re, whe isexposed to the radioactive omud resulting prom the postulated fitaiOt productrelease (during the entire period of 46 passage), are well within the doselimits ~ ~ ' of1IFR101 as shown in Table 15.5 10.(3) Since the actiVit', releases from the SB3LOCA are less than those fromn alarge break LOCA. ([BLOC=9A), any conrol1 room dose which Might occuwould be well within the established criteria discGussed in Sectio 15.5.7On the basis of this conservative comparison approach, it is concluded that the doseconsequences at the EAB and LPZ following a SBLOCA will remain within theacceptance criteria listed in Section 15.5.11.1.15.5.12 RADIOLOGICAL CONSEQUENCES OF MINOR SECONDARY SYSTEMPIPE BREAKS15.5.12.1- Acceptance CriteriaThe radiological consequences of accidents analyzed in Section 15.3 such as minorsecondary system pipe breaks shall not exceed the dose limits of 10 CFR 100.1 1 asoutlined below;10 CFR 50.67, and will meet the dose acceptance criteria of RegulatoryGuide 1.183, July 2000 as outlined below:Regulatory Guide 1.183 does not specifically address Condition III scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area forany 2-hour period following the onset of the postulated fission product releaseshall not receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.15.5-46Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.An indiVidu1-al locnated at any point on the boundary of the exclusion area for the twohours immediately foelowing the onset of the postulated fission product release shall notreceive a total radiation dose to the whole body in excess of 25 rem Or a Moa! radiationdose in excess Of 300 remA to the thyroid from iodine exposure.An individual located at any point on the outer boundary of the low population zone, whoisepsed to the radioactive cioud reSUlting fromn the postulated fissio product4rekease-(uigthe entire period Of its passage), shall not receive a total radiation dose to theWhO ho~ bdy in excess Of 25 rem~ or a total radiation dose in.,cs of 300 Frem to the-w- ý ý --J".ITn rUOn iiro i loinp xOouunre.15.5.12.2- Identification of Causes and Accident DescriptionThe effects on the core of sudden depressurization of the secondary system caused byan accidental opening of a steam dump, relief or safety valve were described inSection 15.2 and apply also to the case of minor secondary system pipe breaks. Asshown in that analysis, no core damage or fuel rod failure is expected to occur. InSection 15.&484.2, analyses are presented that show the effects on the core of a majorsteam line break, and, in this case also, no fuel rod failures are expected to occur.The analyses presented in Section 15.3.2 demonstrate that a departure from nucleateboiling ratio (DNBR) of less than the safety analysis limit will not occur anywhere in thecore in the event of a minor secondary system pipe rupture.The steam releases following a minor secondary line break is expected to besignificantly less than that associated with a main steam line break.As demonstrated in Table 15.5-34, the dose consequences at the EAB and LPZfollowing a MSLB is within the acceptance criteria applicable to the minor secondary linebreak.?a .... :l-l- --= J.'--I----:----IX -1r~n~FI~nrf~ fiT Ifl!~ f~"f~flT flJU~ Ifi 1flU~ [flif P-~I? fT ~fi~T1I9 ~II-~i!IIthat Might contain radioactive iedines, are disncussed Hn. Section 1551.The resulting-thyroid doses are presented parametrically in Figures 15.5 2 through 15.5 5 as a-fucinof quantity of steamn released and -eodr'system activity. in the event thata comnplete plant coo Idown wifthut condenser cooling capacity is necessary followingthe break, the potential exposures would be the same as those Fepo~ted in Table 15.5 9foqr lios of eIectricanlad15.5-47Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.12.3 ConclusionsOn the basis of this conservative comparison approach, it is concluded that the doseconsequences at the EAB and LPZ following a minor secondary system pipe rupture willremain within the acceptance criteria listed in Section 15.5.12.1.On the basis of the discussed results, it can be concluded that the potential exposuresf4llowing a minor econdla- system pipe rupture would be insignifcantAdditionally, the ancalysis demonstates that the acceptance criteria are met as follows;The radiatien dose to the whole bodiy and t the thyroid of an individual located at anypoint on the brundag' of the exu salon area for the no hours immediately feollogingthe onset of the postulated fission product release arc insignificant as shown in TableThe radiation dose to the whole body and to the thyroid of an individual located at anypoint on the outer boundas, of the flw population zone, who is exposed to theradioactive lonud resulting from the postulated fission produgt release (durithpeleentire pewrid of its passage), are insignifcant as shown in15.5.13 RADIOLOGICAL CONSEQUENCES OF INADVERTENT LOADING OF AFUEL ASSEMBLY INTO AN IMPROPER POSITION15.5.13.1 Acceptance CriteriaFuel assembly loading errors shall be prevented by administrative proceduresimplemented during core loading. In the unlikely event that a loading error occurs,analyses supporting Section 15.3.3 shall confirm that no events leading to radiologicalconsequences shall occur as a result of loading errors.15.5.13.2 Identification of Causes and Accident DescriptionFuel and core loading errors such as inadvertently loading one or more fuel assembliesinto improper positions, loading a fuel rod during manufacture with one or more pelletsof the wrong enrichment, or loading a full fuel assembly during manufacture with pelletsof the wrong enrichment will lead to increased heat fluxes if the error results in placingfuel in core positions calling for fuel of lesser enrichment. The inadvertent loading ofone or more fuel assemblies requiring burnable poison rods into a new core withoutburnable poison rods is also included among possible core loading errors. Because ofmargins present, as discussed in detail in Section 15.3.3, no events leading toradiological consequences are expected as a result of loading errors.15.5-48Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.13.3 ConclusionsBecause of margins present, as discussed in detail in Section 15.3.3, no events leadingto radiological consequences are expected as a result of loading errors.15.5.14 RADIOLOGICAL CONSEQUENCES OF COMPLETE LOSS OF FORCEDREACTOR COOLANT FLOW15.5.14.1 Acceptance CriteriaThe radiological consequences of small amounts of radioactive isotopes that could bereleased to the atmosphere as a result of atmospheric steam dumping required for plantcooldown following a complete loss of forced reactor coolant flow shall not exceed thedose limits of 10 CFR 100.11 as outloned below:50.67, and will meet the doseacceptance criteria of Regulatory Guide 1.183, July 2000 as outlined below:Regulatory Guide 1.183 does not specifically address Condition III scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area forany 2-hour period following the onset of the postulated fission product releaseshall not receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.15.5-49Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE--f~L.. L...A

• i* I I I I I I IAn hndliul loc-ated at any polnt on ihe i~unnar;of the eXcIU6on areaT r Tne twohourS immediately following the onset of the postulated fission product release shall notree, a totai Facmation doscT Me ncWnoie assy in 2.5e 01 F em OF a total FaiatIGRdose in excess of 300 remn to the thyroid from iodine exposure.An individual !ocated at any point On the outer boundary of the low population zone, whois 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 to thewhole body iR eXcess of 25 rem. or a total radiation dosinecs of 300 rem to thethyroid from iodine exposure.15.5.14.2 Identification of Causes and Accident DescriptionAs discussed in Section 15.3.4, a complete loss of forced reactor coolant flow mayresult from a simultaneous loss of electrical supplies to all reactor coolant pumps(RCPs). If the reactor is at power at the time of the accident, the immediate effect ofloss of coolant flow is a rapid increase in the coolant temperature.The analysis performed and reported in Section 15.3.4 has demonstrated that for thecomplete loss of forced reactor coolant flow, the DNBR does not decrease below thesafety analysis limit during the transient, and thus there is no cladding damage orrelease of fission products to the RCS. For this reason, this accident has no sigRnftiantefferts.The possible radiological consequence of a complete loss of forced reactor coolant flowis expected to be bounded by the conservative Loss-of-Load scenario with a coincidentLoss of offsite power described in Section 15.5.10.As demonstrated in Table 15.5-9, the dose consequences at the EAB and LPZ followinga Loss of Load is within the acceptance criteria applicable to the complete loss of forcedreactor coolant flow.15.5-50Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.14.3 ConclusionsOn the basis of this comparison approach, it is concluded that the dose consequencesat the EAB and LPZ following a complete loss of forced reactor coolant flow willremain within the acceptance criteria listed in Section 15.5.14.1.The .analysis-decrinrbed- in SectA-ion 1 5.3.4 demonStrates that there are no significant enViFronment-aleffects of the Complete Loss6 Of Forced Reactor Coolan~t Flow event. Teeoe hacceptance criteria are mnet as followfs:The radiation dose to the whole body and to the thyroid of an individual located at anypoint on the boundary of the eXclusion area for the he hoUrM immdatl folowingthe onset of the postulated fissioi; product release are insignificant.-The radiation dose to the whole body and to the thyroid of an individual located at anypoint on the outer boundary of the low population zone, who is exposed to theradioactive cloud Fesulting from the postulated fission pro~duct release (duringthentire period of its passage), are inSfignifca-Rt.15.5.15 RADIOLOGICAL CONSEQUENCES OF AN UNDERFREQUENCYACCIDENT15.5.15.1 Acceptance CriteriaThe radiological consequences of small amounts of radioactive isotopes that could bereleased to the atmosphere as a result of atmospheric steam dumping required for plantcooldown following an underfrequency accident shall not exceed the dose limits of 10CFR 100. 11 as outlined below:. 50.67, and will meet the dose acceptance criteria ofRegulatory Guide 1. 183, July 2000 as outlined below:Regulatory Guide 1.183 does not specifically address Condition Ill scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area for any2-hour period following the onset of the postulated fission product release shallnot receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.15.5-51Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.An individual located at any point on the boundary of the exclusion area for the twehourS immediately following the onset of the postulated ,fision produt release shall notrecei5e2a tetal radiation dose to the whole body in excess of 25 rem Or a total radiationdose in exess of 300 frem to the thyreid frem iodine exposure.An individual located at any point on the outer boundary of the low population zone, whoes epotsed to the sadioaative aloud resulting from the postulated fissien product release(duing the eatire peried of its passage), shall not receive a total radiation dose to thehole body in exess ef 25 rem, 9o a total radiation dose in excess of 300 rem to thethyroid from iodine exu reqew.15.5.15.2- Identification of Causes and Accident DescriptionA transient analysis for this unlikely event has beecr offsite pois discussed in Section15.3.4. The analysis demonstrates that for an underfrequency accident, the DNBRdoes not decrease below the safety analysis limit during the transient, and thus there isno cladding damage or release of fission products to the RCS. However, smallamounts of radioactive isotopes could be released to the atmosphere as a result ofatmospheric steam dumping required for plant cooldown.The possible radiological consequence of this event is expected to be bounded by theconservative Loss-of-Load scenario with a coincident Loss of offsite power described inSection 15.5.10.As demonstrated in Table 15.5-9, the dose consequences at the EAB and LPZ followinga Loss of Load is within the acceptance criteria applicable to an underfrequencyaccident.A detailed discussiORon the ptnilenyironmental ogsqeRe f accidents-invelving atmospheric steam dumping is prescnted in Section 15.5-10. From the15.5.15.3 ConclusionsOn the basis of this comparison approach, it is concluded that the dose consequencesat the EAB and LPZ following a complete loss of forced reactor coolant flow will remainwithin the acceptance criteria listed in Section 15.5.15.1. On the basis of the potential15.5-52Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEexposures dicussed, it an be concaluded that, although ,er.' unlikely, the occurrence ofthis would not cause undue risk to the health and safety of the public.Additionally, the anal.sis dem;onStrates that the ... ....a. m.....et as follows.The rad4atin dose to the whole body and to the thyroid of an indiv.'idual loated at anypoint on the boundary of the exclusion area foF the two hours immediately following theonset of the portulated fissinn pFrduct release are inignRiAfant as shoWn in Table 15.544~-The radiation dose to the whole body and to the thyroid of an individual located at anypoint on the outer boundar; of the low population zone, who is exposed to theradieaGtiey cld r, ,esulting from the postulated fission product release (during the entireperiod of its passage), are insignificant as shown in Table 15.5 11.15.5.16 RADIOLOGICAL CONSEQUENCES OF A SINGLE ROD CLUSTERCONTROL ASSEMBLY WITHDRAWAL AT FULL POWER15.5.16.1 Acceptance CriteriaThe radiological consequences of a single rod cluster control assembly withdrawal shallnot exceed the dose limits of 10 CFR 100.11 as outlined below,,,50.67, and will meet thedose acceptance criteria of Regulatory Guide 1.183, July 2000 and outlined below:Regulatory Guide 1.183 does not specifically address Condition III scenarios. However,per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows alicensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all doseconsequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicatesthat for events with a higher probability of occurrence than those listed in Table 6 ofRegulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed thecriteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will belimited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limitimposed by 10 CFR 50.67.EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area for any2-hour period following the onset of the postulated fission product release shallnot receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.15.5-53Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEw J iI I I iI:-\n. ina:'iauai Iocatea a; any point on tne bounaar; orinhe exclusio~n. area ror Me Merec5i16.2-total radiation dose to the whole body idnt eessoriof 25rdometin exaess of 300 rem to the thyroid feom iodine expoSure.An individual located at any powirt o the outer boundar; of the low populatieonzone, whoas exposed te the radioBactive chloud resulti from the postulated fissien pfrdeut felease(during the .entiepeod of its passage), shall not receive a total radiation dose to theiwhele body in eXcess of 25 rem, Or a total radiation dose in eXcess of 300 Frem to thethyroid frem iodine exposure.15.5.16.2- Identification of Causes and Accident DescriptionA complete transient analysis of this accident is presented in Section 15.3.5. For thecondition of one rod cluster control assembly (ROCA) fully withdrawn with the rest of thebank fully inserted, at full power, an upper bound of the number of fuel rodsexperiencing DNBR less than the safety analysis limit is 5 percent of the total fuel rodsin the core.The possible radiological consequence of this event is expected to be bounded by theCREA discussed in Section 15.5.23.The dose consequences following a single rod cluster control assembly withdrawal willbe less than a CREA since the CREA is postulated to result in 10% fuel damage,whereas the condition of one rod cluster control assembly fully withdrawn with the restof the bank fully inserted, at full power has only 5% fuel damage.As demonstrated in Table 15.5-52, the dose consequences at the EAB and LPZfollowing a CREA is within the acceptance criteria applicable to the condition of one rodcluster control assembly fully withdrawn with the rest of the bank fully inserted, at fullpower. A det-ailed diScussion of the potential radiological consequences of accidentsinvolving small amounts of fuel rod failure is included in Section 15.5.21. From~ theparamnetric analyses presented in that sec-tion, the potential exposures from an RCCA15.5-54Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.16.3 ConclusionsOn the basis of this comparison approach, it is concluded that the dose consequencesat the EAB and LPZ following the condition of one rod cluster control assembly fullywithdrawn with the rest of the bank fully inserted, at full power will remain within theacceptance criteria listed in Section 15.5.16.1.On the basis of the potential exposures discaussed, it can be concluded that thoccuFF-Grrenecf thisacdn ol not cause u-ndue Prik to the health and safet of thepublir..AdtIIe 1 the ay d AIt tha cc Ie aI II l sThe radiation dose to the whole body and to the thyroid of an individual located at anypoint on the boundar,' of the exclusion area for the two hours immediately followingthe onset of the postulated fission product release are insignificant as show n On TableThe radiation dose to the whole body and to the thyroid of an individual located at anypoint on the outer boundar; of the low population zone, who is exposed to the-radioactive cloud resulting from the postulated fission proeduct release (duringthentire period of its passage), are insignificant as shownnTable 15.5 12.15.5.17 RADIOLOGICAL CONSEQUENCES OF MAJOR RUPTURE OF PRIMARYCOOLANT PIPESVarious aspects- of the radiologicsal consequenc6es f a large break loss of cooelantaccident (LIBLOGA) are presented in this section.15.5.17.1 Acceptance CriteriaThe radiological consequences of a LOCA shall not exceed the dose limits of 10 CFR50.67, and will meet the dose acceptance criteria of Regulatory Guide 1.183, July 2000and outlined below:EAB and LPZ Dose Criteria(1) An individual located at any point on the boundary of the exclusion area forany 2-hour period following the onset of the postulated fission product releaseshall not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE.(2) An individual located at any point on the outer boundary of the low populationzone, who is exposed to the radioactive cloud resulting from the postulatedfission product release (during the entire period of its passage), shall notreceive a total radiation dose in excess of 0.25 Sv (25 rem) TEDE.15.5-55Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEControl Room Dose CriteriaAdequate radiation protection is provided to permit access and occupancy of thecontrol room under accident conditions without personnel receiving radiationexposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.Technical Support Center Dose CriteriaThe acceptance criteria for the TSC dose is based on Section 8.2.1(f) of NUREG-0737,Supplement 1, as amended by Regulatory Guide 1.183, Section 1.2.1, and 10 CFR50.67. The dose to an operator in the TSC should not exceed 5 rem TEDE for theduration of the accident.(1) The radiologiGal consequences of a maj;r rupture of oolant pipesshall take into consideration fission product releases due to leakage from thecontainment, post LOCA recircUlation Loop leakage in the Auxiliary Buildirn(inclusive of a residual heat removal (RHR) pump seal failure resuting in a gpm 3 leak for 30 minutes staring at T-24 hrs post LOCA), and containmentshhine.(2) The radiological consequences6 of a mnajor rupture of primnary coolant pipesshall not exceed the dose limits of 10- C-FR 10-0.11 a6 outlined below:i. An individual located at any point on the boundary of the exclusion areaforo the b.vo immediately following the onset of the postulatedfission product release shal! not receive a total radiation dose to thewhole body in excess of 25 rem OF a total radiation dose in excess of300 remA to the thyroid from iodine exposure.ii nindiv.idual located at any point on the outer boundary of the low-population zone, who is exposed to the radioactive cloud reesultingfrom the postulated fission product release (during the entire periodof its passage), shall not r-eceive a total radiation dose to the wholebody in excess of 25 rem, Or a total radiation dosinecs of 300rem to the thyroid fromn iodine exposure.(3) in acc.ordancGe with the requiremnents of GDC 19, 1971, the dose to thecontra! roomR operator u-nderaccidWent conditions shall not be in excess" of 5rem whole body or its equivalent to any part of the body (i.e., 30 remRthyroid and beta skin, Referencse 51) for the duration of t-he accioden-t.(4) in the event controlled venting of the containment is imlmned otLOCA using the containment hydrogen purge system (serves as a back upGapabilit' for hydrogen control to the hydroegen recoembiners), an individuallocated at any point on the boundary of the excGucio ara w.ho is exposedto the radioactive cloud Frsulting from the postulated fission product15.5-56 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATErelease (during the entire period of its passage), shall not roceive a totalradiation dose to the whole body in eQes of05 rem/year in accordance_with 10 CFR Part 20.15.5.17.2 Identification of Causes and Accident Description15.5.17.2.1 B-ai-c Events and Release FractionsActivity Release PathwaysThe accidental rupture of a main coolant pipe is the event assumed to initiate a -L-Blargebreak LOCA. Analyses of the response of the reactor system, including the emergencycore cooling system (ECCS), to ruptures of various sizes have been presented inSections 15.3.1 and 15.4.1. As demonstrated in these analyses, the ECCS, usingemergency power, is designed to keep cladding temperatures well below melting and tolimit zirconium-water reactions to an insignificant level. As a result of the increase incladding temperature and the rapid depressurization of the core, however, somecladding failure may occur in the hottest regions of the core. Following the claddingfailure, some activity would be released to the primary coolant and subsequently to theinside of the containment building. Active mechanisms include radioactive particulateand iodine removal by the containment sprays inclusive of the containment air mixingprovided by the CFCUs. Section 6.2 describes the design and operation of the CSSand the CFCUs. Bccause of the pr.essuization of the containment building by theprimay coolant water escaping from the pipe break, some of the volatile radioactiveiodines and noble gases could leak f..ro the containmhent building to the atmosphere.it is not expected that a significant amount of organic iodine would be iberated froM thefuel as a result of a LBLOhA. This conclUsion is based on the results of fuel meltdownexeients conducted by the Oak Ridge National Laboratory. The fraction of the totalioiethat is released in organic formns is expected to be on the order of 0.2 percent, Orless, since the rate of thermal radiolytic decomposition would exceed the rateOrganic compounds of io-dine can be for~med by reaction of absorbed elemen~tal iodineon su.faces of the containment vessel. E*pe.rim.ents have shown that the rate of-formation is dependent on specific ,onditions such as the oncentration of iodipe,concentration Of impurfities, radiation level, pressure, temperature, and relative humnidity.Tho ra2te Of conversion of airborne iodine is- proportional to the surface to volumne ratio ofthe enclosure, whether the process is limited to diffusion to the surface Or by threaction ra2te of the- abhsorbePd_ iodine-. The obscr.'ed yields of organic iodine asfunction of aging time invros test encloGsures, with various volume to surfacae arearatios, were extrapolated to determine the values; for the DCPP containment vessel.The iodine conversion rates pmdreditedd in thi;s mnanner did not exceed 0.0005 percent ofthe atmospheric iodine per hour.The potential exposures following the postulated sequence of events in LBLOC=0As havebeen analyzed forF tWO cases. in the expected case, it has, been assumned that the entireinventor,' of volatile fission produc~ts oentaie inte 'elet cladding gap spaces i15.5-57Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEreleased to the coolant during the time the corc is being flooded by the ECrCS. of thisgap inyentor;, 25 percent of the iodines an 10- 0 percen;-t- of the noble gasesarconsidered to be released to the containment atmosphere immediately fo llewing thepip .ru ptur. In thiS respect, the expected case does contain some degree of-n-se-atism -ince the EGGS is designed to rfevent gross cladding damage. Inraccordance with the exe imetal data reported in the previous paragraph, the fractionof iodine that is relese inonrgani fo-rm. is- assumed to be 0.2 percen~t, and theproductionR rate of organic form~s is considered negligible. The iodine plateout rates areregligible (Reference 10) compared to the spray w,-ashout rates and are assum-ed t bezero. The partic~ulate fraction of iodine is also assumned to be zero for the expectedcas sine this fraction is small and the Spray removal rates for particulates is large asshown in Reference10-.For the design basis LOCA, it has been assumed that 25 percent of the equilibriumRadeiactiye iodine invent;en in the re is lavail able for leakage from theeactor caontainment. Ninety one percent of this 25 percent is assumed to be in the formof elemental iodine, 4 percent of this 25 percent is in the form of organic. iodides, and5 percent of this 25 percent is in the formn of particulate iodine. In addition, 100 percentof the noble gas inventod; in the ore is assumed to be immediately released to theC aientbuildig. As diussed in earlie paragfaph, releases of these Magnitudesa1e Rot expected to our, even if the EGGS does not pelform as expected. An analysisusing these assumptions is presented because these values are onsidered acceptablefor a design basis analysis in Regulator; Guide 1.4, Rev*ision 1.Regulatory Guide 1.183, Appendix A, identifies the large break LOCA as the designbasis case of the spectrum of break sizes for evaluating performance of releasemitigation systems and the containment, and for facility siting relative to radiologicalconsequences.DCPP has identified six activity release paths following a LOCA1. Release via the Containment Pressure / Vacuum Relief pathway to theenvironment until the containment isolation valves are closed.2. Containment leakage to the environment after containment isolation is achieved.3. Sump water leakage from ESF systems that recirculate sump water outsidecontainment.4. Failure of the RHR pump seal at T=24 hrs resulting in a 50 gpm leak of sumpwater for 30 mins.Note: DCPP design includes an ESF atmosphere filtration system, so from a regulatorystandpoint per Standard Review Plan Section 15.6.5, Appendix B (Reference 77), aswell as Regulatory Guide 1. 183, July 2000, inclusion of this leakage path in doseconsequences is not required. However, as noted in the following Sections, the RHR15.5-58 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEpump seal failure resulting in a "filtered" release is DCPP's licensing basis with respectto passive single failure.-Section 3.1.1.1 (Single Failure Criteria / Definitions), Item 2; discusses passivefailures -"The structural failure of a static component that limits the component'seffectiveness in carrying out its design function. When applied to a fluid system, thismeans a break in the pressure boundary resulting in abnormal leakage notexceeding 50 gpm for 30 minutes. Such leak rates are assumed for RHR pumpseal failure."-UFSAR Appendix 6.3A.3.2 (discusses passive failures), indicates that -the designof the auxiliary building and related equipment is based on handling of leaks up to amaximum of 50 gpm. Means are provided to detect and isolate such leaks in theemergency core cooling pathway within 30 mins. A review of the equipment in theRHR system loop and the CSS loop indicates that the largest leakage would resultfrom the failure of an RHR pump seal. Evaluation of RHR pump seal leakage rate,assuming only the presence of a seal retention ring around the pump shaft, showsthat flows less than 50 gpm would result (Chapter 6). Circulation loop piping leaks,valve packing leaks, and flange gasket leaks are much smaller and less severe thanan RHR pump seal failure leak.-UFSAR Section 15.5.17.2.8, indicates that- failure of an RHR pump seal at 24 hrsis assumed as the single failure that can be tolerated without loss of the requiredfunctioning of the RHR system.Therefore, the RHR Pump Seal Failure is retained as a release pathway for the ASTdose consequence analysis.5. Releases to the environment from the Miscellaneous Equipment Drain Tank(MEDT) which collects component leakage hard-piped to the MEDT. Thecollected fluid includes both post-LOCA sump water and other non-radioactivefluid.6. Releases to the environment via the refueling water storage tank (RWST) ventdue to post-LOCA sump fluid back-leakage into the RWST via the mini-flowrecirculation lines connecting the high head and low head safety injection pumpdischarge piping to the RWST.The LOCA dose consequence analysis follows the requirements provided in thepertinent sections of Regulatory Guide 1.183 including Appendix A. Table 15.5-23Alists the key assumptions / parameters utilized to develop the radiologicalconsequences following a LOCA at either unit.Computer code RADTRAD 3.03, is used to calculate the control room and site boundarydose due to airborne radioactivity releases following a LOCA.15.5-59Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.17.2.2 Activity Release Transport ModelSpray System I.dine Rem"ova'.RatesThoo Cntainment spray system (CSS) is described in detail, aRlon with a performanceanalysis, in Sectiens 6.2.2 and 6.2.3. The perfoancixAe analysioincludes theropresentation of the spatial distribution of driOpets and iodine in the containment, aswell as drop coalesmence and other effectd.Fori the expeted case analyses, the SS is assumed to funition with both epray pumpsoperating, gi ane effective elemental iodine remou val eoeffiient of 92 per houif. Onthe basis of experiments at Battelle, as decfribed in Reference 10, the npray remeoalrate for organic iodides was assumed to be 0.058 per hour.For the design basis case, it is assumed that one of the nO spay pumps fails ooperate, and the elesental iodine remoayl coefficient is prcdued to 31 per hour. Thisassumption is ossetwtqh au f3 per hour-used in the PSAR analysis. itTha anlyso beeassumed, for0 the design basis case, that the CSS has no effect nr theorgassumand parto iulate iedines.Alhours. It iscosequr afiety eassumt soed that 40tohelae Dlsesig anise Ginstantanousl31nde homoenoul mixed 26i n tpahe nainent atm)souldere anU~d tohatthex activiy2ssciaed with theF 2466ies ipnSpa he., 100%of the poblentiale and40%dofste iodn~e due the15.5.17.2.2.1 Containment Pressure /Vacuum Relief Line ReleaseIn accordance with Regulatory Guide 1.183, Appendix A, Section 3.8, for containmentssuch as DCPP that are routinely purged during normal operations, the doseconsequence analysis must assume that 100% of the radionuclide inventory in theprimary coolant is released to the containment at the initiation of the LOCA. Theinventory of the release from containment should be based on Technical Specificationsprimary coolant equilibrium activity (refer to Table 15.5-78). Iodine spikes need not beconsidered.Thus, in accordance with the above guidance, the 12 inch containment vacuum / overpressure relief valves are assumed to be open to the extent allowed by TechnicalSpecifications (i.e., blocked to prevent opening beyond 50 degrees), at the initiation ofthe LOCA, and the release via this pathway terminated as part of containment isolation.The analysis assumes that 100% of the radionuclide inventory in the primary coolant,assumed to be at Technical Specification levels, is released to the containment at T= 0hours. It is conservatively assumed that 40% of release flashes and is instantaneouslyand homogeneously mixed in the containment atmosphere and that the activityassociated with the volatiles, i.e., 100% of the noble gases and 40% of the iodine in thereactor coolant is available for release to the environment via this pathway.Containment pressurization (due to the RCS mass and energy release), combined withthe relief line cross-sectional area, results in a 218 acfs release of containment air to the15.5-60Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEenvironment for a conservatively estimated period of 13 seconds. Credit is taken forpressure boundary integrity of the containment pressure / vacuum relief systemductwork which is classified as PG&E Design Class II, and seismically qualified; thus,environmental releases are via the Plant Vent.Since the release is isolated within 13 seconds after LOCA, i.e., before the onset of thegap phase release, releases associated with fuel damage are not postulated. Thechemical form of the iodine released from the RCS to the environment is assumed to be97% elemental and 3% organic.15.5.17.2.2.2 Containment LeakageThe inventory of fission products in the reactor core available for release into thecontainment following a LOCA is provided in Table 15.5-77 which represents aconservative equilibrium reactor core inventory of the dose significant isotopes,assuming maximum full power operation at 1.05 times the current licensed thermalpower, and taking into consideration fuel enrichment and burnup. The notes provided atthe bottom of Table 15.5-77 provide information on isotopes used to estimate theinhalation and submersion doses following a LOCA, vs isotopes that are considered toestimate the post-LOCA direct shine dose.Per Regulatory Guide 1.183, the fission products released from the fuel are assumed tomix instantaneously and homogeneously throughout the free air volume of the primarycontainment as it is released from the core.In accordance with Regulatory Guide 1.183:a. Two fuel release phases are considered for DBA analyses: (a) the gap release,which begins 30 seconds after the LOCA and continues to t=30 mins and(b) the early In-Vessel release phase which begins 30 minutes into the accidentand continues for 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (i.e., t=1.8 hrs).b. The core inventory release fractions, by radionuclide groups, for the gap andearly in-vessel damage are as follows:Early In-VesselGroup Gap Release Phase Release PhaseNoble gas 0.05 0.95Halogens 0.05 0.35Alkali Metals 0.05 0.25Tellurium Group -0.05Ba, Sr -0.02Noble Metals -0.0025Cerium Group -10.0005Lanthanides -10.000215.5-61Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATENote: Footnote 10 criterion is met in that peak fuel rod burnup is limited to62,000 MWD/MTU.The elements in each radionuclide group released to the containment following a LOCAare assumed to be as follows (note that the groupings were expanded from that inRegulatory Guide 1.183 to address isotopes in the core with similar characteristics; theadded isotopes are in bold font):Noble gases: Xe, KrHalogens: I, BrAlkali Metals: Cs RbTellurium Grp: Te, Sb, Se, Sn, In, Ge, Ga, Cd, As, AgBa,Sr: Ba, SrNoble Metals: Ru, Rh, Pd, Mo, Tc, CoCerium Grp: Ce, Pu, Np, ThLanthanides: La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am, Gd, Ho, TbAs discussed in Section 6.2.3.3.7, the design includes chemical addition into thecontainment spray system which ensures a long term sump pH equal to or greater than7.0. Thus, the chemical form of the radioidine released from the fuel is assumed to be95% particulate (cesium iodide (Csl)), 4.85% elemental iodine, and 0.15% organiciodine. With the exception of noble gases, elemental and organic iodine, all fissionproducts released are assumed to be in particulate form.The activity released from the core during each release phase is modeled as increasingin a linear fashion over the duration of the phase. The release into the containment isassumed to terminate at the end of the early in-vessel phase, approximately 1.8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />safter the LOCA.Isotopic decay, containment leakage and spray removal are credited to deplete theinventory of fission products airborne in containment.Containment spray in the injection and recirculation mode is utilized as one of theprimary means of fission product cleanup following a LOCA. Mixing of the effectivelysprayed volume of containment, with the unsprayed volume of the containment isenhanced by operation of the PG&E Design Class I containment fan coolers. In order toquantify the effectiveness of the containment spray system, both the volume fraction ofcontainment that is sprayed, and the mixing rate between the sprayed and unsprayedvolumes are quantified.The LOCA analysis is based on an assumed worst case single failure of loss of oneESF train. A single train ESF consists of one train of ECCS, one train of CSS, and twoContainment Fan Cooling Units (CFCUs). A single train scenario is selected to beconsistent with the use of reduced iodine and particulate removal coefficientsassociated with single train operation.15.5-62Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEa. Containment Spray Duration: Containment Spray in the injection mode is initiatedat 111 seconds after the LOCA and terminated at 3798 seconds. Manualoperation is credited to initiate containment recirculation spray within twelve (12)minutes after injection spray is terminated. Thus, based on single trainoperation, containment spray in the recirculation mode is initiated at 4518seconds, and terminated 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later at 22,518 seconds. In summary,containment spray operation (injection plus recirculation) is credited for 6.25 hrspost-LOCA, with a twelve minute gap after injection spray is terminated.b. Containment Spray Coverage: As discussed in Section 6.2.3.3.7.1, thecontainment sprays are estimated to effectively cover 82.5% of the containmentfree volume during the containment spray injection as well as spray recirculationmode.c. Mixing between Sprayed and Unsprayed Regions of Containment: Thecontainment mixing rate between the sprayed and unsprayed regions following aLOCA is determined to be 9.13 turnovers of the unsprayed regions per hour.This mixing rate is based on the operation of two CFCU with a total volumetricflow rate that addresses surveillance margins and uncertainty, between theunsprayed regions and sprayed regions. Review of the layout and arrangementof the intake and exhaust registers of the CFCUs indicate that the air intakes areall located above the operating floor (sprayed region) and the air dischargeregisters are all located below the operating floor in the unsprayed region.Additional review of the containment configuration including the location of themajor openings in the containment structure, and various active and passivemixing mechanisms, results in the conclusion that following a LOCA, credit canbe taken for a) the entire flowrate provided by each operating CFCU to supportmixing between the sprayed and unsprayed regions, and b) homogeneousmixing within the sprayed and unsprayed regions, of the volume of air transferredfrom one region to the other due to CFCU operation. In accordance withRegulatory Guide 1.183, Appendix A, Section 3.3, prior to CFCU initiation, thedose consequence model assumes a mixing rate attributable to naturalconvection between the sprayed and unsprayed regions of 2 turnovers of theunsprayed region per hour.d. Fission Product Removal: The fission product removal coefficients developed forthe LOCA reflect the following guidance documents:i. Elemental iodine removal coefficients are calculated using guidanceprovided in Standard Review Plan Section 6.5.2, Revision 4 (Reference80) which is invoked by Regulatory Guide 1.183, Appendix A, Section 3.3ii. Time dependent particulate aerosol removal coefficients are estimated usingRegulatory Guide 1.183, Appendix A, Section 3.3, which permits the use oftime-dependent particulate aerosol removal coefficients by invokingNUREG/CR 5966, June 1993 (Reference 81), and indicates that no reduction15.5-63Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEin particulate aerosol removal coefficients is required when a DF of 50 isreached, if the removal rates are based on the calculated time-dependentairborne aerosol mass. There are several aerosol mechanics phenomena thatpromote the depletion of aerosols from the containment atmosphere. ForDCPP, agglomeration of the aerosol is considered in both sprayed andunsprayed regions. In the sprayed region, the particulate removal calculationtakes credit for the removal effectiveness of sprays and diffusiophoresis(aerosol removal due to steam condensation). Computer program SWNAUAis used to develop the time dependent particulate aerosol removal coefficientswhich reflect the effect of diffusiophoresis and sprays. Gravitational settling isconsidered only in the unsprayed region.The methodology used to develop the elemental iodine and particulate removalcoefficients in the sprayed and unsprayed region of the containment is discussed inSection 6.2.3.3.7.2. The total elemental iodine and particulate removal coefficients inthe sprayed and unsprayed region of the containment as a function of time aresummarized in Table 6.2-32.In summary, the activity transport model takes credit for aerosol removal due to steamcondensation and via containment spray based on spray flowrates associated withminimum ESF during the containment spray injection and recirculation mode. Itconsiders mixing between the sprayed and unsprayed regions of the containment,reduction in airborne radioactivity in the containment by concentration dependentaerosol removal lambdas, and isotopic in-growth due to decay.During spray operation in the injection mode, the elemental iodine removal rate for thesprays exceeds 20 hr', the maximum value permitted by NUREG-0800, StandardReview Plan Section 6.5.2; thus the elemental iodine removal rate attributable to spraysis limited to 20 hr1.During recirculation spray operation, the elemental removal rate forthe sprays is 19.34 hr'. As discussed in 6.2.3.3.7.2, the wall deposition removalcoefficient for elemental iodine has been calculated with the model provided in NUREG-0800, SRP Section 6.5.2. In sprayed and unsprayed regions, prior to spray actuation,the wall deposition removal coefficient is estimated to be 2.74 hr', while during sprayoperation, and in the sprayed region only, the wall deposition removal coefficient isestimated to be 0.57 hr1.In the unsprayed region, the aerosol removal lambdas reflect gravitational settling. Nocredit is taken for elemental iodine removal in the unsprayed region.Since the spray removal coefficients are based on calculated time dependent airborneaerosol mass, there is no restriction on the DF for particulate iodine. The maximum DFfor elemental iodine is based on Standard Review Plan Section 6.5.2 and is limited to aDF of 200.Radioactivity is assumed to leak from both the sprayed and unsprayed region to theenvironment at the containment technical specification leak rate for the first day, and15.5-64Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEhalf that leak rate for the remaining duration of the accident (i.e., 29 days). To ensurebounding values, the atmospheric dispersion factors utilized for the containment releasepath reflects the worst value between the containment wall release point, the plant Vent,the Containment Penetration Area GE (EL 140') and the Containment PenetrationAreas GW/FW (EL 140').15.5.17.2.3 Offsite E=XPoSUrcs fromF Containment LeakageAs a result of the pressurization of the containment following a LOCA, there is apossibility Of containment leakage du.rig the time that the contai.nment pressureabove atmospheric. For; the design basis case, the leakage rate has been assumed tobhe 0.1 percent 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 /> following the accident, and 0.05 percent perday after the 46rt day. These assumed rates are consistent with the Technical-Specifications (Reference 22)ilimit, the assumed rates con"sidered acceptable i..Regulator; Guide 1.1, Revision 1, and the values assumed in the PSAR analyses.For the expected case, the containment leakage rates used are 0.05 percent per day for-the first day and 0.025 percent per day for the periods after 1 day. These rates weredetermnined from averages of the actual predicted containment pressures presented in;preiou section;s, with the assumption that somne of the heat remova! systems do- notfunct-ion at full capacit.in this regard, the leakage rates. assdegree o-f cosea ism sine the- ccnreduce the containment pressur~e totermin~ating the leakage.umned-for the expected case analysis retain somenenRt hea;tr:emov~al systems are designed to;pheric following the initial pressure rise, thus15.5.17.2.4 Containment Leakage Exposure Sensitivity StudyLJIT-rTDI'AI IKIzC'DhTAAIt(KI IK ITAI 1CýQ OEI nIA1 KInT DIC'iI IIDf1 l TCt Dl\/I1CCSens&tivty studies wcro peiformed to ilustrato the dependencc of the thyroid ex-posuresOn th9 Spray system remova! cOnistant and the fraction of nonremgovable iodines presentin theo containment. The results or tnese stua~es are shown 0n -:ur-es 1&5 &, 557-f,and 15.5 8. The thyrid exposure., normal.zed to the .xp.sur.e for zero spray removalconstan9,,,,,00,0,t, A ho, fnin of .spray eo6ntant in F..i.. e 15.5 6, fo a fied nof normvboidine forms of 15 percGent. In Figur-es 15.5 7 an 155 tyrithe percGent of nonremovable iodines. To determine an absolute ex-posure (rem) fromFigure 15.5 7, the normalized ex-posure should be multilid by 940.9 rem, which is ME~reference 2 hour- 800 meter exposure for the design bai cas wi--- th a zero spray'constan-9t and a zero nonremovable fraction. To9 determine an absolu1te exposure (rem)from Figure 15.5 8, the normalized ex-posure should be m~uktilid by 197-.4 rem, whichis theQ reAferene 30 day 1 0, 000 meter exposure for- the design baisis case with a zerospray constant and a zcro nonremovable fraction. As shown in these figures,coPmbinations, of thege parameteQrsq that resuwlt in normalized exposures below thecriero lin UA~d resul0PQ699t in4 a- calculated absolute exposur-e less than the 300 rem.t15.5-65Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEguidaine leve! spe.ified in 10 FR P124 100.15.5.17(.2. ladietoqicaI Genseciuences wltf UI- OT 1-0UThe design basis LO CA was reviewed to evaluate potential differences in the offsiteradiological dose conRsequences us-ing a containment decontamin*ation factor Of 100 anda containment mnixing flowrate of 94,000 cfmA.A containment riin ate of 9400 cf orreSPOnds oih ur current Mim design-basis opwatiorn f .O contai;nmeRt ce-,,'r unit- (CFCU). CalculatioR were basedOn reload fuel. A containment spray delay of 80 seconds was used. The radionuclideinventor,' source termns for the various fuel conditions were calculated ueing thORIGEN 2 comnputer code with a power level of 3580 MINt. The radionuclide-atmospheari reale6tse and- offsite doerRs,- weire c-alc'ulated- with the IOCADOSE computerGede.Calculations .. ere..ade relative to 10 CFR 100.11 requiements for.ff.site doses, at the800 mneter e)xclusion area boundar; (EAB3) at 2 hoUrs and the 10,000 meter lopopulation Zone (LPZ) at 30 days, fm os .leakage.Table 15.5 75 Ipresents- the calculWated offsite dose consequences from post LOCA fromnvarious pathways. The limiting dose is the thyroid at the EAB. ForF the containm~entleakage pathw,,'ay, the maximum thyroid dose- of rem ev.. ceeds the origiRal designbasi, LOCA thyrod dose of 95.9 Frm in Table 15.5 23. For the p-e ex";ting smalllekgtier=M tHFl uiywiu 1ue .6 remF. Thiese uuese aie uumpalable witl tIAcorresponding oiginal design basis LOCA large leakage and small leakage casesAll doses are within the 10 CFR 100.11 15.5.17.2.6 Offsite Exposures from Containment ShineThe site boun~dar; 30 day DB3A exposure froem direct containment gamma radiation(containm~ent shine) is estimated to be 0.0048 rem. Containment shine is a funcati-on ofthe civt present in the conta-inment a~tmostne-phere. The EMVERALD com~puter code wasused to caIcUlate the post accident containment activity time histor,', and theISQSHLD 11 computer code was- then used to calculate the conta;inment shine exposure.The shin exosr model assumes a cylindrical radiation; source havin~g the sameradius, adh ighs the containment structure with a 3.5 foot thick concrete shie-ldsurroundinig i.The site boundary receptor point is assumed tbe800 meters fromn thecontain~ment struc-ture.15.5.17.22.7 O-ffreite Population Exposrmes from Containment LeakageT-hp calculated neaulation exiosgureis for the design bas eas asumotions. and for thee.pected case, are summarized in Table 15.5 223. These whole body population15.5-66Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEexposures do not include the effets of any population redistribution due to evacuation.Tho-se exposur~es were caiculated using the EMVERALD computer code. Theatmospheric dilution factors and population distribution utili;zed inl the populationFe xpo9s uire calcu-latiioens ae discussed i S;ct io 15.5.5.155.7..8Offsoite EXPosures fromA Post LOC,_A. RecGircUlation Loop Leakage in theAuxiliar,' Building Reactor coolant water that collects in the containment recirculationsump after a LOCA would contain ra;dioactive fission pr-oducts.15.5.17.2.2.3ESF System Leakage Outside ContainmentThe fluid that collects in the containment recirculation sump after a LOCA (i.e., the fluidcontents of reactor coolant system, the RWST, the NaOH tank and the accumulators)contain radioactive fission products that has been released from the core as a result ofthe LOCA. Because containment sump water *i ..cicuated outside thecontainment, problems of potential exposure due to post LOCA operation of externalcirulation loops with leakage have been evaluated.Reactor coolant water, EGGS injection water, and containment spray water accumRulatei n the containment rFe;ircuation ,um.p following a LOA. .Containment The containmentrecirculation sump water is circulated by the RHR pumps, cooled via the RHR heatexchangers, returned to the containment via the RHR system piping and the CSS piping-(if recir.culation spray is used), passed through the RCS and the containment spraynozzles (if spray is used), and finally returned to the containmentrecirculation sump. In the event of circulation loop leakage in the auxiliary building,post-LOCA activity has a pathway to the atmosphere.An illus6tration of this pathway for a small leak is given in F.igue 15.5 9. For the s.allleakage situatieo, fisRion poducts in the leakage water are exposed to auxilia" buildingventilation aif flow for a long period of time. Thus, for the smina leakage situation, allaSctivont released to- the auxiliary building would be released to the auxiliar' building air,i.e., n credit foR liquid gas paGutitioenig.An illus6tration of post LOCGA acti'Vity pathway for a large leak is given inFigure 15.5 10.For the large leakage situation, fission products in the leakage water are exposed toauxiliar; building ventilation air flow for a short period of time. Thus, modt of the aptiViyreleased to the auxiliary building would be transferred to the floor: drain reGeiver tank,i.e., credit for liquid gas partitionIfing.The complete RHR system and OSS description, including estimates of leakage,detection of leakage, equipment isolation, and corrective maintenance, are contained inSections 5.5.6 and 6.2.2, respectively.In accordance with Regulatory Guide 1.183, with the exception of noble gases, all thefission products released from the core during the gap and early in-vessel releasephases are assumed to be instantaneously and homogeneously mixed in the primary15.5-67Revision 19 May 2010 IDCPP UNITS 1 & 2 FSAR UPDATEcontainment recirculation sump water at the time of release from the fuel. A minimumsump water volume of 480,015 gallons is utilized in this analysis.In accordance with Regulatory Guide 1.183, the ESF systems that recirculate sumpfluids outside containment are analyzed to leak at twice the sum of the administrativelycontrolled total allowable leakage applicable to all components in the ESF recirculationsystems. With the exception of iodine, all radioactive materials in the recirculating liquidare assumed to be retained in the liquid phase.ESF leakage is assumed to occur at initiation of the recirculation mode for safetyinjection. Since the maximum temperature of the recirculation fluid supports a flashfraction less than 10%, per Regulatory Guide 1.183, ten percent (10%) of the halogensassociated with this leakage are assumed to be airborne and are exhausted (withoutmixing and without holdup) to the environment. The iodine release from the core is 95%particulate (Csl), 4.85% elemental and 0.15% organic, however after interactions withsump water the environmental release is assumed to be 97% elemental and 3%organic.The environmental release of ESF system leakage can occur via the 2 pathways listedbelow.a. Environmental release of ESF System leakage via the plant vent: The sum ofthe maximum allowable simultaneous leakage from all components in the ESFrecirculation systems located in the auxiliary building is limited to 120 cc/min.Thus, and in accordance with the requirements of Regulatory Guide 1.183, theanalysis addresses an ESF leakage of 240 cc/min in the auxiliary building. Theareas where these components are located are covered by the PG&E DesignClass I ABVS which discharges to the environment out of the plant vent. Onlyselected portions of the Auxiliary Building ventilation system are processedthrough the PG&E Design Class I AB ventilation filters. For purposes ofestimating the dose consequences, it is assumed that with the exception of theRHR pump rooms (refer to Section 7.2.3.4), this release pathway bypasses thePG&E Design Class I AB ventilation filters.b. Environmental release of ESF System leakage via Containment Penetration AreaGE and Areas GW & FW: The sum of the maximum allowable simultaneousleakage from all components in the ESF recirculation systems located in thecontainment penetration areas is limited to 6 cc/min. Thus, and in accordancewith the requirements of Regulatory Guide 1.183, the analysis addresses an ESFleakage of 12 cc/min in the containment penetration areas. The ventilationsystem covering this area is not PG&E Design Class I, thus the release path tothe environment is unfiltered and could occur via the Plant Vent or via the closeststructural opening in the Containment Penetration Areas GE and Areas GW &FW.15.5-68Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATEPost-LOCA aux!!iar,' building loop leakage exposures were calculated for four differentleakage Gases-(1) Expected small leakage case(2) Expected large leakage case(3) DBA small leakage case(4) DBA large leakage caseAssumption~s and numerical values used to calculate loop leakage exposures are listedin Table 15.5 21. Table 15.5 63 shows the rcsults of the calculations based On theseassumptions. Becaus an nsgnificant amount of noble gases would be in thecontainment recirculation sump water, the whole body exposures are negligible.One possible to therecircUlatien loop leakage wo.!evaluation of offsite exposures from postL AId include the following assumptions:0+1A LOCA. as an initiating event(2) F~ailure of twvo EGGS trains resulting in gross fuel damage: Releasse of50 percent Of core iodine inventor'; and 100 percent of core noble gasinventor', to the coentainmentA(3) F'ailure of an RHR pump seal, resulting in the release of a significantamount of the above containment activity to the auxilial' building(4) Failure of the passive auxiliar,' building charcoal filters resulting in theunfiltered release of iodine fission products to the enV;rFnm,ntThe assumption of this sequence of fa*ilures for analysis of offsite exposues, however,would be requiring plant design fe-atures in excess of the current guides and regulations,and in particular the requirements of ANS Standard N1 8.2, Nuclear Safety Criteria forthe Design of Stationar,' PreSSUrized W~ater Power Plants. (See proposed addendum toANS Standard N18.2, Single Failure Criteria for Fluid Systems (Reference 16))-.Applying the proposed standard to post LOCA recircu'ation loop leakage the L13LOCA.wasg assumed- ;as the initiating event:"The unit shall be designed to tolerate an initiating ev.ent vwhich may be a singleactive or passive failure in any system intended- for use durling normnal operation."The EGGS was assumed to function properly, as required by the EGGS acceptancecriteria, preventing gross fue! damage. Although meeting these criteria is expected toprec'ude gross cladding damage, it was assumed for this analysis that 100 percent of15.5-69Revision 19 May 2010 IDCPP UNITS 1 & 2 FSAR UPDATEthe gap iodine and noble gae, inventories were released to the containmnent reircGulationFor the !argo leakage cases,15.5.17.2.2.4 RHR Pump Seal Failurefailki.e-Failure of an RHR pump seal was assumed to be as-the worst case single failureandG to be tolerated without loss of the required functioning of the RHR system, aswas required by the following clauses in the pr-pesed-addendum to the ANSStandard N18.2 proposed at the time of original license:"Fluid systems provided to mitigate the consequences of Condition III andCondition IV events shall be designed to tolerate a single failure in addition to theincident which requires their function, without loss of the function to the unit."A single failure is an occurrence which results in the loss of capability of acomponent to perform its intended safety functions when called upon. Multiplefailures resulting from a single occurrence are considered to be a single failure.Fluid and electrical systems are considered to be designed against a singlefailure if neither (a) a single failure of any active component (assuming passivecomponents function properly); nor (b) a single failure of a passive component(assuming active components function properly) results in a loss of the safetyfunction to the nuclear steam electric generating unit."An active failure is a malfunction, excluding passive failures, of a componentwhich relies on mechanical movement to complete its intended function upondemand."Examples of active failures include the failure of a valve or a check valve tomove to its correct position, or the failure of a pump, fan or diesel generator tostart."Spurious action of a powered component originating within its actuation systemshall be regarded as an active failure unless specific design features or operatingrestrictions preclude such spurious action."A passive failure is a breach of the fluid pressure boundary or blockage of aprocess flowpath."For the expected and .1BA Iarge leakage ca.es, iheThe failure of auxiliary buildingcharcoal filters, a second failure, was not assumed, in accordance with the standard.For the expected and DBA small leakage cases, failure of auxilia',' building charcoalfilters was assumed as the single failure and can be tolerated without loss of the-reuie fUncGtion Of the auxiliar,' building ventilation system, which provides cooling for15.5-70Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEEGGS rcompone nt.For the longterm .m-,a! leakage cases, the charcoal fiet e "rc t needed to reduceexposres below the guideline values given in 10 CFR Pa~t 100. in an" carse, the fansin the ventilation system We redundant, and only the passive charcoal beds themSe.Vesare not redundant.Fer the experted small and Iarge leakage cases, it was that two EGS tra;ins,five fan coo)Icrs, and bNo containm~ent Spray trafins functioned. For the IDBA small andarge leakage cases, it is assu-med that two EGGS trains, two fan coolers, and onecontainment spray train functioned. The DBA assumptions result in high recirculation sump water temperatures and minimum containment recirculationI sumpFor all four c;rculation loop leakage cases it was assumed that 100 peFrent of the gapio-dine ..vente. ' was- deposited iW Glo+ewn conanentreiruatio sump water.ForF the expected small and large leakage cases, the assumed gap iodine inventoriesare listed- in Tab;-le- 11.1 7. The expected case gap iodine was assumed to be onlyelemental iod-ne. For the IDBA small and Iarge leakage cases, the asrmed gap iodinReinveRtWries are based An release fract'Rns given in Safety Guide 25, March 1972(Reference 23). The nBA case gap iodiRe was assumed to be 99.75 percent elemental:odine and 0.25 percent organic iodine per Safet Guide 25, March 1972.Radiological decay of activity in the containlment recirculation sump was assumed for allleakage cases for both the time periods before and during loop leakage. No credit wastaken for cleanup of activity in the containment recirculation sum~p.ReacGtor oolant water, accum+ulator water, and refue;ng water storage tank (R. ST)water up thn total volume of water 4n which activity is deposited. Consideration ofemergency core cooling injection flowrates and containment spray injection flowmrte6yields the volume of RAST water (Chapter 6). Table 15.5 21 lists the assumed volumeof which activity is deposited for the four leakage cases. For the large leakageases, the voelume of diluting water wars taken as the volume when the leakage began.NO credit was taken for the extra diluting water added from the RV.A.ST duFrig the30 minute leakage period.Sodium hydroxide spray additive will provide for an increased pH in the containmentrecirculation sump water. Consideration of emergency corecooling injection f1oW2atesand containment spray injection flOWrates yields the pH4 of the containment Frecirculationsump water (Chapter 6). Table 15.5 24 lists the assumed pH Of recircU~ation loopleakage water for tht e, for leakage carses. For the !arge leakage cases, the pH;Awas;taken athpHwen the leakage began. No credit was taken for the extra sodiumnh~ydroxide in the spray water added during the 30 minute leakage period.The design evaluation conducte-d foAr the onanmn f~unc-tionRal design yields the15.5-71 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEtemperaturc of containment recircUlation sum~p water as a function Of timne (Chapter 6).Tab;-le- 15.5 24 lists the assumed temnperature of recirculatien loo)p leakage water forF thefour leakage cases. For the large leakage Gases, the water temnperature was taken asthe temperature when the leakage began. No credit was taken for the decrease[; ofwater temperature duFrin the 30 mninute leakage period.A review of the equipment in the RHR system loop and the CSS loop indicates that thelargest leakage would result from the failure of an RHR pump seal. Evaluation of RHRpump seal leakage rate, assuming only the presence of a seal retention ring around thepump shaft, shows that flows less than 50 gpm would result (refer to Section 3.1 andChapter 6). Circulation loop piping leaks, valve packing leaks, and flange gasket leaksare much smaller and less severe than an RHR pump seal failure leak. Leakage-fr,,.,these components during normal post LOCA operation of the RHR system loop and theCSS loop is estimatedto be, !910 cc/hr (Chapter 6). On this basis, a 50 gpm leakratewas assumed for both the expected larg leaag. e. cas and the DBA largeleakageLOCA. case, and a 1910 cc/hr leakrate was assumed for both the expectedsmall leakage case and the DBA smnall leakage case.For the DBA large leakage case, recirculation loopLOCA pump seal leakage wasassumed to commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of the LBLOCA. This assumption isconsistent with the discussion in Sections 3.1.1.1 and 6.3.3.5.3, and with the guidancein Standard Review Plan 15.6.5, Appendix B. In this context, the limiting recirculationloop long term passive failure is 50 gpm leakage at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of theLBLOCA.Evaluation of an RHR pump seal failure shows that the failure could be detected andthe pump isolated well within 30 minutes (Chapter 6). A-Thus a leakage duration of 30minutes is conservatively assumed for both the expected and the DBA LOCAlafge-leakage-Gases.15.5-72Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEA lea.kage 'o" days is assumed , both 'he e".e.ted and DBA sma"l l.ak.,,cases. As diScussed earler, the auxiliary bufilding DF is a function of the PartitioFactor (PF=) for a particular isotope (Equation 15.5 7).For both the expected and DBA large leakage cases, it was assumed that leakagewater was.... pupe away to the ,1,,,, ,4,.,;, receiver tank. l",4-ime in- he, leakage ......ewas assumed to be exposed to auxiliarl buid " entio air flow for a short period oftime (0.05 0.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />), and thus, liquid gas partitioning w~as assumed for elementaliedne isetepes.The leakage case elemental PFs were using the perViousYhpresented PF eXperession. Because the circu"lation water WAill be above 212OF(Chapter 6), a flashing process .musqt be cosd red.Fo heat energy conSer~atien onthe basis of 1 Ilahro =h, (I x) _(Rearranging yieldshf0 -hf (15.5 12)h9 -hfwheFel-initial enthalpy of liquid, Btullbmhl -final enthalpy, of liquid, Btullbmh9 -final enthalpy of vapor, Btuilbmnx -fraction of initial mass that became vaporon this temprature. The mass fraction, X, is the ratio of the nal mass of vapor to thetotal initial mass of water, so the mass ratio at the end of the flashing process becomes-Mvapor xMliquid 1-x(I5.6 13)Figures 15.5 11 and 15.5 12 present the expected and DBA large leakage caseelemental iodine PFs as a function of both temperature and pH. For small PFs, the DF(see Equation 15.5 7) is approximately equal to the reciprocal of the PF.Figures 15.54 11 and 15.5 12 illustrate that auxiliay building iodine PFs and resultingD)Fs are relatively insensitive to water temperture, but muc;h mor~e sensitive to pH.Tab;-le- 15.5 24 lists the assumed temperatures and pHs along with the resulting-elemental iodine PFs and auxiliary building decontamination factors for both the-expetdandIge a~e eakage Gases.15.5-73 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFor both the expe..ted. anmd DB1A small leakage cases, it was a.sumed that leakagewater was not pumped away. Elemental iodine in the leakage water was assumned tobe exposed to auxiliary building ventilation alir floew foar a- long period of time(1 00, 150 heUrS), and thus, liquid gas partitioning for elemental iodine isotopes w'as notassumedl. For the small leakage ease all elemental iodine actiVity released tothauxiliary building was assumed to be released to the auxiliar,' building atmosphere, i.e.,a' _Q F -- f 41leakage cases. All organic. iodine activity released to the auxi!i;;n building was-assumned to be released to the auxiliary building atmosphere, i.e., a decontamination;faotOF f-ef1For all fouFr lOP leakage cases, no Gcedit 1 taken for auIVI1ia1y building radiologicaldecay Or fission product plateout.Foar the expected and 96A large cases, credkit fo auxiliary buildin~g charcoal filters wasdecn -+ ..... " ....a + .. .. .+la9---I.nnn*L 'r~. nr~na ml, Ae., onA.anl rnarnr~n n ANIIEl Standard N18.2 single failure criteria). Table 15.5 24 lists the assumed iodine filterefficiencies, for each loop leakage case.From the cGalcuated D1BA case offsote exposures from post LOCA recircu+ation loopleakage in the auxiliary building listed in Table 15.5 63, it ran be concluded that anyexposures that occur via this combination of unlikely events would be well below theguideline levels in 1 0 CFR Part 100. In addition, even if no consideration is given to theeffectiveness of the auxiliay buildinHg charcoal filters for the DBA leakage cases, thearculated exposures would- still be below guideline levels specified on 10 CFR Part 100.15.5.17-.2.8.1 Maximum Allowable Leakage From Post LOCA Recirculation LoopCalculations have been pe~fGFmed to determine the maximum allowable leakage fromn.y.rfl~1(I%~ ~I PAL-------** r-'- '.i-JepeFatiens beTGFe 9"Ite ant, GeRfl-191 FROM uplot-atut- 0-5 -at Wt$UA i f-%f% nncc A A *.basis EAB and low population zone outer boundar,' (L=PZ7) off-site radiation doses andcontrol room operator airborne radiation dose from porst 10 LC--A containm~ent leakage,RHR pump seal leakage, and pre existing leakage from recirculation loOP componentsoutside cnotainment. The calculations determined the amontu of pre existingrecirculation leakage Which could exist before effsite exposures would excee10 CFR 100.11 limi~ts OF control room operator exposr~es would exc-eed GIDC 19, 1971limiOts, if a LOC.A w..ere to simultaneously occur.Table 15.5 63 shows the results of the calcaulations based on the above assumptionswhic~h determined that the maximum allowable leakage (in; additionA to the RHR pumpseal leakage) from the Frcirculation loop at post LOCA conditions of pressure andl15.5-74Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEtemperature was 1.85 gpmA w here the airborne activity is filtered by cha~rca! filters Or0..16 gp wer the airboFre activity is unfi!tered. The limitation is the GDC 19, 1971allowabe dse for the control room.In summary, the RHR pump seal failure resulting in a filtered release via the plant ventis DCPP's licensing basis with respect to the worst case passive single failure in theRHR system. Therefore, the RHR pump Seal Failure is retained as a release pathwayfor the AST LOCA dose consequence analysis.The activity transport model is based on a 50 gpm leak of sump water activity for 30minutes that occurs 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA. The temperature of the recirculation fluidis conservatively assumed to remain at the maximum temperature of 259.90F. Thus asdiscussed above in Section 15.5.17.2.2.3 under ESF system leakage, the amount ofiodine that becomes airborne is assumed to be 10% of the total iodine activity in theleaked fluid.The ventilation exhaust from the RHR pump rooms is covered by the PG&E DesignClass I Auxiliary Building ventilation system and processed through the PG&E DesignClass I AB ventilation filters. Thus, credit for filtration of the release of a RHR pumpseal failure by the Auxiliary Building Ventilation system is taken in determining thedose consequences to the public at the EAB and LPZ, to the operator in the controlroom, and to personnel in the technical support center.Credit for filtratin of therelease of a RHR System pump seal failure by the auxiliar,' building vecntilatiOns6ystemnis taken in determining the dose consequences to the public at the EAB and LPZ,cOntrol room and ,TSC.The efficiency of the auxiliary building charcoal filters is determined usingmethodology similar to that documented in Section 15.5.9 for the CRVS Mode 4ventilation filters. The allowable methyl iodide penetration / filter bypass for theauxiliary building charcoal filter is controlled by DCPP Technical Specification 5.5.11;and are 5% and <1%, respectively. Based on the above, an efficiency of 88% isassigned to the charcoal filters in the AB ventilation system prior to environmentalrelease via the plant vent. Similar to the ESF system leakage, the environmentalrelease of iodine is assumed to be 97% elemental and 3% organic.15.5.17.2.2.5 Refueling Water Storage Tank Back LeakageThe safety injection and containment spray systems function to provide reactor corecooling and mitigate the containment pressure and temperature rise, respectively, in theevent of a LOCA. Both systems initially take suction from the RWST. Once the RWSTwater supply is depleted, both the containment spray and safety injection systems aresupplied by the RHR System. The RHR pumps take suction from the containmentrecirculation sump water. Under LOCA conditions, the recirculation sump water isassumed to be radioactively contaminated by fission products, of which the maincontributors to airborne dose are the various isotopes of iodine.15.5-75Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEAs discussed in NRC Information Notice 91-56, September 1991 during containmentsump water recirculation, there is the potential for leakage from the mini-flowrecirculation lines connecting the high head and low head safety injection pumpdischarge piping to the RWST. Since the RWST is vented to the atmosphere, thispresents a pathway for iodine release to the atmosphere. The acceptance criteria in theDCPP administrative test procedures ensure that the total as-tested back leakage intothe RWST from the containment recirculation sump is less than or equal to 1 gpm.Dose consequences of RWST back-leakage assumes that leakage starts at theswitchover to recirculation following the LOCA and continues for 30 days. Perregulatory guidance, a safety factor of 2 is applied to the leak rate, i.e., a 2-gpm leakagerate is assumed for the full duration of the event, which is two times the allowableleakage of 1 gpm. With the exception of noble gases, all fission products released fromthe fuel to the containment are instantaneously and homogeneously mixed in the sumpwater at the time of release. Only iodine and their daughter products are releasedthrough RWST back-leakage since the particulates would remain in the sump water.A significant portion of the iodine associated with sump water back-leakage into theRWST is retained within the RWST fluid due to the equilibrium iodine distributionbalance between the RWST gas and liquid phases. The time dependent iodine partitioncoefficient takes into consideration the temperature and pH of the RWST liquid andsump fluid, the RWST liquid and gas volumes, and the temperature, pH and volume ofthe incoming leakage. The iodines that evolve into the RWST gas space as a result ofthe equilibrium iodine distribution balance, and the noble gas daughters of iodines, arereleased to the environment via the RWST vent, at a vent rate established by thetemperature transient in the RWST (which includes the effect of decay heat), theincrease in the liquid inventory of the RWST due to the incoming leakage, the gasesevolving out of incoming leakage, and the environmental conditions outside the RWST.The average time-dependent RWST iodine release fractions along with the fractionalRWST gas venting rates (may be applied to the noble gas daughters of iodines) to theatmosphere from the Unit 1 and Unit 2 RWSTs due to RWST back-leakage followingswitchover to the sump water recirculation mode of operation is summarized in Table15.5-23C. As discussed earlier, the release fractions / rates presented in Table 15.5-23C reflect a safety factor of 2 on the leak rates, i.e., are developed based on a RWSTback-leakage of 2 gpm. The iodine released to the environment is assumed to be 97%elemental and 3% organic.The equilibrium iodine concentration in the RWST gas space utilized to develop Table15.5-23C is based on the iodine mass in the sump fluid entering the RWST vapor spaceas back-leakage or the total iodine mass contained in the RWST liquid, whicheverresults in higher RWST vapor phase concentrations. The RWST maximum venting rateaveraged over an interval is primarily based on RWST back-leakage entering the RWSTgas space and thermally equilibrating, and is used in conjunction with the higher RWSTgas space iodine concentration to calculate an iodine mass release rate as a function oftime. An interval based averaging approach is utilized in preparing Table 15.5-23C to15.5-76Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEreduce the number of input values to the dose analysis while preserving the boundariesfor the time periods used for atmospheric dispersion; the actual iodine releasecalculated in an interval is normalized to the iodine mass leaking into the RWST duringthat time interval.Examination of the average gas space venting rates indicate that after the first day, thenoble gases formed by decay of iodine will primarily remain in the RWST during the 30day period of evaluation and not be released. However, the dose consequenceanalysis conservatively releases the noble gases formed by decay of iodine, directly tothe environment without taking any credit for tank holdup.15.5.17.2.2.6 Miscellaneous Equipment Drain Tank (MEDT) LeakageThe DCPP Unit 1 and Unit 2 MEDT is a covered rectangular stainless steel linedconcrete tank located in the auxiliary building below El 60 foot. The MEDT tank vent ishard-piped to the auxiliary building ventilation ductwork; thus the airborne releases fromthe MEDT are ultimately discharged to the environment via the plant vent (refer toSection 9.4.2).Following a LOCA, the MEDT will receive both post-LOCA sump fluids as well as non-radioactive fluids (i.e., ESF system leakage from the accident unit as well as non-radioactive fluids from equipment drains / RWST leakage from the non-accident unit)which are hard-piped to the MEDT. The acceptance criteria in the DCPP administrativetest procedures ensure the total as-tested flow hard piped to the MEDT is less than950 cc/min of ESF system leakage and 484 cc/min of non-radioactive fluid leakage.Similar to the RWST back-leakage model, dose consequences due to releases from theMEDT assumes that leakage starts at the switchover to recirculation (829 secondfollowing the LOCA) and continues for 30 days. Per Regulatory Guide 1.183, a safetyfactor of 2 is applied to the leak rate, i.e., 1900 cc/min of ESF system leakage and 968cc/min of non-radioactive fluids into the MEDT is assumed for the full duration of theevent, which is two times the allowable leakage. With the exception of noble gases, allfission products released from the fuel to the containment are instantaneously andhomogeneously mixed in the sump water at the time of release. Only iodine and theirdaughter products are released through MEDT leakage since the particulates wouldremain in the sump water.The methodology used to determine the post-LOCA iodine and noble gas releases viathe MEDT vent and Plant Vent is similar to that used to address RWST back-leakage.Adaptation of the methodology to address overflows/room ventilation releases isstraightforward with the room ventilation rate being treated as the tank exhaust rate.The transport model utilized to determine airborne releases from the MEDT takes intoaccount the fact that the MEDT is a small tank with an auto-transfer capability which isPG&E Design Class II. Consequently, and for purposes of conservatism, it is assumedthat a) the LOCA occurs when the MEDT water level is at the normal maximum setpoint15.5-77Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEto initiate auto transfer, b) the auto-transfer capability is not initiated because it is not asafety function, and c) the MEDT contents will spill over into the Equipment DrainReceiver Tank (EDRT) Room after the tank is full. Thus, for the post-LOCA scenario,the MEDT is conservatively assumed to overflow via its manway into the EDRT Room.The EDRT room drains into the auxiliary building sump, which ultimately overflows intothe Unit 1/Unit 2 pipe tunnels. The auxiliary building sump is also a covered stainlesssteel lined concrete tank with a vent that is hard-piped to the auxiliary buildingventilation system (ABVS) with a PG&E Design Class II auto transfer capability. Theauxiliary building sump is located adjacent to the MEDT.The bounding transient release of iodine along with the gas venting rate to theatmosphere as a result of post-LOCA leakage of radioactive and non-radioactive fluidhard-piped into the MEDT is developed in 2 parts: a) prior to MEDT overflow and b) postMEDT overflow.a) Prior to MEDT overflow -The iodines evolve into the MEDT gas space as a result ofthe equilibrium iodine distribution balance between the MEDT gas and liquid phases(either the MEDT liquid inventory or the incoming leakage), and are released to theenvironment via the plant vent, at a vent rate established by the temperaturetransient in the MEDT (including the effect of decay heat), the increase in the liquidinventory of the MEDT due to the incoming leakage, and the gases evolving out ofthe incoming leakage.b) After MEDT overflow -The equilibrium iodine distribution balance is conservativelyassumed to be between the iodine concentrations in the MEDT overflow liquid andthe EDRT room (or Unit 1/Unit 2 pipe tunnels) ventilation flow (rather than theaverage concentration in the EDRT room (or Unit 1/Unit 2 pipe tunnels) freevolume). This maximizes the iodine release rate. Thus, the iodines released are asum total of the following:i) the iodines that evolve into the EDRT room air space as a result of theequilibrium iodine distribution balance between the spilled liquid from theMEDT (at the temperature of the MEDT), and the EDRT room ventilation flow,and is released to the environment via the plant vent, at the vent rateestablished by the EDRT room ventilation system, andii) the iodines that evolve into the Unit 1/Unit 2 Pipe Tunnel air space as a resultof the equilibrium iodine distribution balance between the spilled liquid fromthe MEDT (at the maximum temperature of the Unit 1/Unit 2 Pipe Tunnel),and the U1/U2 Pipe Tunnel ventilation flow, and is released to theenvironment via the plant vent, at the vent rate established by the U1/U2 PipeTunnel ventilation system.The exhaust fans servicing the EDRT room and pipe tunnel are PG&E Design Class I.There is also a potential that the non-LOCA unit's ABVS will be operating with the flowexhausting to its unit specific plant vent. Thus, it is conservatively assumed that the15.5-78Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEnon-LOCA unit's ABVS is also operating, and together with the accident units' exhaustfans, are providing the motive force to exhaust the airborne releases to the respectiveunit vents.The average time-dependent MEDT iodine release fractions, along with the fractionalMEDT gas venting rates (which may be applied to the noble gas daughters of iodinesprior to MEDT overflow) to the atmosphere following switchover to the sump waterrecirculation mode of operation, is summarized in Table 15.5-23D. As discussedearlier, the release fractions / rates presented in Table 15.5-23D reflect a safety factorof 2 on the leak rates, i.e., are developed based on an input of 1900 cc/min of ESFsystem leakage and 968 cc/min of non-radioactive fluids into the MEDT. Through theuse of extremely conservative assumptions, the calculated iodine release fractions / gasventing rates presented in Table 7.2-4 when used in combination with the analyzed ESFsystem leak rate, bound the iodine releases of all combinations of radioactive and non-radioactive leakages less than or equal to the leak rates analyzed. The iodine releasedto the ventilation system is assumed to be 97% elemental and 3% organic, and isreleased to the environment via the plant vent. In addition, the dose consequenceanalysis conservatively releases the noble gases formed by decay of iodine, directly tothe environment without taking any credit for tank holdup.15.5.17.2.3 Offsite Dose AssessmentDue to the delayed post-LOCA fuel release sequence of an AST model, and the rate atwhich aerosols and elemental iodine are removed from the containment, the maximum2-hour EAB dose for a PWR LOCA typically occurs between 0.5 hrs to 2.5 hrs.To establish the "worst case 2-hour release window" for the DCPP EAB dose, theintegrated dose versus time for each of the six pathways discussed in Section15.5.17.2.3 was evaluated. The 0-2 hr EAB Atmospheric Dispersion Factor from Table2.3-145 was utilized for all cases.The analysis demonstrated that for DCPP the maximum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose will occur, asa result of the RHR pump seal failure, between T=24 hrs to T=26 hrs, and is unrelatedto the post-LOCA fuel release sequence associated with AST.The direct shine dose at the EAB due to a) the airborne activity inside containment,and b) the sump water collected in the RWST due to RWST back-leakage, was alsoevaluated. Based on the results of the EAB evaluation which determined that thedose contribution due to direct shine was minimal (<0.01 rem), the dose at the LPZdue to direct shine is deemed negligible.The bounding EAB and LPZ dose following a LOCA at either unit is presented in Table2.3-145.15.5.17.2.9 Off:'te EXOc.Ur.S from Controlld Post a,,dcnt Centa.nm....15.5-79Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEVe nt-0 n,pi o -+ t sýn;f al.tntfo nfn.n ,n. ne .rr nn r 3t pggnrn l.trl^j~t, fUhUl.%5 S.* t~ttii tR lt tSt i .jAP5i i- ..;A -FIlLt...... ...... ....... ... a ............. ...... .-...s .s .. .........control"ing the post accident concentration of hydrogen in thc containment atmosphere-Redlundant theFrmal hydFrogen recoMbiners are the pnrian; means of post acGcidnthydFrogen control. As a backup, controlled containment venting (via the containmenthydFogen purge system) with offshoe flow, wind directionS from no.,hwe s.througheat-, southeast measured p-rVides, hydFrogen contrl with a high pFebability ofno inland exposures. As shown in Table 15.5 26, offs-hore wind- d rections occur over50 percent of the time regardless of the season and, as shown in Table 15.5 27, have ahigh degFree of persistene. The large time period (312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> for rDBA case) betweenthe hydrogen venting level (3.5 vieo) and the hydrogen flammability leve(4.0 We) iS much greateFr than the longest reorFded period (37- consecutiie houFr) ofons-hore winds in any 22.50 sector. These-d-at-a ensure a velr; high probability thatventing can be arrFied out duFring the occurrenAe of ofshore winds.Even though there is a high prebability that cOntainmcnt venting can be carried outwhen the wind is blowing offshore, if at all, an evaluation is presented in thefollowing paragraphs to determine potential exposures if venting were .care out duringefdUHU 15.3.U bfUfiILtJIIRw LII", dlyb!b um pubt c2bblucl lytil W!5W@ I lit touUtAltol m M ltl-i r, 4. 44U lr -li A i tI u a.i.... -.......... -..t .... -........... ...it- ............... ...- t...also described in Section 6.2.5. The purge stFeam is withdrawn from the containmentthroeugh one of te penetFration ines. The stream is rotcd thromugh a flow device, chaFGoal filter, exhaust fans, the radiation monitors, and finally to the plantPost accident containment venting activity releasos arc caIGLuated with the followingequatieni[1.0 -0.01FILEFF(l)]60 f T(2) \/FNIpATYA('(I't-JX(l)t11t.I I I IVULUML-T(1)wheFe-.ACT(!) -,AG(l)-. activity of istp F eleased to the containment atmosphere, CVOLUME -VENRAT -I(I) -removal onstAntO fAr isottop I, hr4f4 IEFFAIk ---4:.,.... \,.IC/' k fa.-f-is~tn , n a. n.n, hPitt,_r/rv -44...t 4.a. 1 ('r(A 4-1, f +afnn +n f;n~t tAo kiNLiii ~ ~ ~ ~ c is-l tcit ir.5Si i*f.I i.iiiiisii Is-r SituP t* 1 U- P .PJ15.5-80Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE~f-l^old III;auL pe!onThe above equation considers radiological decay during the time period priortcontainment venting and the time period during containment venting. It also assumesthat the L=BLOCA actiVity released to the containment atmosphere is homo~geneouslydispersed throughout the containment atmospheri volume. Exposures from activityreleascd to the atmosphere were calculated u6ing the EMERALD computer code.EMERALD assumes there is, no radiological decay duFrig the atmos~pheiric disprsonContainment venting exposure6 were calculated for both the expe ctd-- -case- awn.d th eIDBA case. Assumnptions- and numerical values, used to calculate Yenting exposures areitemized in Table 115.5-28. Onshore controlled coentainment venting thyroid and wholebody exposures are listed in Table 15.5 29.Post accident containment venting schedules, are evaluated in Section 6.2.5. Assmnthe venting system will operate an average 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per day, the systemn flowrates duFrig-sheot venting periods are 120 cfmn (expected) and 300 cfms (D)BA). E~quivalentcontinuous venting rates, 10 cfmA and 25 cfm, were used to calculate venting acativity~Feleases-in the event containment venting should be required during periods with onshore flow,the venting would be limited to those periods when Pasguill Stability Categor,' D exists.Categor; D and an elevated release height of 70 m~eters were evaluated using-a-coenventional Gaussian plume model and are listed in Table 15.5 30. Themneteorological input parameters utilized were detefrmined from onsite mfeasuremnents,given in References 18, 19, and 20. Because an individual is assumed to be located ethe plume centerline for the entire venting duration, exposures, are centerline exposureand represent worst case conditions. The probability of an individual being located onthe plume centerdine for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period is very small, and thurs centerline exposuresIiqbtrd "m T;IhlaIr F1' fi 29 nm "w' nng~~ghrB-During the time period prior to venting, activity released to the containment atmosphereis significantly reduced by both radigolgical decay and functioning of the safet features-systemsr,. The main contribut,,or,, Of several hundFed hours after thaccident are the noble gases: Kr 85, Xe 133, and, to some extent, Xe 131m. BecauseKr 5 hs ahalf li~fe of 10.6 years, the exposueres resulting fromn containment ventingwould not be significantly reduced if the venting could be fu~theF delayed for Man"m,'n~nho-It can be cncn;lude-d- from the results prresnted in Table 15.5 29, along with theconsideration of the very high probabillity of OPPOFWnitiees for offshore venting and theother favorable factors associated with the DCPP design an~d site, that, as a backup tothe i;ntrnal hydrogeAn systemr, ve-nting usiRg the GcntainRenthydrogen purge systemF is an acceptable contAngency methd of post accident hydrogencontro! for this plant. in addition, it can be concGluded that the expected exposures dueto venting e-e uing the assumnptions in Safety Guide 7, will not exceed the annual15.5-81Revision 19 May 2010