Text

Diablo Canyon, Units 1 and 2 - Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 1 of 3
	ML16004A361
Person / Time
Site:	Diablo Canyon
Issue date:	12/17/2015
From:	; Pacific Gas & Electric Co
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16004A363	List: DCL-15-152, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03 - Updated Final Safety Analysis Report Markup, Revision 1. Part 3 of 3; ML16004A354; ML16004A355; ML16004A356; ML16004A357; ML16004A361;
References
	DCL-15-152, TAC MF6399, TAC MF6400
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EnclosureAttachment 4PG&E Letter DCL-1 5-1 52License Amendment Request 15-03, Attachment 4Diablo Canyon Power PlantUpdated Final Safety Analysis Report Markup(For Information Only), Revision 1 DCPP UNITS 1 & 2FSAR UPDATEreleases of radioactive materials to the atmosphere and (2) coping with radiologicalemergencies.2.3.1.4 Safety Guide 23, February 1972 -Onsite Meteorological ProgramsAn onsite meteorological monitoring program that is capable of providing meteorologicaldata needed to estimate potential radiation doses to the public as a result of routine oraccidental release of radioactive material to the atmosphere and to asses otherenvironmental effects is provided.2.3.1.5 Regulatory Guide 1.97, Revision 3 -Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs ConditionsDuring and Following an AccidentControl room display instrumentation for use in determining the magnitude of therelease of radioactive materials and in continuously assessing such releases during andfollowing an accident is provided.2.3.1.6 Regulatory Guide 1.111, March 1976 -Methods for EstimatingAtmospheric Transport and Dispersion of Gaseous Effluents in RoutineReleases from Light-Water-Cooled ReactorsAnnual average relative concentration values are used during the postulated accident toestimate the long-term atmospheric transport and dispersion of gaseous effluents inroutine releases.2.3.1.7 Regulatory Guide 1.111, Revision 1, July 1977 -Methods for EstimatingAtmospheric Transport and Dispersion of Gaseous Effluents in RoutineReleases from Light-Water-Cooled ReactorsIn accordance with the requirement of Regulatory Guide 1.145, Revision 1 annualaverage relative concentration values are developed for each sector, at the outer lowpopulation zone (LPZ) boundary distance for that sector, using the method described inRegulatory Position 0.1 .c of Regulatory Guide 1.111, Revision 1. This information isused as input to develop the design basis radiological analysis 7./ values at the LPZusing Regulatory Guide 1.145, Revision 1 methodology.2.3.1.8 Regulatory Guide 1.145, Revision 1, February 1983 -AtmosphericDispersion Models for Potential Accident Consequence Assessments atNuclear Power PlantsThe method outlined in Regulatory Guide 1 .145, Revision 1, (with the exception ofmethodology associated with elevated or stack releases, i.e., Regulatory PositionsC.1.3.2, 0.2.1.2 and 0.2.2.2), is used for calculating short-term atmospheric dispersionfactors for off-site locations such as the exclusion area boundary or the low populationzone for design basis radiological analysis dispersion factors.2.3-22.3-2Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE2.3.1.9 Regulatory Guide 1.194, June 2003 -Atmospheric RelativeConcentrations for Control Room Radiological Habitability Assessmentsat Nuclear Power PlantsThe method outlined in Regulatory Positions C.1 through C.3, and the adjustment factorfor vertically orientated energetic releases from steam relief valves and atmosphericdump valves allowed by Regulatory Position C.6 of Regulatory Guide 1.194, June 2003is used to determine short-term on-site atmospheric dispersion factors in support ofdesign basis radiological habitability assessments.2.3.1.710 NUREG-0737 (Item III.A.2), November 1980 -Clarification of TMl ActionPlan RequirementsItem Ill.A.2 -Improving Licensee Emergency Preparedness-Long-Term:Reasonable assurance is provided that adequate protective measures can and will betaken in the event of a radiological emergency. The requirements of NUREG-0654,Revision 1, November 1980, which provides meteorological criteria to ensure that themethods, systems and equipment for monitoring and assessing the consequences ofradiological emergencies are in use, is implemented.Item III.A.2.2 -Meteorological Data: NUREG-0737, Supplement 1, January 1983provides the requirements for III.A.2.2 as follows:Reliable indication of the meteorological variables specified in Regulatory Guide 1 .97,Revision 3, for site meteorology is provided.2.3.1.811 IE Information Notice 84-91, December 1984- Quality Control Problemsof Meteorological Measurements ProgramsMeteorological data that are climatically representative, of high quality, and reliable inproviding credible dose calculations and recommendations for protective actions in anemergency situation, and for doses calculated to assess the impact of routine releasesof radioactive material to the atmosphere are available.2.3.2 REGIONAL CLIMATOLOGYHISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.2.3.2.1 Data SourcesThe information used in determining the regional meteorological characteristics ofDiablo Canyon Power Plant (DCPP) site consists of climatological summaries, technicalstudies, and reports by Dye (Reference 2), Edinger (Reference 3), Elford (Reference 4),2.3-32.3-3Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE22.5° interval. The 1 -year gap (April 1971 through March 1972) in the period of record,October 1970 through September 1972, resulted from an unauthorized bivanemodification.Frequency distributions of wind speed and wind direction classified into seven stabilityclasses as defined by the vertical temperature gradient are shown in Tables 2.3-21through 2.3-28. The column headings are labeled in terms of mean hourly wind speedin miles per hour. The six wind speed categories are as follows: 1-3, 4-7, 8-12, 13-18,19-24, and 25-55. The rows are labeled with the wind direction at the midpoints of 22.50intervals. Table 2.3-28 shows the number of observations in each of the seven stabilityclasses (Pas quill A through G) for the period of record July 1, 1967, throughOctober 31, 1969, when the mean hourly wind speed is less than 1 mph. The winddata were measured at the 76 meter level, and the vertical temperature differencemeasurements are the 76 meter level minus the 10 meter level.The radius of the low population zone (LPZ) at DCPP has been established to be 6miles. Cumulative frequency distributions of atmospheric dilution factors at each 22.50intersection with a 10,0O00-meter radius (slightly greater than 6 miles) for the period May1973 through April 1975 are presented in Table 2.3-4 1, Sheets 7, 8, 9 and 10. Eachdata set used to compile the frequency distribution is comprised of averages taken over1 hour, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months , 3 days, or26 days, using overlapping means updated at 1-hour increments as specified by the NRC.Because of overlapping means, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months z/Q is included in several observation periods:for example, an hourly J/Q is included in 624 estimates of the 26-day averages. As aresult, a single hourly measurement may influence the value of over 5 percent of theobservations. Since overlapping means are used in the distributions, the data are notindependent and no assumption of normality can be made. These data show z/Qestimates from the 25th through the 100th percentile levels for each of the averagingperiods.2.3.5.2 Design Basis Radiological Analysis Dispersion Factors2.3.5.2.1 Exclusion Area Boundary and Low Population Zone AtmosphericDispersion FactorsAtmospheric dispersion factors (i.e., x/Qs) are calculated at the EAB and LP7_ for post-accident environmental releases originating from Unit 1 and Unit 2. These 7/Qs areapplicable to all dose consequence analyses documented in Section 15.5 with theexception of the tank rupture events. The methodology used for the tank ruptureaccidents is discussed in Section 15.5.5.2 and the associated ylQs are reported inTable 15.5-3.The applicable methodology is identified in Regulatory Guide 1 .145, Revision 1(Reference 22). The methodology is implemented by executing the CB&l computerprogram "Atmospheric Dispersion Factors" EN-i113 (Refer to Section 15.5.8.10 for a2.3-282.3-28Revision 21 September 2013 DCPP UNITS 1 & 2FSAR UPDATEdescription of computer program EN-i113) using a continuous temporally representative5-year period of hourly meteorological data from the onsite meteorological tower (i.e.,January 1, 2007 through December 31,2011). EN-I113 calculates j/Q values for thevarious averaging periods using hourly meteorological data related to wind speed, winddirection, and stability class.Equations used to determine the %/Q's are as follows:X/Qi = + (A/2)]}1 (2.3-7)Z/Q2 = 1 (2.3-8)Z/Q3 = [ ( u) (7;) (X,) (Oz)]-1 (2.3-9)where:/Q = relative concentration (sec/in3);z= horizontal and vertical dispersion coefficients, respectively, based onstability class and horizontal downwind distance (in);u = wind speed at the 10-meter elevation (m/sec);A = cross-sectional building area (in2);; = (M)(ay) for distances of 800 meters or less; and: =[(M-1)(ary800rn) + for distances greater than 800 meterswith M representing the meander factor in Reference 22, Figure 3.Per Regulatory Guide 1.145, Revision 1, x/Q1 and z/Q2 values are calculated by EN-113 and the higher value selected. This value is then compared to the x/Q3 valuecalculated by EN-I113, and the smaller value is then selected as the appropriate value.The EAB distances for the sixteen 22.5°-azimuth downwind sectors are derived fromFigure 2.1-2, taking into consideration a 45-degree azimuth sector centered on each22.5°-azimuth sector as described in Regulatory Guide 1 .145, Revision 1, RegulatoryPosition C.1 .2. The EAB x/Q values for the radiological releases from each unit areconservatively based on the EAB distances from the outer edge of each containmentbuilding.An LPZ distance of 6 miles (9,654 meters) is used in the analysis. The use of one LPZdistance in all downwind directions from the center of the site for all release points isreasonable given the magnitude of this distance relative to the separation of the releasepoint locations from one another.The containment building cross-sectional area along with the containment buildingheight is used for the annual average x/Q calculations (used as input to develop theaccident x/Q values at the LPZ using Regulatory Guide 1.145 methodology). Theapplicable methodology for the annual average %/ calculations is identified in2.3-292.3-29Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATERegulatory Guide 1.111, Revision 1, Regulatory Position C.1.c (Reference 28). Theseannual average x/Q values are used to calculate the intermediate averaging time %/Qvalues for the periods of 2-8 hours, 8-24 hours, 1-4 days, and 4-30 days by logarithmicinterpolation.The following conservative assumptions are made for these calculations:* Releases are treated as point sources;* Releases are treated as ground-level as there are no release conditions thatare sufficiently high to escape the aerodynamic effects of the plant buildings;* The distances from the Unit 1 and Unit 2 releases are determined from theclosest edge of the containment buildings to the EAB;* The plume centerline from each release is transported directly over thereceptor; and.* A terrain recirculation factor of 4 is used in the calculation of the annualaverage x/Q valueso anid-ne-Rfadioactive decay or plume depletion due to deposition is notconsidered.The highest EAB and LP7 x/Q values from among all 22.50-downwind sectors for eachrelease/receptor combination and accident period are summarized in Table 2.3-1 45.EAB %/Q values are presented for releases from Unit 1 and Unit 2, while the LPZ ;(/Qvalues are applicable to both units. The 0.5% sector dependent z/Q values arepresented with the worst case downwind sector indicated in parentheses.2.3.5.2.2 On-Site Atmospheric Dispersion FactorsThe control room and technical support center %IQ values for radiological releases fromUnit 1 and Unit 2 are calculated using the NRC "Atmospheric Relative CONcentrationsin Building Wakes" (ARCON96) methodology as documented in NUREG/CR-6331,Revision 1 (Reference 29). Input data consist of: hourly on-site meteorological data;release characteristics (e.g., release height, building area affecting the release); andvarious receptor parameters (e.g., distance and direction from release to control roomair intake and intake height). Refer to Section 15.5.8.11 for a description of computerprogram ARCON96). -A continuous temporally representative 5-year period of hourly on-site meteorologicaldata from the DCPP onsite meteorological tower (i.e., January 1, 2007 throughDecember 31, 2011) is used for the ARCON96 analysis. Each hour of data, at aminimum, has a validated wind speed and direction at the 10-meter level and atemperature difference between the 76- and 10-meter levels. This period of data istemporally representative and meets the requirements of Safety Guide 23, February1972 (Reference 21).The ARCON96 modeling follows the ground level release requirements of RegulatoryPosition C.3 of Regulatory Guide 1 .194, June 2003 (Reference 30) relative to2.3-302.3-30Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATEdetermination of: (1) release height (i.e., ground-level vs. elevated); (2) release type(i.e., diffuse vs. point); and (3) configuration of release points and receptors (i.e.,building cross-sectional area, release heights, line-of-sight distance between releaseand receptor locations, initial diffusion coefficients etc.).Releases are assumed to be ground-level as none of the release points meet thedefinition of an elevated release as required by Regulatory Position C.3.2.2 ofRegulatory Guide 1.194, June 2003 (i.e., do not meet the requirement to be at aminimum 2.5 times the height of plant buildings).Only the containment building edge releases are treated as diffuse sources as thereleases occur from the entire surface of the building. In these cases, initial values ofthe diffusion coefficients (sigma y, sigma z) are determined in accordance with therequirements in Regulatory Guide 1.194, June 2003 Regulatory Position C.3.2.4.Release and receptor locations are applied in accordance with Regulatory Guide 1.194,June 2003 Regulatory Position C.3.4 requirements for building geometry and line-of-sitedistances.The following recommended default values from Regulatory Guide 1.194, June 2003,Table A-2, are judged to be applicable to DCPP:Wind direction range = 90 degrees azimuth;Wind speed assigned to calm = 0.5 m/sec;Surface roughness length = 0.20 m; andSector averaging constant = 4.3 (dimensionless)The following assumptions are made for %/Q calculations:o The plume centerline from each release is transported directly over the controlroom or technical support center air intake/receptor (conservative);oThe distances from the Unit 1 and Unit 2 containment building surfaces to thereceptors are determined from the closest edge of the containment buildings andthe release/receptor elevation differences are set to zero (conservative);* The applicable structure relative to quantifying building wake effects on thedispersion of the releases is based on release/receptor orientation relative to theplant structures;* The releases from the Unit 1 and Unit 2 containment building surfaces aretreated as diffuse sources;2.3-312.3-31Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE0All releases are treated as ground level as there are no release conditions thatmerit categorization as an elevated release (i.e., 2.5 times containment buildingheight) at this site (conservative); andThe x/Q value from the accident release point to the center of the control roomboundary at roof level is utilized for control room in-leakage since the above %/Qcan be considered an average value for in-leakage locations around the controlroom envelope. The y/Q from the accident release point to the center of thecontrol room boundary at roof level is also utilized for control room.ingress/egress. The outer doors to the control room are located at approximatelythe middle of a) the east side (i.e., auxiliary building side) wall of the control roomand b) the west side (i.e., turbine building side) wall of the control room. Similarly,the z/ from the accident release point to the center of the TSC at its roof level isutilized for TSC in-leakage since the above 7J can be considered an averagevalue for in-leakage locations around the TSC building envelope.Summarized below are some of the other salient aspects of the control room andtechnical support center %/ analyses, as applicable.Control Room Receptors within 10-meters of ReleaseRegulatory Guide 1.194, June 2003, Regulatory Position C.3.4 recommends thatARCON96 methodology not be used for analysis at distances less than about 10meters. However, as an exception to Regulatory Guide 1.194, June 2003,Regulatory Position C.3.4 the ARCON96 methodology has been applied for twocases when the distance from the release to the receptor is less than 10 meters.The distances in question (i.e., 9.4 meters for Unit 1 containment building to Unit1 control room normal intake and 7.8 meters for Unit 2 containment building toUnit 2 control room normal intake) is considered acceptable since the dominatingfactors in the calculation are building cross-sectional area and plume meander,not the normal atmospheric dispersion coefficients.Control Room Receptors at 1.5-meters from ReleaseSince the Unit 1 and Unit 2 MSSVs, 10% ADVs, and MSLB release points arelocated within 1.5 meters line-of-sight distance from the affected unit's controlroom normal intake, this near-field distance is considered outside of theARCON96 application domain. Although ARCO N96 is capable of estimatingnear-field dispersion, the 1 .5-meter line-of-sight distance from the releases to thereceptors is much less than the 10-meter distance recommended as theminimum applicable distance in Regulatory Position 0.3.4 of Regulatory Guide1.194, June 2003. Thus no z/Qs are developed for the above release point /receptor combinations.Enernqetic Releases2.3-322.3-32Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATEThe vertical velocity of the MSSV and 10% ADVs releases is at least 95 timeslarger than the 95th percentile wind speed of 1 rn/sec and approximately 5 timeslarger thanthe highest observed 10-meter wind speed (i.e., 18.9 m/sec) withinthe 5-year meteorological data base. The large vertical velocities of the MSSVand 10% ADVs releases, ranging from 94.9 to 98.9 m/sec, preclude any down-washing of the releases by the aerodynamic effects of the containment buildingssuch that the control room normal intake of the same unit as the release (e.g.,Unit 1 MSSV/IO% ADVs releases to Unit 1 CR normal intake) is notcontaminated given that the horizontal distance is only 1 .5 meters. Moreover,this short distance precludes the releases from reaching the control room normalintakes of the same unit given the height of the MSSV and 10% ADVs releases(i.e., 27.1 and 26.5 meters, respectively) relative to the height of the normalintakes (i.e., 22 meters). Plume rise calculations indicate that the MSSV andADV release heights will be enhanced by 11 meters at the 95th percentile windspeed of 1 rn/sec due to the large vertical velocities of the releases. Thus, forpurposes of estimating dose consequences, it is appropriate to use the x/Qassociated with the normal control room intake of the opposite unit for releasesfrom the MSSVs / 10% ADVs as the worst case control room intake location.Vertically-Oriented Enerqetic ReleasesRegulatory Position C.6 of Regulatory Guide 1 .194, June 2003 establishes theuse of a deterministic reduction factor of 5 applied to ARCON96 7./Q values forenergetic releases from steam relief valves or atmospheric dump valves. Thesevalves must be uncapped and vertically-oriented and the time-dependent verticalvelocity must exceed the 95th-percentile wind speed at the release point heightby at least a factor of 5. Since the DCPP MSSVs and 10% ADVs are vertically.oriented / uncapped and will have a vertical velocity of at least 94.9 rn/sec for thefirst 10.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months of the accident, the reduction factor of 5 is clearly applicable tothe DCPP MSSV and 10% ADVs releases. Note that since %!Q values areaveraged over the identified period (i.e., 0-2 hours, 2-8 hours, 8-24 hours, etc.),and the vertical velocity has been estimated to occur for 10.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months , applicationof the factor of 5 reduction is not appropriate for %/Q values applicable toaveraging periods .beyond the 2-8 hours averaging period. For assessment of anenvironmental release between 8 to 10.73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months , continued use of the 2-8 hour%IQ, with the factor of 5 reduction, is acceptable and conservative.Dual IntakesThe Unit 1 and Unit 2 control room pressurization air intakes which also serve thetechnical support center, may be considered dual intakes for the purpose ofproviding a low contamination intake regardless of wind direction for any of therelease points since the two control room pressurization air intakes are neverwithin the same wind direction window; defined as a wedge centered on the lineof sight between the release and the receptor with the vertex located at therelease point. The size of the wedge for each release-receptor combination is 902.3-332.3-33Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATEdegrees azimuth with the use of ARCON96, as described in Regulatory Position0.3.3.2 of Regulatory Guide 1.194, June 2003.Redundant Radiation MonitorsPer Regulatory Guide 1.194, June 2003, Regulatory Position 0.3.3.2.3, based onthe dual intake design of the control room pressurization intakes, and theavailability of redundant PG&E Design Class I radiation monitors at eachpressurization intake (which provide the capability of initial selection of thecleaner intake and support the expectation that the operator will manually makethe proper intake selection throughout the event), allows the x/Q valuesapplicable to the more favorable control room pressurization intake c-anto bereduced by a factor of 4 and utilized to estimate the dose consequences.PG&E Desiqjn Class II Lines Connecting to PG&E Design Class 1 Plant VentThe 16 inch PG&E Design Class 11 gland seal steam exhauster line connects tothe PG&E Design Class I plant vent. In addition, the plant vent expansion jointmay experience a tear during a seismic event, however the plant vent will remainintact and functional.a) The gland seal steam 16 inch exhauster line connects to the plant vent at El144'-6" (Centerline) on the North-East side / South-East side of the Unit 1and Unit 2 containments, respectively. It has been determined that should afailure occur due to a seismic event, it would occur at the interface of thisline and the plant vent.b) The plant vent expansion joint is located at El 155.83' North-East side ISouth-East side of the Unit 1 and Unit 2 containments, respectively. Asdiscussed earlier, the plant vent expansion joint may experience a tearduring a seismic event.An assessment of the potential release locations identified above indicates thatthe %/Q values developed for the plant vent are either conservative orrepresentative of these potential release points.Release points and receptor locations are provided in Figure 2.3-5, while Table 2.3-1 46provides the release point I receptor combinations that were evaluated. Tables 2.3-147and 2.3-148 provide the control room %/Q values for the individual release point-receptor combinations for Unit 1 and Unit 2, respectively.The XIQ values selected for use in the dose consequence analyses are intended tosupport bounding analyses for an accident that occurs at either unit. They take intoconsideration the various release points-receptors applicable to each accident in orderto identify the bounding x/Q values and reflect the allowable adjustments and reductionsin the values as discussed earlier and further summarized in the notes of Tables 2.3-1 47and 2.3-1 48.2.3-342.3-34Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATETable 2.3-1 49 presents the 7J values for the individual post-LOCA release point TSCreceptor combinations for Unit 1 and Unit 2 applicable to the TSC normal intake and thecenter of the TSC boundary at roof level (considered an average value for potential TSCunfiltered in-leakage locations around the envelope). The Unit 1 and Unit 2 controlroom pressurization air intakes also serve the TSC during the emergency mode. Thus,the 7J~ presented in Tables 2.3-1 47 and 2.3-148 for the control room pressurizationintakes inclusive of the credit for dual intake design and ability to select the morefavorable intake are also applicable to the TSC.2.3.6 LONG-TERM (ROUTINE) DIFFUSION ESTIMATESHISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.2.3.6.1 ObjectiveAnnual relative concentrations (z/Q) were estimated for distances out to 80 kilometersfrom on site meteorological data for the period May 1973 through April 1975. Theserelative concentrations are presented in Table 2.3-2; they were estimated using themodels described in Reference 18. The same program also produces cumulativefrequency distributions for selected averaging periods using overlapping means havinghourly updates. For critical offsite locations, measured lateral standard deviations ofwind direction, GrA, and bulk Richardson number, Ri, were used as the stabilityparameters in the computations. The meteorological input data were measured at the10 meter level of the meteorological tower at DCPP site. Annual averaged relativeconcentrations calculated by the above methods are presented in Table 2.3-4.2.3.6.2 CalculationsThe meteorological instrumentation that was used to obtain the input data for thepreviously discussed relative concentration calculations at DCPP site is described inSection 2.3.4. Procedures for obtaining annual averaged relative concentrations aredescribed in detail in Reference 15.2.3.6.3 Meteorological ParametersThe following assumptions were used in developing the meteorological inputparameters required in the dispersion model:(1) There is no wind direction change with height(2) Wind speed changes with height can be estimated by a power lawfunction where the exponent, F, varies with stability class and is assignedthe following values:Pas quill Stability Class Exponent (P)2.3-352.3-35Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE2.3.8.4 Safety Guide 23, February 1972 -Onsite Meteorological ProgramsAs discussed in Section 2.3.4, the pfeepefatipna-meteorological data collectionprogram was designed and has been updated continually to meet the requirements ofSafety Guide 23, February 1972.2.3.8.5 Regulatory Guide 1.97, Revision 3 -Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs ConditionsDuring and Following an AccidentWind speed, wind direction, and estimation of atmospheric stability indication in thecontrol room provide information for use in determining the magnitude of the release ofradioactive materials and in continuously assessing such releases during and followingan accident (refer to Table 7.5-6 for a summary of compliance to Regulatory Guide1 .97, Revision 3).2.3.8.6 Regulatory Guide 1.111, March 1976- Methods for EstimatingAtmospheric Transport and Dispersion of Gaseous Effluents in RoutineRe leases from Light-Water-Cooled ReactorsThe pre-operational values of dilution factor and deposition factor used in the calculationof annual average offsite radiation dose are discussed in Section 11 .3.7. The values ofdeposition rate were derived from Figure 7 of Regulatory Guide 1.111, March 1976, fora ground-level release.2.3.8.7 Regulatory Guide 1.111, Revision 1, July 1977 -Methods for EstimatingAtmospheric Transport and Dispersion of Gaseous Effluents in RoutineReleases from Light-Water-Cooled ReactorsThe annual average relative concentration values are developed for each sector, at theouter LPZ boundary distance for that sector, using the method described in RegulatoryPosition C.1 .c of Regulatory Guide 1.111, Revision 1. These values are used tocalculate the intermediate averaging time 7/0 values at the LPZ for the periods of 2-8hours, 8-24 hours, 1-4 days, and 4-30 days following the postulated accident. Thisinformation is used as input to develop the accident x/O values at the LPZ usingRegulatory Guide 1.145, Revision 1 methodology. Refer to Section 2.3.5.2.2.3.8.8 Regulatory Guide 1.145, Revision 1, February 1983 -AtmosphericDispersion Models for Potential Accident Consequence Assessments atNuclear Power PlantsThe short-term atmospheric dispersion factors applicable to the exclusion areaboundary and the low population zone for post-accident accident releases from Unit 1and Unit 2 are calculated using methodology applicable to "ground level" releasesprovided in Regulatory Guide 1.145, Revision 1. Refer to Section 2.3.5.2.2.3-382.3-38Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE2.3.8.9 Regulatory Guide 1.194, June 2003 -Atmospheric RelativeConcentrations for Control Room Radiological Habitability Assessmentsat Nuclear Power PlantsThe control room and technical support center atmospheric dispersion factors forradiological releases from Unit 1 and Unit 2 are calculated using methodology outlinedin Regulatory Positions 0.1 through C.3, and the adjustment factor for verticallyorientated energetic releases from steam relief valves and atmospheric dump valvesallowed by Regulatory Position C.6, and NRC ARCON96 methodology as documentedin NUREG/CR-6331, Revision 1. Refer to Section 2.3.5.2.2.3.8.710 NUREG-0737 (Item III.A.2), November 1980 -Clarification of TMIAction Plan RequirementsItem llII.A.2 -Improving Licensee Emergency Preparedness-Long-Term:As discussed in Section 2.3.4, the primary and backup meteorological data areavailable in the control room and emergency response facilities via the TRS servers andEARS, in accordance with NUREG-0654, Revision 1, November 1980.As discussed in Section 2.3.4, the measurement subsystems consist of a primarymeteorological tower and a backup meteorological tower. The primary meteorologicalcomputer and the backup meteorological computer communicate with each other, theEARS and also with the TRS server. Primary and backup meteorological data areavailable on the PPCs via the TRS servers and thus in the control room and emergencyresponse facilities.Item III.A.2.2 -Meteorological Data: NUREG-0737, Supplement 1, January 1983:Table 7.5-6 and Section 2.3.8.5 summarize DCPP conformance with Regulatory Guide1 .97, Revision 3. Wind direction, wind speed, and estimation of atmospheric stabilityare categorized as Type E variables, based on Regulatory Guide 1 .97, Revision 3. ThePPC is used as the indicating device to display meteorological instrument signals. Inaddition, Type E, Category 3, recorders are located in the meteorological towers.2.3.8.8---1 IE Information Notice 84-91, December 1984- Quality ControlProblems of Meteorological Measurements ProgramsIn addition to the primary meteorological towers, a supplemental meteorologicalmeasurement system is provided in the vicinity of the plant site in order to meetIE Information Notice 84-91. As discussed in Section 2.3.4.5, this supplementalmeasurement system consists of three Doppler SODAR and seven tower sites locatedas indicated in Figure 2.3-4. The primary and secondary meteorological towers inconjunction with the supplemental system adequately predict the meteorologicalconditions at the site boundary (800 meters) and beyond.2.3-392.3-39Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE24. ANSI/ANS 2.5, American National Standard for Determininq MeteoroloqicalInformation at Nuclear Power Sites, American Nuclear Society, 1984.25. National Oceanic and Atmospheric Administration, An Evaluation of WindMeasurements by Four Doppler SODARS, NOAA Wave Propagation Laboratory,1984.26. Deleted in Revision 20.27. PG&E reports previously submitted as Appendices 2.3A-K, 2.4A-C, and 2.5A-Fof the FSAR Update, Revision 0 through Revision 10 (Currently maintained atPG&E Nuclear Power Generation Licensing office files).28. Regulatory Guide 1.111, Revision 1, Methods for Estimating AtmosphericTransport and Dispersion of Gaseous Effluents in Routine Releases from LightWater Cooled Reactors, USNRC.29. Ramsdell, J. V. Jr. and C. A. Simonen, Atmospheric Relative Concentrations inBuilding Wakes. Prepared by Pacific Northwest Laboratory for the U.S. NuclearRegulatory Commission, PNL-10521, NUREG/CR-6331, Revision 1, May 1997.30. Regulatory Guide 1.194, June 2003, Atmospheric Relative Concentrations forControl Room Radiological Habitability Assessments at Nuclear Power Plants,USNRC.2.3-422.3-42Revision 21 September 2013 6tACT-6AY500-450- _400-- (300 @ _-+ -F ---:I , , IA100 A t -li- -I- -- f--1oo +-J ---L -~I Li ii iHiI ii ~iRTOL I , N- ----VTHIItUmD o. -- I-- -- ---.... I-l FUEL HAM)I.4GI BLDG. kI I 1 0 I-1 -I--. -- TC -i- ... ..! -- GW,I14II IK_ _15--CAuxX.,ARY BWI.OI4G---i-+-I---I-i----- -----l --[I-- --..... -_ l -I -....... BLDEQU.E.T .-... --/I ,',,,178

FFEi+.o l- --C- --M M -. ---I-h4, TC 2~113 27'0'416.33382,33337.833T2.83280.33II_ I i iZ -- £88.83+KD0R0©087233V~'7~.2I I I IIII II I-I--I--I-..--I---* I-I.-.-I--: I I I i I Ti I i i i i I i---I ---i -f i -I ,I t tI++,"- ',, I[ Ii -i-i-5-ii i-I-T-i--d-!-d-I- ---T -- T -F --- -i--I--H--- -T --i -- i T i -I -----1 ---I -? ---? ----------1 ', I SS_ T m _ _ _ _II,,.0. To 01, H, U i ;l , l + i : i t , !-,-,. : I 1 T200 250 300 350 400 450 I 500 550 600 650COOROW4ATE-COLUUN LtINE CROSS REFERENCE T ,700 "750 800 650#%FSAR UPDATEUNITS 1 AND 2DIABLO CANYON SITEFIGURE 2.3-5Post-Accident EnvironmentalRelease Point / Receptor Location DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.1-1 Set f1Sheet 8 of 12CRITERIA ITITLEAPPLICABILITYEnierdSafety Fetue Containment Containment Containment Containment Combustible Emergency Control Technical AuxiliaryIFunctional Heat Removal Air Purification Isolation Gas Control in Core Cooling Room Support FeedwaterDesign Systems and Cleanup System Containment System Habitability Center System_ _ _ _ _ _ _ _ _ _ _ _ I _ _ System _Section [6.2.1 6.2.2 6.2.3 6.2.4 6.2.5 6.3 6,4.1 6.4.2 6.55. Recqulatory Guides (contd.)Regulatory Guide Performance-Based1.163, Containment Leak- X XSeptember 1995 Test ProgramAlternativeRadiological SourceRegulatory Guide Terms for Evaluating1.163, July 2000 Design BasisAccidents at Nuclear______________Power ReactorsDemonstratingRegulatory Guide Control Room1.197, Revision 0, Envelope Integrity atXMay 2003 Nuclear PowerReactors6. NRC NUREGClarification of TMINUREG-0737, Action Plan X X X X X X XNovember 1980 Rqieet7. NRC Generic Letters -____Generic Letter Safety-Relatedt89-10, Motor-Operated I I xJune 1989 Valve Testing andISurveillanceRevision 22 DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.2-32LOSS-OF-COOLANT ACCIDENTTOTAL ELEMENTAL IODINE & PARTICULATE REMOVAL COEFFICIENTSElemental Iodine Removal Particulate RemovalCoefficient CoefficientFrom To Time (hr-1 -Note 1 (hr1)Time (sec) Sprayed Unsprayed Sprayed Unsprayed(sec) Region Region Region Region0 30 N/A N/A30 111 272.45.89 0.0062111 1,800 2.24 0.00711,800 3,798 20.57 (Note 2) 9.35 0.11443,798 4,518 0.00 (Note 3) __1.02_0.1224,518 5,030 7.50 0.12395,030 6,4806.0.136,480 7,200 19.91 (Note 2) 0.00 4.74____ 0.1236__7,200 8,004 3.39____ 0.1222__8,004 22,1521.300022,152 22,518 0022,518 720 hrs 0.00 _______0.00 (Note 4)______Notes:1. Per Regulatory Guide 1.183, July 2000 and SRP 6.5.2, Revision 4, removal credit forelemental iodine by sprays is eliminated after a DF=200 is reached in the containmentatmosphere.2. Wall deposition removal coefficient (0.57 hr"1) is included.3. Time period without spray.4. For purposes of conservatism, no credit is taken for particulate removal in the sprayedregion after termination of recirculation spray DCPP UNIT 1 & 2FSAR UPDATETABLE 6.2-33Sheet 1 of 3LOSS-OF-COOLANT ACCIDENTCONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATAPost-LOCA Time Containment Containment Containment -RHPressure TemperatureSeconds psia 0F%0.00 16.00 120 180.52 18.81 145.9 72.31.04 21.45 163.8 88.21.54 23.76 176.78 93.52.04 -25.88 186.8 962.54 27.80 194.83 97.43.04 29.44 201.03 98.13.54 30.86 205.97 98.54.04 32.10 210.01 98.94.54 33.21 213.43 99.15.04 34.24 216.47 99.27.04 38.04 226.75 99.57.54 38.96 229.03 99.58.54 40.68 233.06. 99.610.04 43.06 238.34 99.710.54 43.81 239.9 99.711.54 45.21 242.76 99.813.04 47.12 246.47 99.814.54 48.62 249.65 99.816.04 50.36 252.4 99.917.54 51.70 254.7 99.919.04 52.87 256.67 99.920.54 53.88 258.34 99.921.54 54.24 258.9 99.922.04 54.36 259.08 10023.54 54.48 259.25 10025.04 54.40 259.12 10029.54 53.87 258.24 10032.54 53.62 257.85 10048.54 53.14 257.02 10054.54 53.37 257.03 10068.04 53.70 256.79 10086.54 53.18 255.93 100144.18 50.88 251.95 100158.18 50.44 251.15 100 DCPP UNITi1 & 2FSAR UPDATETABLE 6.2-33Sheet 2 of 3LOSS-OF-COOLANT ACCIDENTCONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATAPost-LOCA Time Containment Containment Containment -RHPressure TemperatureSeconds psia 0F%188.18 49.70 249.82 100200.18 49.48 249.41 100212.18 49.33 249.15 100266.18 49.21 248.92 100333.18 49.37 249.2 100400.18 49.70 249.82 99.9534.18 50.56 251.76 99.1668.18 51.49 254.34 97.4803.18 52.60 256.32 97.3816.18 52.43 255.22- 98.7857.18 52.08 254.17 99.5912.18 51.73 253.6 99.41021.19 51.21 252.76 99.31131.19 50.83 252.11 99.21240.19 50.55 251.61 99.11458.19 50.16 .250.91 991677.19 .49.94 250.49 98.91730.19 50.43 251.61 98.51746.19 50.26 250.56 99.91859.19 49.41 248.91 1001988.19 48.57 247.32 1002247.19 47.07 244.4 1002505.19 45.75 241.75 1002764.19 44.56 239.29 1003022.19 43.47 236.96 99.93281.19 42.45 234.71 99.83604.24 41.26 231.94 99.93798.24 40.10 229.1 1003888.29 40.52 230.99 98.23978.29 40.92 233.84 9.4068.29 41.24 235.73 9.4158.29 41.48 237.00 9.4338.29 41.86 238.43 9.4518.29 42.13 239.07 9.4536.29 42.05 237.81 9.4555.29 41.94 235.91 9.

DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.2-33Sheet 3 of 3LOSS-OF-COOLANT ACCIDENTCONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATAPost-LOCA Time Containment Containment Containment -RHPressure TemperatureSeconds psia 0F%4573.29 41.87 234.69 97.24592.29 41.82 233.97 98.44666.29 41.74 233.23 99.55110.73 41.37 232.3 99.65700.73 40.79 230.94 99.66890.73 39.55 227.94 99.68080.73 38.36 224.96 99.510000.80 36.64 220.36 99.411001.50 35.82 218.08 99.412001.50 35.08 215.93 99.434.39 213.85 99.414001.50 33.74 211.85 99.415001.50 33.13 209.91 99.416001.50 32.57 208.07 99.318001.50 31.59 204.69 99.320001.50 30.71 201.53 99.321001.50 30.31 200.06 99.322518.00 29.78 198.01 99.4 DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.2-34Sheet 1 of 4LOSS-OF-COOLANT ACCIDENTCONTAI NMENT STEAM CONDENSTION DATAPost- Steam Condensation RateLOCA Thermal Containment Injection Recirculation Total SteamTime Conductor Fan Coolers Spray Spray Condensation Rat __Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/se0.00 0.00 0.00 0.00 0.00 0.00 0.0OC_0.52 8.37 0.00 0.00 0.00 8.37 3796. i71.04 49.71 0.00 0.00 0.00 49.71 22548 082.54 204.71 0.00 0.00 0.00 204.71 92854,903.04 250.47 0.00 0.00 0.00 250.47 .113611.293.54 290.16 0.00 0.00 0.00 290.16 131614.374.04 325.28 0.00 0.00 0.00 325.28 147544.545.04 384.15 0.00 0.00 0.00 384.15 174247.527.54 509.97 0.00 0.00 0.00 509.97 23131 .5210.04 611.01 0.00 0.00 0.00 611.01 27714£.4912.54 684.86 0.00 0.00 0.00 684.86 310647.2915.04 734.83 0.00 0.00 0.00 734.83 33331 .3017.54 766.16 0.00 0.00 0.00 766.16 347524.3520.54 783.18 0.00 0.00 0.00 783.18 355244.5024.54 742.58 0.00 0.00 0.00 742.58 33682 .6428.54 684.40 0.00 0.00 0.00 684.40 31043 .6432.54. 644.28 0.00 0.00 0.00 644.28 29224 .5137.04 623.09 0.00 0.00 0.00 623.09 28262 .8953.54 538.96 0.00 0.00 0.00 538.96 24446 .1670.04 469.30 0.00 0.00 0.00 469.30 21287 .9187.04 412.48 0.00 0.00 0.00 412.48 187097.7987.57 410.75 44.28 0.00 0.00 .455.03 20639 .1588.07 409.15 45.29 0.00 0.00 454.44 20613 .53107.14 354.70 44.87 13.61 0.00 413.18 18741 .31124.18 312.19 44.28 72.76 0.00 429.23 19469 .47146.18 267.97 43.51 70.37 0.00 381.85 17320 .26169.18 231.50 42.78 68.46 0.00 342.74 155464.26197.18 197.26 41.97 66.66 0.00 305.89 13874! .38234.18 166.95. 41.32 49.25 0.00 257.52 11680£.11262.18 149.95 41.00 48.48 0.00 239.43 .63327.18 121.98 40.50 47.28 0.00 209.76 95145 54403.18 100.43 40.23 46.58 0.00 187.24 849306B4449.18 91.02 40.17 46.66 0.00 177.85 80671 41502.18 82.32 40.11 47.29 0.00 169.72 76983 70Revision 19 May 2010 DCPP UNITI1 & 2FSAR UPDATETABLE 6.2-34LOSS-OF-COOLANT ACCIDENTCONTAINMENT STEAM CONDENSTION DATASheet 2 of 4Post- Steam Condensation Rate _________LOCA Thermal Containment Injection Recirculation Total SteamTime Conductor Fan Coolers Spray Spray Condensation Rat__Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/se _558.18 74.79 40.09 48.64 0.00 163.52 74171 43560.18 74.55 40.08 48.66 0.00 163.29 74067 10629.18 67.45 40.12 47.87 0.00 155.44 70506 40683.18 63.43 40.22 47.23 0.00 150.88 68438 02754.18 59.41 40.45 47.01 0.00 146.87 66619 12802.18 57.28 40.64 46.99 0.00 144.91 65730 07832.18 49.92 40.37 57.62 0.00 147.91 67090 85876.18 43.40 40.00 58.41 0.00 141.81 64323 94937.18 37.29 39.57 "57.66 0.00 134.52 61017 251013.19 32.26 39.16 56.94 0.00 128.36 58223 121094.19 28.58 38.82 56.39 0.00 123.79 56150 201148.19 26.76 .38.64 56.09 0.00 121.49 55106 941243.19 24.16 38.39 55.68 0.00 118.23 53628 231341.19 22.03 38.19 55.34 0.00 115.56 52417 141423.19 20.55 38.07 55.11 0.00 113.73 51587 061492.19 19.48 37.98 54.94 0.00 112.40 50983 791564.19 18.50 37.90 54.79 0.00 111.19 50434 941607.19 17.97 37.86 54.72 0.00 110.55 501446341644.19 17.54 37.83 54.66 0.00 110.03 49908 771672.19 17.23 37.81 54.61 0.00 109.65 49736 411678.19 17.19 37.71 54.73 0.00 109.63 49727 331730.19 19.93 36.37 52.72 0.00 109.02 494506541794.19 16.03 35.05 60.34 0.00 111.42 50539 271859.19 14.01 33.89 60.13 0.00 108.03 49001 591985.19 11.88 32.19 59.74 0.00 103.81 47087 432052.19 11.05 31.50 59.53 0.00 102.08 46302 712116.19 10.33 30.96 59.34 0.00 100.63 45645 002244.19 9.09 30.09 58.96 0.00 98.14 44515 562311.19 8.50 29.72 58.76 0.00 96.98 43989 392439.19 7.47 29.12 58.40 0.00 94.99 43086 742567.19 6.53 28.62 57.93 0.00 93.08 42220 382695.19 5.66 28.19 57.37 0.00 91.22 41376 702763.19 5.23 27.94 57.08 0.00 90.25 40936 712890.19 4.47 27.51 56.53 0.00 88.51 401471463018.19 3.76 27.11 56.00 0.00 86.87 39403 57Revision 19 May 2010 DCPP UNIT 1 & 2 ESAR UPDATETABLE 6.2-34Sheet 3 of 4LOSS-OF-COOLANT ACCIDENTCONTAINMENT STEAM CON DENSTION DATAPost- ______Steam Condensation Rate _________LOCA Thermal Containment Injection Recirculation Total SteamTime Conductor Fan Coolers Spray Spray Condensation Rat,__Seconds Ibmn/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/se __3082.19 3.43 26.92 55.74 0.00 86.09 39049 773210.19 2.80 26.56 55.23 0.00 84.59 38369 383338.19 2.21 26.22 54.74 0.00 83.17 37725 283466.19 1.64 25.89 54.25 0.00 81.78 370941793594.19 1.12 25.58 53.78 0.00 80.48 36505 123722.24 0.24 25.02 54.13 0.00 79.39 36010,7013796.24 0.02 24.72 53.61 0.00 78.35 35538 3843.29 2.24 24.91 0.00 0.00 27.15 12315 033901.29 4.39 25.13 0.00 0.00 29.52 13390 053995.29 6.72 25.42 0.00 0.00 32.14 14578 464105.29 8.33 25.70 0.00 0.00 34.03 154351754189.29 9.21 25.87 0.00 0.00 35.08 15912 0324291.29 10.03 26.06 0.00 0.00 36.09 16370 154383.29 10.63 26.21 0.00 0.00 36.84 16710 344463.29 11.06 26.33 0.00 0.00 37.39 169591824515.29 11.25 26.41 0.00 0.00 37.66 17082294518.29 11.26 26.41 0.00 0.92 38.59 17504 134584.29 10.17 26.48 0.00 8.26 44.91 20370 834592.29 10.15 26.49 0.00 9.02 45.66 207110334654.29 10.18 26.55 0.00 11.12 47.85 21704404698.29 10.19 26.59 0.00 11.33 48.11 21822 334734.29 10.21 26.60 0.00 11.39 48.20 21863 154785.29 10.19 26.62 0.00 11.43 48.24 21881 304807.29 10.17 26.63 0.00 11.44 48.24 21881 304843.29 10.14 26.63 0.00 11.46 48.23 21876764851.29 10.13 26.64 0.00 11.46 48.23 21876764895.29 10.09 26.64 0.00 11.48 48.21 21867694926.29 10.05 26.64 0.00 11.49 48.18 218541084932.29 10.06 26.63 0.00 11.49 48.18 21854i084988.29 9.99 26.63 0.00 11.51 48.13 21831 406120.73 8.70 25.98 0.00 11.43 46.11 20915 157400.73 7.52 25.02 0.00 11.13 43.67 19808388680.73 6.51 24.04 0.00 10.86 41.41 18783269510.73 6.06 23.47 0.00 10.70 40.23 18248 0214001.50 4.42 20.91 0.00 9.98 35.31 16016 35Revision 19 May 2010 DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.2-34Sheet 4 of 4LOSS-OF-COOLANT ACCIDENTCONTAINMENT STEAM CONDENSTION DATAPost- ____________Steam Condensation Rate _________LOCA Thermal Containment Injection Recirculation T Total SteamTime Conductor Fan Coolers S ray Spray Condensation RatSeconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec {Ibm/sec g/se__18001.50 3.69 18.95 0.00 9.59 32.23 14619 2822518.00 3.23 17.41 0.00 9.27 1 29.91 13566 95Revision 19 May 2010 DCPP UNIT 1 & 2 FSAR UPDATETABLE 6.2-35PARAMETERS FOR FISSION PRODUCT REMOVAL ANALYSISParameter ValueTotal Containment Volume, ft3 2.55 x 106 .Containment Spray Coverage 0.825FractionAverage Spray Fall Height, ft 116 (Note 1)Spray Flow Rate, gpm 2,456 for 111 sec-< t -3,798 sec0 for 3,798 sec <t-<4,518 sec______________________1,211 for 4,518sec<t-<22,518secSpray Droplet Radius, cm 500 x i 04Note 1: The average fall height is conservatively approximated as the distance from the lowestspray header to the operating deck as follows:Elevation of Header #1 =256'-0"Elevation of Deck = 140'-0'Fall Height to the Deck = 1 16'-0"Note that this fall height is more conservative than the area-weighted average drop fall height of128 ft shown in Table 6.2-37.

DCPP UNIT 1 & 2 FSAR UPDATE-ABLE462-36PARDAMEII:TERSI l:q RESU TS FORh SpRAY IOD'MINE: DREMOVt'\AL ANkALYISVlINJCION-fTf'k PHIASEI- Poara m tertContainment free ;'ol, fiUnaprayedve'lume, %Spray pump flo,-'iate, gppmC...nta...inet p~ressure, psigEstimate2. x 10'.-t----_-2(bMitnmtur--p4-4 !0---4 -7--4R-esultsremoval --9 --46 eonstxi~t for thespray system (lrl )ssntai. nt i~ntsri anaysi rs pr.... s.. in.. pp. ndi.:.. ... .... .. .potential. offsic, .ii- !nra dsc IA, thi s1; hangs is exreel s... .mall a,.d san' be ...ns.dercd, insignifisant(Refr~sneo~0O DCPP UNITS 1 & 2 FSAR UPDATETABLE 6.2-13CORE FIS .... PRODUCT ENERG ATEROPEDR-ATIONM WAITH- EXVTENDI'EDr FEl CYCLVPrESTrime After Reactor Trip,8-1020408O400Energy Re!ease Rate, Intcgrated Energy Release,watts!MWt 1 0~ watts days!MWt x 1 p404968A414344022_28642,404424(a) As..u.e. 50o/ coere halogens +99%oo/cther fission and4 eRevision 14 November 2001 DCPP UNITS 1 & 2 FSAR UPDATEF4SSION PRODUCT DECAY DEPOSTION IN SUMP SOLUTIONf-aAbump ,l-ISn :ss o autEnr '---1/2P¶eester-Tnp-1/240)-1-52040watts/M\At x 1 02-&65462g06w nt-tgas/Mt x 0&44$(a}-onsioers reiease 0ot u percent or core natogens,-noenoble-gasesan 1 ecnaroto ~inpodcst n sm ouinRevision 14 November 2001 DCPP UNITS 1 & 2 FSAR UPDATE15.5 RADIOLOGICAL CONSEQUENCES OF PLANT ACCIDENTSThe purposes of this section are: (a) to identify accidental events that could causeradiological consequences, (b) to provide an assessment of the consequences of theseaccidents, and (c) to demonstrate that the potential consequences of these occurrencesare within the limits, guidelines, and regulations established by the NRC.An accident is an unexpected chain of events; that is, a process, rather than a singleevent. In the analyses reported in this section, the basic events involved in variouspossible plant accidents are identified and studied with regard to the performance of theengineered safety features (ESF). The full spectrum of plant conditions has beendivided into four categories in accordance with their anticipated frequency of occurrenceand risk to the public. The four categories as defined above are as follows:Condition I: Normal Operation and Operational TransientsCondition 11: Faults of Moderate FrequencyCondition Ill: Infrequent FaultsCondition IV: Limiting FaultsThe basic principle applied in relating design requirements to each of these conditions isthat the most frequent occurrences must yield little or no radiological risk to the public;and those extreme situations having the potential for the greatest risk to the public shallbe those least likely to occur.These categories and principles were developed by the American Nuclear Society(Reference 1). Similar, though not identical, categories have been defined in the guideto the Preparation of Environmental Reports (Reference 3). While some differencesexist in the manner of sorting the different accidents into categories in these documents,the basic principles are the same.It should also be noted that the range of plant operating parameters included in theCondition I category, and some of those in the Condition 11 category, fall in the range ofnormal operation. For this reason, the radioactive releases and radiological exposuresassociated with these conditions are analyzed in Chapter 11 and are not discussedseparately in this chapter. The analyses of the variations in system parametersassociated with Condition I occurrences or operating modes are discussed in Chapter 7since these states are not accident conditions. In addition, some of the events identifiedas potential accidents in Regulatory Guide 1.70, Revision 1 (Reference 2), have nosignificant radiological consequences, or result in minor releases within the range ofnormal releases, and are thus not analyzed separately in this chapter.15.5-115.5-IRevision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.1 DESIGN BASESThe following regulatory requirements, including Code of," Federal Regulati.D .. FR)10 CFR Par.t 100, Genr- De...igr,,n Critria; (DC), Safety and. Regul...ator,,Gudsare applicable to the DCPP radiological consequence analyses presented in thisThey form the bases of the acceptance criteria and methodologies asdescribed in the following Sections:(1) 10 CFR Part 100, "Reactor Site Criteria"(2) 10 CFR 50.67, "Accident Source Term"(3) General Design Criterion 19,-1-97-1-1999 "Control Room"(4) Regulatory Guide 1 .4, Revision 1, "Assumptions Used for Evaluating thePotential Radiological Consequences of a Loss of Coolant Accident forPressurized Water Reactors"(5) Safet Gu.dc 7, March 1971, ,Conro.l of C~,ombusibl Gas Co..... rtios in(6~)(5) Safety Guide 24, March 1972, "Assumptions Used for Evaluating thePotential Radiological Consequences of a Pressurized Water ReactorRadioactive Gas Storage Tank Failure"Consequences a Fuel, Handl,,,ing, A^ccident in,, t,;he,, Handlingand Storg for Boilin and. Pressuri..ed Water Rea,-tors"{8)(6) Regulatory Guide 1.183, July 2000, "Alternative Radiological SourceTerms for Evaluating Design Basis Accidents at Nuclear Power Reactors"(9) Reguato,,,,-,,'_,Guide !.!9,,5,Ma 20,--,"'°n "Methods,4, Assumption for,., E,.a-luating,-15.5.1.1 List of Analyzed AccidentsThe following table summarizes the accident events that have been evaluated forradiological consequences. The table identifies the applicable UFSAR Sectiondescribing the analysis and results for each event, the offsite/onsite locations andapplicable dose limits, and the radiological analysis and isotopic core inventory codesused.15.5-215.5-2Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATEFSRRadiological Isotopic CoreAcietEet Section Boundary Dose Limit Analysis InventoryCode(s) Code(s)Loss of Electrical 15.5.10 EAB and LPZ RADTRAD SAS2 /Load (LOL) Control ORIGEN-RoomE-A8- rem SEMEP.ALD~TEDE5 rem TEDECONDITION IllSmall Break 15.5.11 EAB and LPZ 2.5 remn TEDE N/A N/ALOCA (SBLOCA) Control ;=04mRefer to Refer toRoomE-AB- 25-84m Section Sectiona~4~5 rem TEDE 15.5.23E-ME-RA 15.523E-ME-RATrhyre4 L-bL15.5-315.5-3Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFSRRadiological Isotopic CoreAcietEet Section Boundary Dose Limit Analysis InventoryCode(s) Code(s)Minor Secondary 15.5.12 EAB and LPZ 2.5 rem TEDE N/A N/ASystem Pipe Refer to Refer toBreaks :T-Ayred OQ-re* Section SectionWhGle-Beety 25re 15.5.18N/A 15.5.18NIARefe-re R4ef-e~4Scin55.2 Section 15.5.12Inadvertent 15.5.13 EAB and LPZ 2.5 rem TEDE N/A N/ALoading of a Fuel E-AS-ani-L-P- Refer to Refer toAssembly Section 15.5.13 Section 15.5.13Complete Loss of 15.5.14 LAB and LPZ 2.5 rem TEDE N/A N/AForced Reactor Refer to Refer toCoolant Flow Thri Section SectionW^h,,e,,,,.-p,,ody ,?,0e-re 15.5.4410 15.5.141025 remUnder-Frequency 15.5.15 LAB and LPZ 2.5 rem TEDE N/A N/AEA-n-PZRefer to Refer toSection SectionWholeBedy 3004en4 15.5.10E-ME-RA 15.5.10E-ME-RA2-84emL-L-Single Rod 15.5.16 LAB and LPZ 2.5 rem TEDE N/A N/ACluster Control Refer to Refer toAssembly Thri 3G4ei Section SectionWithdrawal Whl-Bd 2,5--rein 15.5.23E-MER#A 15.5.23E-ME-RACONDITION IVLarge Break 15.5.17 LAB and LPZ 25 rem TEDE RADTRAD 3 03 SAS2 /LOCA (LOCA) Control Room 5 rem TEDE PERC2EME-RA ORIGEN-TSCEAB-.ard 5 rem TEDE L-& SE-MF-RAI4LZ300 rem L-QGADOSE OP4GEN-2Whole-Bod'Control Room 30 rem15.5-415.5-4Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFSRRadiological Isotopic CoreAcietEet Section Boundary Dose Limit Analysis InventoryCode(s) Code(s)Main Steam Line 15.5.18 EAB and LPZ RADTRAD SAS2 /Break (MSLB) ORIGEN-Pre-Accident E-SORlGEN-2Iodine SpikeThyid0O-25 remWhele-Bedy TEDE25 ,remnAccident-initiatedIodine SpikeWoe-ey 2.5 rem TEDEControl Room~ em5 rem TEDEMain Feedwater 15.5.19 EAB and LPZ N/A N/ALine Break Refer to Refer to(FWVLB) Pre-Accident 25 rem TEDE Section SectionIodine Spike 15.5.189 15.5.198Accident- 2,5 rem TEDEinitiatedIodine SpikeE-AB nd LPZ 30rm15.5-515.5-5Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEFSRRadiological Isotopic CoreAcietEet Section Boundary Dose Limit Analysis InventoryCode(s) Code(s)Steam Generator 15.5.20 EAB and LPZ RADTRAD SAS2 /Tube Rupture 3.03RADTR-A, ORIGEN-(SGTR) Pre-Accident 25 rem TEDE SEME-RAbQ--Iodine Spike NO'RMA4=Accident- 2.5 rem TEDEinitiatedIodine SpikeControl Room 5 remn TEDEz-300remnTh~4 25-remAccident (LIA) ACcidntrl 5rmT E 3.E-RAD OIGNinitiAted ;30 -remSEM-L-_ _-LP- 2 5 em30-remnControl Room eLoced Rotorng 15.5.21. EAB and LPZ 2.5 rem TEDE RADTRAD SAS2 /Accident (LRA) Control 5 rem TEDE 3.03L-ORADLB ORIGEN-.03O S'- EDEE- SOMRAGENAraad-L-P-Z ~Gontb!Room~~e155-hRvsin 9Mai21 DCPP UNITS 1 & 2 FSAR UPDATEFSRRadiological Isotopic CoreAcietEet Section Boundary Dose Limit Analysis Inventory__________Code(s) Code(s)Hand!.n ,AB-aJd-bJP-_ L nGAnnS OR!GEN-2.Inside- Trhyr4 75 .remControl RoomControl Rod 15.5.23 EAB and LPZ 6.3 rem TEDE RADTRAD SAS2 IEjection Accident Control 5 rem TEDE 3.03EMERALD ORIGEN-(CREA) RoomF=AB- ;=380em S=ME-RAL-DWhoe~eodyControl Room S-romWaste Gas Decay 15.5.24 EAB and LPZ EMERALD EMERALDTank Rupture Thyroid 300 remWhole Body 25 remLiquid Holdup 15.5.25 EAB and LPZ LOCADOSE EMERALDTank Rupture Thyroid 300 remWhole Body 25 remVolume Control 15.5.26 EAB and LPZ EMERALD EMERALDTank Rupture Thyroid 300 remWhole Body 25 rem15.5.1.2 Assumptions associated with Loss of Offsite PowerThe assumptions regarding the occurrence and timing of a Loss of Offsite Power(LOOP) during an accident are selected with the intent of maximizing the doseconsequences. A LOOP is assumed for events that have the potential to cause gridperturbation.i. The dose consequences of the LOCA, MSLB, SGTR, LRA, CREA and LOL eventare evaluated with the assumption of a LOOP concurrent with reactor trip.ii. The assumption of a LOOP related to a postulated design basis accident whichleads to a reactor trip does not directly correlate to an FHA. Specifically, a FHAdoes not directly cause a reactor trip and a subsequent LOOP due to gridinstability; nor can a LOOP be the initiator of a FHA. Thus the FHA doseconsequence analyses are evaluated without the assumption of a LOOP.In addition, in accordance with current DCPP licensing basis, the non-accident unit isassumed unaffected by the LOOP.15.5-715.5-7Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE15.5.2 APPROACH TO ANALYSES OF RADIOLOGICAL EFFECTS OFACCIDENTS15.5.2.1 IntroductionThe potential radiological effects of plant accidents are analyzed by the evaluation of allphysical factors involved in each chain of events which might result in radiationexposures to humans. These factors include the meteorological conditions existing atthe time of the accident, the radionuclide uptake rates, exposure times and distances,as well as the many factors which depend on the plant design and mode of operation.In these analyses, the factors affecting the consequences of each accident areidentified and evaluated, and uncertainties in their values are discussed. Becausesome degree of uncertainty always exists in the prediction of these factors, it hasbecome general practice to assume conservative values in making calculated estimatesof radiation doses. For example, it is customarily assumed that the accident occurs at atime when very unfavorable weather conditions exist, and that the performance of theplant engineered safety systems is degraded by unexpected failures. The use of theseunfavorable values for the various factors involved in the analysis provides assurancethat each safety system has been designed adequately; that is, with sufficient capacityto cover the full range of effects to which each system could be subjected. For thisreason, these conservative values for each factor have been called design basis values.In a similar way, the specific chain of events in which all unfavorable factors arecoincidentally assumed to occur has been called a design basis accident (DBA). Thecalculated doses for the DBA, provide a basis for determination of the design adequacyof the plant safety systems. In the process of safety review and licensing, the radiationexposure levels calculated for the DBA are compared to the regulatory limits gu-ideln-h-values-established in 10 CFR 100.11 and 10 CFR 50.67including acceptance criteriaproposed in regulatory guidance, and if these calculated exposures fall below theregulatory guidelines-L-evels, the plant safety systems are judged to be adequate.-T-hc calculated4 e..po..ur ... reulin from a BA are...n... "" far in exce.. of what+.ouldbe..p.ctd... do....not... prvde a... relsi enosesngteepcein tho ,ernidPnft which osfimafos of tho 2ntuaI ,J~hme efnnd to oc'ir if thoaccidcnt took place. The resulting doses were close to the doses expected to resultfrom an .accident of tpe. The, second',r,,, -,-case, DBlA,,, use customary',..,,e-stimate÷ of doses, a1bais o r n,+,, of the desig~rn15.5-815.5-8Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEAs noted in Section 111.2.a of Standard Review Plan Section 15.0.1, Revision 0,(Reference 59), a full implementation of AST addresses a) all the characteristics of AST(i.e., the radionuclide composition and magnitude, chemical and physical form of theradionuclides, and the timing of the release of these nuclides), b) replaces the previousaccident source term used in all design basis radiological analyses, and c) incorporatesthe Total Effective Dose Equivalent (TEDE) criteria of 10 CFR 50.67, and Section II ofStandard Review Plan 15.0.1, Revision 0.The dose consequences of the following accidents have been re-evaluated using ASTin accordance with Regulatory Guide 1.183, July 2000.1. Loss of Coolant Accident (LOCA) -Section 15.5.172. Fuel Handling Accident (FHA) -Section 15.5.223. Locked Rotor Accident (LRA) -Section 15.5.214. Control Rod Ejection Accident (CREA) -Section 15.5.235. Main Steam Line Break (MSLB) -Section 15.5.186. Steam Generator Tube Rupture (SGTR) -Section 1.5.5.207. Loss-of Load (LOL) Event -Section 15.5.10The tank rupture events (i.e., Rupture of a Waste Gas Decay Tank, Section 15.5.24;Rupture of a Liquid Holdup Tank, Section 15.5.25; Rupture of a Volume Control Tank,Section 15.5.26) represent accidental release of radioactivity accumulated in tanksresulting from normal plant operations, thus the source term characteristics of AST arenot applicable to these events.The dose consequences for the remaining accidents are addressed by qualitativecomparison to the seven accidents listed above (with the exception of the tank ruptureevents).Note reference to Regulatory Guide 1.183, July 2000 is used extensively within thissection, as a result any reference to "Regulatory Guide 1.183" within Section 15.5 refersto Regulatory Guide 1.183, July 2000.The methodology used to assess the dose consequences of the DBAs, including thespecific values of all important parameters, data, and assumptions used in theradiological exposure calculations are listed in the following sections. The computerprograms used to assess the dose consequences of the DBAs are described briefly inSection 15.5.8.As discussed previously, certain- radiological source terms for accidents and some ofthe releases resulting from Condition I and Condition II events have been included inChapter 11.15.5.2.3 Dose Acceptance CriteriaEAB and LPZ Dose15.5-915.5-9Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe dose acceptance criteria presented below for the EAB and LPZ reflect use of ASTand are applicable to all accidents with the exception of the tank rupture events. Thetank rupture events are evaluated against 100 CFR100.11 (refer to Sections 15.5.1.1and 15.5.24 through 15.5.26 for detail)The acceptance criteria for the Exclusion Area Boundary (EAB) and the Low PopulationZone (LPZ) Dose are based on 10 CFR 50.67, and Section 4.4, Table 6 of RegulatoryGuide 1.183:(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 the accident-specific TEDE valuenoted in Reference 55, Table 6.(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 radiation dose in excess of the accident-specific TEDE value noted inReference 55, Table 6.EAB and LPZ Dose Acceptance Criteria -Condition II and Condition III events:Regulatory Guide 1 .183, does not specifically address Condition II and Condition Illscenarios. However, per Regulatory Guide 1.183, Section 1.2.1, a full implementationof AST allows a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in alldose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183indicates that for events with a higher probability of occurrence than those listed inTable 6 of Regulatory Guide 1.183, the postulated EAB and LPZ doses should notexceed the criteria tabulated in Table 6. Thus, the dose consequences at the EABand LPZ will be limited to the lowest value reported in Table 6, i.e., a small fraction(10%) of the limit imposed by 10 CFR 50.67.Control Room DoseThe acceptance criterion for the control room dose is based on 10 CFR 50.67.Adequate radiation protection is provided to permit access to 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.This criteria ensures that the dose criteria of GDC 19, 1999 and NUREG-0737,November 1980, Item lll.D.3.4 (refer to Section 6.4.1) is met.Technical Support Center Dose15.5-1015.5-10Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe 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 CER50.67. The dose to an operator in the TSC should not exceed 5 remn TEDE for theduration of the accident.15.5.2.4 Dose Calculation MethodologyThe dose calculation methodology presented below reflects use of AST and isapplicable to all accidents with the exception of the tank rupture events. Themethodology used for the tank rupture events are discussed in the accident specificsections, i.e., Sections 15.5.24 through 15.5.26.15.5.2.4.1 Inhalation and Submersion Doses from Airborne RadioactivityComputer Code RADTRAD 3.03 is used to calculate the committed effective doseequivalent (CEDE) from inhalation and the effective dose equivalent (EDE) fromsubmersion due to airborne radioactivity at offsite locations and in the control room.The summation of CEDE and EDE is reported as TEDE, in accordance with Section4.1 .4 of Regulatory Guide 1.183.The CEDE is calculated using the inhalation dose conversion factors provided in Table2.1 of Federal Guidance Report 11 (Reference 41).The submersion EDE is calculated using the air submersion dose coefficients providedin Table 111.1 of Federal Guidance Report 12 (Reference 42). The dose coefficients arederived based on a semi-infinite cloud model. The submersion EDE is reported as thewhole body dose in the RADTRAD 3.03 output.RADTRAD 3.03 includes models for a variety of processes that can attenuate and/ortransport radionuclides. It can model the effect of sprays and natural deposition thatreduce the quantity of radionuclides suspended in the containment or othercompartments. In addition, it can model the flow of radionuclides betweencompartments within a building, from buildings into the environment, and from theenvironment into a control room. These flows can be through filters, piping, or simplydue to air leakage. RADTRAD 3.03 can also model radioactive decay and in-growth ofdaughters. Ultimately the program calculates the whole body dose, the thyroid dose,and the TEDE dose (rem) to the public located offsite, and to onsite personnel locatedin the control room due to inhalation and submersion in airborne radioactivity based onuser specified, fuel inventory, nuclear data, dispersion coefficients, and dose conversionfactors. Note that the code uses a numerical solution approach to solve coupledordinary differential equations. The basic equation for radionuclide transport andremoval is the same for all compartments. The program breaks its processing into 2parts a) radioactive transport and b) radioactive decay and daughter in-growth.Computer Code PERC2 is used to calculate the CEDE from inhalation and the EDEfrom submersion due to airborne radioactivity in the TSC. PERC2 is a multiple15.5-1115.5-11Revision 19 May2010 DCPP UNITS 1 & 2 FSAR UPDATEcompartment activity transport code with the dose model consistent with RegulatoryGuide 1.183. The decay and daughter build-up during the activity transport amongcompartments and the various cleanup mechanisms are included. The CEDE iscalculated using the Federal Guidance Report No.11 (Reference 41) dose conversionfactors. The EDE in the TSC is based on a finite cloud model that addresses buildupand attenuation in air. The dose equation is based on the assumption that the dosepoint is at the center of a hemisphere of the same volume as the TSC. The dose rate atthat point is calculated as the sum of typical differential shell elements at a radius R.The equation utilizes the integrated activity in the TSC air space, the photon energyrelease rates per energy group from activity airborne in the TSC, and theANSI/ANS 6.1.1-1991 neutron and gamma-ray fluence-to-dose factors. (Reference 84)Offsite DoseIn accordance with Regulatory Guide 1.183, for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate ofthe public located offsite is assumed to be 3.5xl0A m3/sec. From 8 to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sfollowing the accident, the breathing rate is assumed to be 1 .8x1 0- m3/sec. After thatand until the end of the accident, the rate is assumed to be 2.3x10-4 m3/sec. Themaximum EAB TEDE for any two-hour period following the start of the radioactivityrelease is calculated and used in determining compliance with the dose criteria in 10CFR 50.67. The LPZ TEDE is determined for the most limiting receptor at-the outerboundary of the low population zone and is calculated for the entire accident duration.Control Room DoseThe control room inhalation CEDE is calculated assuming a breathing rate of 3.5x1 04m3/sec for the duration of the event. The following occupancy factors are credited indetermining the control room TEDE: 1 .0 during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 0.6between 1 and 4 days, and 0.4 from 4 days to 30 days. The submersion EDE iscorrected for the difference in the finite cloud geometry in the control room and thesemi-infinite cloud model used in calculating the dose coefficients. The followingexpression obtained from Regulatory Guide 1.183 is used in RADTRAD 3.03 tocorrect the semi-infinite cloud dose, EDEnO, to a finite cloud dose, EDEi'inite, where thecontrol room is modeled as a hemisphere that has a volume, V, in cubic feet,equivalent to that of the control room.EDEJ~~ =EDEh= V0.33s-1173Technical Support Center DoseThe TSC inhalation CEDE is calculated by computer code PERC2 assuming the samebreathing rate and occupancy factors as those used in determining the control roomdose. The submersion EDE developed by PERC2 (which computes the photonfluence at the center of TSC and utilizes the ANSI/ANS 6.1 .1-1991 fluence to effectivedose conversion factors), is a close approximation of the dose determined using Table111.1 of Federal Guidance Report No. 12 (Reference 42) (refer to Section 4.1.4,15.5-1215.5-12Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATERegulatory Guide 1 .183) and adjusted by the finite volume correction factor given inRegulatory Guide 1.183, Section 4.2.7.15.5.2.4.2 Direct Shine Dose from External and Contained SourcesComputer program SW-QADCGGP is used to calculate the deep dose equivalent(DDE) in the control room, TSC and at the EAB due to external and contained sourcesfollowing a LOCA. The calculated DDE is added to the inhalation (CEDE) and thesubmersion (EDE) dose due to airborne radioactivity to develop the final TEDE.Conservative build-up factors are used and the geometry models are prepared toensure that un-accounted streaming/scattering paths were eliminated. The dosealbedo method with conservative albedo values is used to estimate the scatter dose insituations where the scattering contributions are potentially significant.ANSI/ANS 6.1.1-1977 (Reference 83) is used to convert the gamma flux to the doseequivalent rate.The .pecifi value.. o...f all impo,, ant parameters .... data, and,, a...umptions used, in thev,'M ,I ra ,I', r erc I~cfA ;,4 +kn er{rrc Tkhe ,~4.-',i;l +kR.............~........implementation of the equations, models,the original licensing basis computer codethe EMERALD computer program (Referecomputer program (Reference 5), which aSnec 'I) and4 the EMERALD Nr M/RAL:re described briefl,. in Section -1558+:L15.5.3 ACTIVITY INVENTORIES IN THE PLANT PRIOR TO ACCIDENTS15.5.3.1 Design Basis Accidents Excluding Tank RupturesThe fission product inventories in the reactor core, the fuel rod gaps, and the primarycoolant prior to an accident have been conservatively calculated based on plantoperation at 105% of the current licensed rated thermal power of 3411 MWth, withcurrent licensed values of fuel enrichment and fuel burnup. 1 1.1 12 b the= EMRALDA[ code, except for slight` d4iffrences- in somenucidesdue lto di,4frferent initial co,-re and irradPia{tion accdet~pPsecondar,' syste inventories are, lis.ed in{a , Tablea 11.23 I_' t shou"d be noted that+ these-15.5-1315. 5-13Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEstem,,,e f...........e.. andmasses ..... ,pproximat lum1,pcd u.. ed for .., acti+it,ma.se.. WAh;ie thes .alues re.. am dequate for activty balance.., the s..hould!, not+ be15.5.3.1.1 Core Activity InventoryIn accordance with Section 3.1 of Regulatory Guide 1.183, the inventory of fissionproducts in the reactor core available for release to the containment following anaccident should reflect maximum full power operation of the core with the currentlicensed values for fuel enrichment, fuel burnup, and an assumed core power equal tothe current licensed rated thermal power times the ECCS evaluation uncertainty in the10OCFR50 Appendix K analysis (typically 1 .02).The equilibrium core inventory is calculated using computer code ORIGEN-S. Thecalculation is performed using the Control Module SAS2 of the SCALE 4.3 computercode package. The SAS2 control module provides a sequence to calculate the nuclideinventory in a fuel assembly by calling various neutron cross section treatment modulesand the exponential matrix point-depletion module ORIGEN-S. It calculates the time-dependent neutron flux and the buildup of fissile trans-uranium nuclides. It accounts forall major nuclear interactions including fission, activation, and various neutronabsorption reactions with materials in the core. It calculates the neutron-activatedproducts, the actinides and the fission products in a reactor core.The reactor core consists of 193 fuel assemblies with various Uranium-235enrichments. Per control imposed by DCPP core-reload design documentation, thepeak rod burnup limit at the end of cycle is not allowed to exceed 62,000 MWD/MTU.The current licensed maximum value for fuel enrichment is 5.0%. To account forvariation of U-235 enrichment in fresh fuel, the radionuclide inventories were calculatedfor a 4.2% average enriched core (representing minimum enrichment at DCPP), and 5%average enriched core (representing maximum enrichment). The higher activity foreach isotope from the above two enrichment cases is chosen to represent the inventoryof that isotope in the equilibrium core.The equilibrium core at the end of a fuel cycle is assumed to consist of fuel assemblies.with three different burnups, i.e., approximately 1/3 of the core is subjected to one fuelcycle, 1/3 of the core to two fuel cycles and 1/3 of the core to three fuel cycles. Thisapproach has been demonstrated to develop an isotopic core inventory that is areasonable and conservative approximation of a core inventory developed using DCPPspecific fuel management history data. Minor variations in fuel irradiation time andduration of refueling outages will have a slight impact on the estimated inventory oflong-lived isotopes in the core. However, these inventory changes will have aninsignificant impact on the radiological consequences of postulated accidents. A 4%margin has been included in the final isotopic radioactive inventories in support ofbounding analyses and to address minor changes in future fuel management schemes.15.5-1415.5-14Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEA 19 month fuel cycle length was utilized in the analysis. The 19-month average fuel cycle is anartifact of the current DCPP fuel management scheme which specifies 3 fuel cycles every 5years and refueling outages in Spring or Fall.In summary, the equilibrium isotopic core average inventory is based on:i. A power level of 3580 MWth inclusive of power uncertainty.ii. A range of enrichment of 4.2 to 5.0 w % U-235. Use of a few assemblies withlower enrichment is a common industry practice when replacing assembliespreviously irradiated but proven unsuitable for continued irradiation. As theseassemblies are designed to replace higher enrichment assemblies with ones ofsimilar reactivity for the remainder of the fuel cycle, their inventory is envelopedby the isotopic core average inventory developed to support the doseconsequence analyses.iii. A maximum core average burnup of 50 GWD/MTU.The core inventory developed by ORIGEN-S using the above methodology includesover 800 isotopes. The DCPP equilibrium core fission product inventory of dosesignificant isotopes relative to LWR accidents is presented in Table 15.5-77.15.5.3.1.2 Coolant Activity Inventory1. Desig~n Basis Primary and Secondary Coolant Activity ConcentrationsComputer code, ACTIVITY.2, is used to calculate the design basis primary coolantactivity concentrations for both DCPP Unit 1 and Unit 2 based on the core inventorydeveloped using ORIGEN-S and discussed in Section 15.5.3.1. The source terms forthe primary coolant fission product activity include leakage from 1% fuel defects and thedecay of parent and second parent isotopes. The depletion terms of the primarycoolant fission product activity include radioactive decay, purification of the letdown flowand neutron absorption when the coolant passes the reactor core. The nuclear libraryincludes 3rd order decay chains of approximately 200 isotopes.Computer code, IONEXCHANGER, is used to calculate the design basis halogen andremainder activity concentrations in the secondary side liquid. The source terms for thesecondary side activity include the primary-to-secondary leakage in steam generatorsand the decay products of parent and second parent isotopes. The depletion terms ofthe secondary side liquid activity include radioactive decay, and purification due to thesteam generator blowdown flow, and continuous condensate polishing.The design basis noble gas concentrations in the secondary steam are calculated bydividing the appearance rate (pJCi/sec) by the steam flow rate (gm/sec). The noble gasappearance rate in the steam generator steam space includes the primary-to-secondaryleak contribution and the noble gas generation due to decay of halogens in the SGliquid. The activity concentrations of the other isotopes in the steam are determined by15.5-1515.5-15Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEthe SG liquid concentrations and the partition coefficients recommended in NUREG0017, Revision 1 (Reference 56).2. Technical .Specification Primary and Secondary Coolant Activity ConcentrationsIn accordance with Technical Specifications the primary coolant Technical Specificationactivities for iodines and noble gases are based on 1.0 PJCi/gm Dose Equivalent (DE) I-131 and 270 pCi/gm DE Xe-133, respectively.The Technical Specification based primary coolant isotopic activity reflect the following:a. Isotopic compositions based on the design basis primary coolant equilibriumconcentrations at 1% fuel defects.b. Iodine concentrations based on the thyroid inhalation weighting factors for1-131, 1-132, 1-133, 1-134, and 1-135 obtained from Federal Guidance Report11 (Reference 41).c. Noble gas concentrations based on the submersion weighting factors for Xe-133, Xe-133m, Xe-135m, Xe-135, Xe-138, Kr-85m, Kr-87 and Kr-88 obtainedfrom Federal Guidance Report 12 (Reference 42)The Technical Specification 1 pCi/gm DE 1-131 concentrations per nuclide in theprimary coolant are calculated with the following equation:D~11()(u~ i)=C(i) x C1,o, (15.5-1)= ,T__Z{FQ0 xC(i)}Where:F(i) = DCF(i) / DCF E-131DCF(i)= Federal Guidance Report-il, Table 2-1 (Reference 41) Thyroid Dose ConversionFactor per Nuclide (Rem/Cl)C(i) = design basis primary coolant equilibrium iodine concentration per nuclide (IpCi/gm)CTtot= primary coolant total (DE 1-131) Technical Specification iodine concentration(pCi/gm).The CTtot for the pre-accident iodine spike is 60 pJCi/gm (transient TechnicalSpecification limit for full power operation), or 60 times the primary coolant total iodineTechnical Specification concentration.The accident initiated iodine spike activities are based on an accident dependentmultiplier, times the equilibrium iodine appearance rate. The equilibrium appearancerates are conservatively calculated based on the technical specification reactor coolantactivities, along with the maximum design letdown rate, maximum TechnicalSpecification based allowed primary coolant leakage, and an assumed ion-exchangeriodine efficiency of 100%.The Technical Specification secondary liquid iodine concentration is determined usingmethodology similar to that described above for the primary coolant where CTtot iS15.5-1615.5-16Revision 19 May 2010 DCPP UNITS I & 2 FSAR UPDATE0.1 pCi/gm DE 1-131, and C(i) is the design basis secondary coolant equilibriumconcentrations per nuclide.The Technical Specification noble gas concentrations for the primary coolant are basedon 270 pci/gm DE Xe-133. The DE Xe-I133 for noble gases is calculated as follows:DEX133 .=2{F(i) x C(i)} (15.5-2)Where:F(i) = DCF(i) / DCF Xe-133DCF(i) = EPA Federal Guidance Report No. 12 (Reference 42) Table II1.1, DoseCoefficient per Nuclide [(rem-m3)/(Ci-sec)IC(i) =design basis primary coolant equilibrium noble gas concentration per nuclide(pJCi/gm)The noble gas and halogen primary and secondary coolant Technical SpecificationActivity Concentrations for Unit 1 and Unit 2 are presented in Table 15.5-78. The pre-accident iodine spike concentrations and the equilibrium iodine appearance rates(utilized to develop accident initiated iodine spike values), are presented in Table 15.5-7915.5.3.1.3 Gap Fractions for Non-LOCA EventsRegulatory Guide 1.183, July 2000, Table 3 provides the gap fractions for Non-LOCAevents that are postulated to result in fuel damage for AST applications. Thereferenced gap fractions are contingent upon meeting Note 11 of Table 3 of RegulatoryGuide 1.183. Note 11 indicates that the release fractions listed in Table 3 are"acceptable for use with currently approved LWR fuel with a peak burnup of 62,000MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3kw/ft peak rod average power for burnups exceeding 54 GWD/MTU." The burnupcriterion associated with the maximum allowable linear heat generation rate isapplicable to the peak rod average burnup in any assembly and is not limited toassemblies with an average burnup that exceeds 54 GWD/MTU.DCPP has three design basis non-LOCA accidents that are postulated to result in fueldamage, i.e., the Locked Rotor Accident (LRA), the Fuel Handling Accident (FHA) andthe Control Rod Ejection Accident (CREA)To support flexibility of fuel management, and establish dose consequences that takeinto consideration fuel rods that may exceed the Regulatory Guide 1.183, Table 3, Note11 linear heat generation criteria, the fuel gap fractions provided in Table 3 of DraftGuide (DG)-1 199 (Reference 62) for all No n-LOCA events that are postulated to resultin fuel damage with the exception of the CREA. This approach is acceptable (i.e., inlieu of developing plant specific fission gas release calculations using NRC approvedmethods and bounding power history to establish the gap fractions), since DCPP fallswithin, and intends to operate within, the maximum allowable power operating envelop15.5-1715 .5-17Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEfor PWRs shown in Figure 1 of DG-1199.In summary, the fuel gap activity fractions used to assess the dose consequences of the FHAand LRA are as follows:FHA /LRANuclide Group (based on DG-1199)I-131 0.081-132 0.23Kr-85 0.35Other Noble Gases 0.04Other Halogens 0.05Alkali Metals 0.46In accordance with Regulatory Guide 1.*183 (Appendix H and Note 11 of Table 3),the gap fraction associated with the CREA is as follows:Noble Gases:Halogens:10%10%Refer to Tables 15.5-80 for the isotopic concentrations in the gap assumed for theLRA and CREA. The isotopic concentrations assumed for the FHA are presentedin Table 15.5-47C.15.5.3.2 Tank Rupture EventsActivity inventories,, in.. variou.s ...,waste,..' .... tanksused for the tank rupture eventsare alse-4istedprovided in sections of Chapters 11 and 12 and witi-beare cross-referenced in the sections of this chapter dealing with accidental releases from thesetanks.I 11c'TrDI' 'BtAI ATIf'KI IlI ITAI IC'O DEIl CxAI kIrT DE Dr\I[C'rrRefueling shutdown studios at operating Westin gho uce P WRs indicate that, duringO ......... .... .nrc.................................. .. .............. ..nc-i i- .....h a b nf p -i -,c-i 1441 .5 .I SI.~1. L~IEd 4.1..... -I ........ 1....... S..........I............ ............... ... .... ................c " ~,V I,4~ r~ vr ~ r'c~d i- c-,s-~,c-A, ,rrr. i-I, a Ann c-nc-c., ,r, c-ni-inn cf i-Mn DC' 0 c-sc- i-Mn c-nc-id! .f nn-,nn ;An ni- anti c-Mci, IAs ALI I EI~LI, 5454f47 54.41.1541 I~.LJtI.dIl 541 IEIS. I s.~ 54.4 LII'.4 I 541.41.411. 54I 541 I ,.454.,I5454I IL, I.AI ISA ,.41i54L4I1.Atherefore be taken into account in the calculation of post accident re!~a~~~ of primarycoolant to the environment.Table 15.5 1 illuotra toe the ant'~ipat~d coolant acti~'ity increases of ceveral isotopes forF~iDO A. Ic-icI.o c'M~ li-An,,,,., TMc- #obIa Ic-.,Cc, i-Mn avc-sac-.i-nA pc--i-A ,,~i-;nc' A, ,c-hr, c-i-anti,, c-i-ni-nI 54LIIII I~J 4111.41.1.454 ill I. I I ISA LI 5411.1 Il1.Jt1.J LI~ 54 54fl1.454541.541.j 1.1541.1W ILl 5454 541.541.5.41.54nnnnnnnAnn#n~n..,l-.,sA nnnInn,,,i-mnc-A,,nnnl..,ni-nnnlti-.,,,,nncn!,nnc- TA-sac-n An!-,c- c-. .i -- s .-- sscc-i--finn-sm .-scn ,snc--shn,- OIA/IZ i/'-"J ics c-u-i~l-sc ;n HAc,;nn i-n i-M1.4I 54 I.J1.A1..ll.IS4 *.JII III 545454I4I541II54I 11.54 11541$I tAll S..If.JI..lI 1.4 LII I~I **IS LII 1.41. 154 541111111.41 III SAl..'lJI~j II 1.54 (II 11.115.5-1815. 5-18Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEDCPP and' has oper... "'sigonificant÷ f,,ol defects. "The me......d act,-iit. leel f.. orf.thc opratingflr p~lant also inc~hrluded4 in 15.5 1.dep -rnss,,r;-,at,-,, is 131t.! Tlho lvel in t he coor.lant÷ wasc obso.,n'or to be hig.her"ther ratec varyng betwn approx,'~nc',imatly The ......o soc.... ,- ,-;,; ..tou .leso in.. agni...doter fsir,- onb p... ÷.domir~'rrnoraizor dulring ln olor--Frsenga dt ;romoprati ngL p~ nts tindiat ha maxrrml m ino;~rctr thas of~c approximatlytcoldwnan dprsurzai5pocdue Alhuhastaysaevisio 133 a 21 DCPP UNITS 1 & 2 FSAR UPDATE15.5.4 EFFECTS OFl PILUiTONllIUM INVIENJTORv rON~ AC'IEn~kTDOSESDELETEDHISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.thrml isios nPu 239, estvt td aodce to d 2er3n th" osi15,5.51 Deofthsig Bffsit Anpotentsa (Ecluding donsRptue.This EtAd dand ther7atmosptht erc~d disperso fartonlys!g (t/Q atilzed by the dspluonieumenceantories. Thae resutng diferenesd listingi TabulaoGie 15.542, ievisaon thamthyoiddosesge nde otnos eprally irraefom4t eprcsentandv wholea berody dosf geourllymecreorosogcdat from2t prete assumingt mteoraoiogita tocred Jatur 1,20.locathios study, toas core fisont yelromads wer avb calculated byaueswihing ofhe23fissionspheric Reandv PuO23enfissionyils.i Becusedthe cores(AON6 messhofdo235giconsieranly g1.Reater the thectr aso u29 oa oefsion 2.3.d5arcclo2.tU23 of thereleas .Thin mass esepof Uoc238n anepovdd u21ta isin aigre extremewhly smabll,and14 thusd 2.3816 adpur41hvie essenmtialynoeft on the teesoitalecepore fissionationds.15hat. Designauaed Basies Accident (Ecldng Tank14 prupdetures)nrlromQvlefrthe EAindivda thelPZatmsepheictdispeptrscmiaionsfacors QUtilizead inith dorspecieyconsequence4analysens thae been dvelueopted uingiRegulat ory-LGuide1e15asevisoint 1Smeepthooogy andbiatontinuous Untempoally retapresabentative 5-yea perodma ofak hourlyeUsingeo the TSam houndrly meteorologicavl dataidthed alue appliabevlue tor ponstentaSAllnfilthered nleasae pitadrcpo locations aronhnelpovied. ine Fnigur 2.3-5, while2 cotablereporm combsuinations for Uintak1 ando Unieth2 applCableito the TSeomalgintak ande Thes15.5-2015.5-20Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEthe z/Q presented in Tables 2.3-1 47 and 2.3-148 for the control room pressurizationintakes inclusive of the credit for dual intake design and ability to select the morefavorable intake are also applicable to the TSC.Note that the specific control room x/Q values used in each of the accident analyses(and the specific TSC X/Q values used for the LOCA) are presented in the accident-specific tables presented in Chapter 15.5. The x/Q values selected for use in the doseconsequence analyses are intended to support bounding analyses for an accident thatoccurs at either unit. They take into consideration the various release points-receptorsapplicable to each accident in order to identify the bounding y/Q values and reflect theallowable adjustments and reductions in the values as discussed earlier and furthersummarized in the notes of Tables 2.3-147 through 2.3-149.15.5.5.2 Tank Rupture EventsFor the analyses of offsite doses from the DBletank rupture events, the rare andunfavorable set of atmospheric dilution factors assumed in the NRC-Regulatory Guide1 .4, Revision 1 (Reference 6) was used. On the basis of meteorological data collectedat the DCPP site, these unfavorable dilution factors, assumed for the design basescases, are not expected to exist for onshore wind directions more than 5 percent of thetime. The particular values used for this site are given in Table 15.5-3.Eforect anarlyeaseo odurateoo dowenwfrd theropcnd cavel accientratinhae asuednamoephuric direution facddtorermisted thoeincabley 15.5 4.norwtedge cates 10preonta ofdthvestignl bsspectr se nfumbuerscerohte u bsed.Onthatis ofstd o theor siterdatlyage tatBecaushe hofizothelo coprobabilts of ocurruence nassgoiatd ihtesel ot~ assme diluti~cntpmossible emergeny ieracuatione tand largew saratondss in conenutrtonsy due atoedwnin15.5-2115.5-2 1Revision 19 May2010 DCPP UNITS 1 & 2 FSAR UPDATEdimension of the cloud need be modified for concentration estimates for noncontinuousreleases. Slade (Reference 7) using the approach recommended by Cramer, gives atime-dependent adjustment of the lateral component of turbulence to be:= Ge (To) (T/To)°2(15.543)where:ce (T) = lateral intensity of turbulence of a time period T,where T is a value less than 10 minutesGe (To) = lateral intensity of turbulence measured over a timeperiod T0, where To is on the order of 10 minutesNear a source there is a direct linear relationship between GO and the plume crosswinddimension G-y so that the Gyversus distance curves presented by Slade can be directlyscaled by the factor (T/To) " to provide estimates of a reference Gy at about 100 metersdownwind from the source. Beyond this distance, the lateral expansion rates forcontinuous and noncontinuous point source releases are approximately the same, andthus the ratio of short-term release concentration to continuous release concentrationfor point sources is independent of stability class, downwind distance, or windspeed.For distances less than a few thousand meters the ratio approaches unity as the volumeof the source increases.Using the above scaling concept, the dilution equation in Regulatory Guide 1 .4, and thecloud dimension curves given by Slade, the ratio of short-term release concentration tocontinuous release concentration was calculated for several different release durations(Figure 15.5-1). For a 10-second duration, the short-term dilution factor is only2.3 higher than the continuous release dilution factor, and thus the appropriateshort-term release correction is within the uncertainty limits of the continuous releasedilution factor.The variou.s .,,, .,,.,. , plntacidnt consdere Sections,. 15.2,o 15. !3, and 15.1 may resu.,lt. in.acuvity release lnrougn various painways: conia:nment leaKage, seconoary sieamdumping, vontilation discharge, and radioactive waste system discharge.Post accident containment leakage is a slow continuous process, and thus continuousrelease dilution factors apply for these cases.Because of secondary loop isolation capabilities and because significant activity relea~is accompanied by large steam release, secondar; steam dumping accidents releasesignificant quantities of activity only through relief valves. Relief valve flow limitationscombined with large steam release result in activity releases of long duration. Thuscontinuous release dilution factors apply for these cases.15.5-2215.5-22Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEof-The release duration for liquid holdup tank rupture, gas decay tank ruptureT and thevolume control tank ruptureadfe handling...... are accident... are all over inless than 10minutesT-,-. As discussed above, continuous release dilution factors apply for thesecases=Contnuou, dIlution factor- hav ben app~lied to a1l Conditions II, IIl, and I110 CFR Part 100 limitsShort ter releas dilutionh,. factors ar.... only about twice as. high .... continuou.. relea.edilutiorn factors+rin the contnuousn~ releaset dilutio~n Furt,-hermore, theaboe- reason... that and a more sophisticated or complexshort-term release dilution model is not justified.Th topei diserio fatr for praessurization and in~filtration air flows ._to the:rooxm are analyzedq using the modifierd Halitskyx whic-h isa result of the TM!, accident the NRC, in NURE 073_n7 Secion III 1"3 .A, asked all1nucflear plants" to their post LOCA contfrol roonm habitability designs usi~ng(7/Q) nmethodology reco~mmended in the NI C paer was ronseRantiv;a ndinappnropriate for most of the plant design. The 5/I C equat-io~nsae brashed primarily onthe Hali+s,, data for round topped. EBR-!! (PWP. tye c... ntinment. and ar va...lid onl,,Historically, the, preliminap' building wake ..s wasL,. on ..a series of;' .and Atomic- Energ'1' 198 D. H.1 Slade Editor 7). In 19'71 K." Mrphy and15.5-23 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEactual bull4 ding. wake measurements+,r haveg been at' Rancho cylindr,',cal containment... Inilrtin.i into,. the contro room,.. would come fromabov t.. ,he highest:- roof elvaion. +' of the ....ilar,', Pres....urization,,. air. for.,, thebuldn roofI ... and a., portion,, of the,+ turbi;ne builing;,, Wall. f aci~ng ..... a the wall facing"'2.-/ "/Au ..... ... (15.5 2)A- sectionalara, e--u--in ped r/account for situation and plant specific features:* Stream line flows are used in most wind tunnel tests* Release points are generally much higher than 10 meters above ground* Null wind velocity is obsen'cd at certain periods of time* Isothermal temperatures are used in wind tunnel tests* Buoyancy and jet momentum effects are ignoredTvoical 1 hr field tests account for olume meander effects. while 3 to 5 minute windtu nl tests. do. no.. .. i -..15.5-2415.5-24Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE-/ K x fxf (15.5..3)This, ,.,,dificd, H-alitsky,, methodolog isv: inhcr.... cos.......... be,, cause... the .ind is,4;as.umed to be in t..o;ward.... the cont,,rol room dur"ing first or..wor.t part of theacc idn,,- and. becas eretwndsed aeue;rte ha 0prcn.additon, te adjstmen factrs ar alwas biaed toard the ini rdcintafTs-f chc,-f. d e to"+'< tnhei uncert.inty, wr toth 1hor.ied e, x<.atm asured-1;int some cas-esigtht was-! rsigiicnly higheorsecn.~ n hnawn !tThe .hoic of Kators and. he suggste modifying factors t÷ , ,; etc. are. dsussed,,be~Ow,K- a Jl; ... .. ;ool.. ... s*r k aeth!xQ estimate..r to. bevaid... iTh ky in Reference,. ... ha severa sets, K iopethfr run tppe.cntanmnt (frRev)an lokbidiongs9 (for01 DCPP UNITS 1 & 2 FSAR UPDATECa.se K 1- t Bas Q Pressu riza:tio n 4 1 3690 1.084x! 04Infiltration 5 1 1661 3.0!x!04tr7-ind-peedr'nntiinm~nt Ar hiuilclrir-i T rafv-L~ fl~a Nil C' nr~u-r'ant uuuir~ eu-.~a~rI .~.-..--...-,..~.4A rna+,-~.- I.uak4 rd.au IA ka 4aA +a +k.~ ',,4, alI VVI IA a4 fka -.,-~-.+*.....~,..'ar rala',ea n~nG Tha C nr. in,r.

i,,, A n r,,-*r*A n a r.1u ~ .~ ; p, ti 4k a ~r. in,.,~.. In4~r.t(,,wujp uL,,.r,., L¥ y Vll um.,Jul Il LAL- *AL t~-,,L tIILAW*1..I-( z "~ ,~,I',t I C A\T I~ZRef Iwhe rc:-u-n-- wind speed at height -z~-X l *./ .'.3-- Kel,T -ha ... ,r r.+ h ,, -r~,m.. .h ... ,,, In aa pi- A T a , , a ks:A a l~,~ ,nnI I' tL Jt. *'~ t A A~'~tmeteorological data for a 10 year period of record.0 8hhrs1 .0 1--.-08 24 hrs 0.83 ----0.9296 720 hrs 0.48 0.---Q7f_ in f"rh!.r' -.-.Wilson in Reference 30 and field tests confirm Halitsky's statement that hisK icoeleths are a factor of 5 to 10 too conservative due to not accounting forrnn~r'irn thtfuufu,-+ranc.n rt'ha .,nnnr~ro.hrmn tha= he1IAt;nn Tharafarar -, far-fa~r ,--,u',Oi i ILJ1m ul Lli'u*1 u~lt~utL LLfl0.2 vv~as u.sed,+,,.. foa [.rLii iii'.., i b';f fllfltfl#flA raIn an4'ani~PAl PAJm~C*t~tI~r r4 ii (RAf~rAnm 31Yinc1in~it~ thM thr~rA r~rn nfl tA 10 null ~"indc.raarl ,.,-,nA;+;anc- ,-Juu rnn ., nhau ur ni A.,f-, r'rIIar'tian flu irnn *haea narnAr. +kauJS.sA'.IStSAiISA ILi*tt iSp S4%Ai i* SI LAu S I SttAu Vu'.AVLtAL*.JilS..~..LiS S. L..LAI Eu u,.4 LE u....J*.. L.'St~S.L4VJ LuII i II + i I I* IIenct a etmoeu,m pume rise D,,,a,,y .. ou...' result in 15.5-2615.5-26Revision 19 May 2010 DCPP UNITS 1 & 2 ESAR UPDATE2 th wk ca ny ,A reduction facto-ra of 1' wa.. C_ 1556 AE (15.55) NHLAIO15.5.6.1D sig -protti p A cci entratxluiongTn RutrsThe brathin re moed ntel cn cntations fihlto oe r itdiTabl15.-7A prt otes e vaus amplingd tim e aeaediybetigrtspo einSetormn wind3 tunnegldatar Gisdaen for83.t 0mnt ape.Tufra1huCmA1aueo.55-a27neraivl asviumed19for 201 DCPP UNITS 1 & 2 FSAR UPDATE15.5.6.2 Tank Rupture EventsThe breathing rates used in the calculations of inhalation doses are listed inTable 15.5-7* These values are based on the average daily breathing rates assumed inICRP Publication 2 (Reference 8) which are also used in Regulatory Guide 1.4, Revision15.5.7 DELETED POPULATION DISTRIBUTIONdist.ribu tio u...ed, is te Table 1 5.5 8. The actual post accident÷ population;,ionlr bet' s'l!ignif'ican'lfr.3rtly llowel l~r a'nyl evacuatlion' ,re 15.5.8 RADIOLOGICAL ANALYSIS PROGRAMS15.5.8.1 DESCRIPTION of the EMERALD (Revision I) and EMERALD-NORMAL(Tank Rupture Events)-P-!egfamEMERALD is used to develop the source term for the tank rupture events and assessthe dose consequences at the EAB and LPZ following a waste gas decay tank ruptureand a volume control tank rupture.The EMERALD program (Reference 4) is designed for the calculation of radiationreleases and exposures resulting from abnormal operation of a large PWR. Theapproach used in EMERALD is similar to an analog simulation of a real system. Eachcomponent or volume in the plant that contains a radioactive material is represented bya subroutine, which keeps track of the production, transfer, decay, and absorption ofradioactivity in that volume. During the course of the analysis of an accident, activity istransferred from subroutine to subroutine in the program as it would be transferred fromplace to place in the plant. Fo e...mple, in the ca,,-,lculation of dose r.sulting a..then,., releas-ed...-. ... to.. the, .. tmshere,. The rates of transfer, leakage, production, cleanup,decay, and release are read in as input to the program.Subroutines are also included that calculate the onsite and offsite radiation exposures atvarious distances for individual isotopes and sums of isotopes. The program contains alibrary of physical data for 25 isotopes of most interest in licensing calculations, andother isotopes can be added or substituted. Because of the flexible nature of thesimulation approach, the EMERALD program can be used for most calculationsinvolving the production and release of radioactive materials, including design,15.5-2815.5-28Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEoperational and licensing studies. The complete description of the program, includingmodels and equations, is contained in Reference 4.The EMERALD-NORMAL program (Reference 5) is a program incorporating thefeatures of EMERALD, but designed specifically for releases from normal and near-normal operating conditions. It contains an expanded library of isotopes, including allthose of interest in gaseous and liquid environmental exposures. Models for a radwastesystem are included, using the specific configuration of radwaste system components inthe DCPP. The program contains a subroutine for doses via liquid release pathwaysdeveloped by the Bechtel Corporation and a tritium subroutine. The code calculatesactivity inventories in various radwaste tanks and plant components which are used forthe initial conditions for accidents involving these tasks. In addition, it is used in somenear-normal plant conditions classified in this document as Condition I and Condition IIand discussed in Chapter 11.15.5.8.2,.,D,.o.cc of, thc LOCADOSE-P-FegramThe LOCADOSE program (Reference 47) is designed to calculate radionuclideactivities, integrated activities, and releases from a number of arbitrarily specifiedregions. One region is specified as the environment. Doses and dose rates for fiveorgans (thyroid, lung, bone, beta skin, and whole body) can be calculated for eachregion, and for a number of offsite locations with specified atmospheric dispersionfactors. The control room can be specified as a special region for convenience inmodeling airborne doses to the control room operators.LOCADOSE is also used to assess the dose consequences at the EAB and LPZfollowing a liquid holdup tank rupture.15.5.8.3 nELETED:nn....ript,., of, ORIGEN_2 Program...The core inventor; and gamma ray energy spectra of post accident fission products forselected accidents (See Section 15.5.1) were computed using the ORIGEN 2 computerprogram. ORIGEN 2 (Reference 50) is a versatile point depletion and decay computercode for use in simulating nuclear fuel cycles and calculating the nuclide compositionsof materials contained therein. This code represents a revision and update of theoriginal ORIGEN computer code which has been distribut ed world ~A~ide beginning in theearly 1970s. Included in it arc provisions for incorporating data generated by more~ ophisticated reactor physics codes, free format input, the abili~' to simulate a widevariety of fuel cycle flowshcets, and more flexible and controllable output features15.5.8.4 DELETEDDescription of the ISOSHLD ProgramISOSHLD (Reference 9) is a computer code used to pe~orm gamma ray shieldingcalculations for isotope sources in a wide variety of source and shield configurations.tt~nuation calculations arc performed by point kernel integration; for most geometries15.5-2915.5-29Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEthis is done by Simpson's rule numerical integration. Source strength in uniform ore~nonential distribution (where ~nnlicable~ m~v he ealctthted by the linked fis~on-----------\~rE.I~-----------sour...ce..... and points....;"<, the effective number pa-rt ..,.,sh;ied las unes ote~isechsen, nd hepoit iotopi Ncler eveopentAsocite15.5.8.5 Description of the ISOSHLD II ProgramISOSHLD II (Reference 11) is a shielding code that is principally intended for use incalculating the radiation dose, at a field point, from bremsstrahlung and/or decaygamma rays emitted by radioisotope sources. This program, with the newly-addedbremsstrahlung mode, is an extension of the earlier version (ISOSHLD). Five shieldregions can be handled with up to twenty materials per shield; the source is consideredto be the first shield region, i.e., bremsstrahlung and decay gamma rays are producedonly in the source. Point kernel integration (over the source region) is used to calculatethe radiation dose at a field point.ISOSHLD II is used to determine the dose to the control room operator due to directshine from the airborne activity inside the containment following a LOCA during dailyingress / egress for the duration of the accident.15.5.8.6 DrLE/TEDpescri-pt-in of, the RADTRAD Pogr...mORIDTR-D (Referenc te 52 L uses a somintio of tables andch numericalmodels byofigsoreN emaeutionalphenomenarto(determine theCtierome dependarientdospuer ataluser ospliedsn evlocations.fo SAS give accidntro scnro.ul thals provides ah invuenetorcadecaechainucanddoe covnversion factorl taseblesb neededg foriteouse calulaton.os Thetocotrolaromn aouls elas sthe bxouential doeandi to -esltimat thmosue attGENuatondu15587SAS2 / ORIGEN-S(Rfrne6)aluaethtmeeenetnurnlxanthNationldLabofislrato-ryOnLum forlthes NRCto perfoprmy sacondadie oamputer anaclyesafrinteractions including fission, activation, and various neutron absorption reactions. It15.5-3015.5-30Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEcan calculate accurately the neutron-activated products, the actinides and the fissionproducts in a reactor core.SAS2/ORIGEN-S is used to develop the equilibrium core activity inventory and thedecayed fuel inventories after shutdown utilized to assess the design basis accidentsexcluding the tank ruptures.15.5.8.8 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.1ONEXCHANGER 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-i113 Atmospheric Dispersion Factors (Reference 73) calculates z/ 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 0.173 hours 0.00103 weeks 2.37432e-4 months 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-i113 performs the same calculations as the NRC PAVAN code15.5-3115.5-31Revision 19 May 2010 DCPP UNITS 1 & 2 FSARUPDATEexcept that EN-I113 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-i113 is used to develop the DCPP site boundary atmospheric dispersion factorsutilized to assess the design basis accidents excluding the tank ruptures.15.5.8.11 AROON96ARCON96 (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 0.2 hours 0.00119 weeks 2.7396e-4 months 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 NAUN/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.15.5-3215 .-32Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.8.13 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 othercompartments. 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.15.5-3315.5-33Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEPERC2 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 SWV_QADCGGP 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.15.5.8.15 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 nmulti-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 interfac:e 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 up15.5-3415.5-34Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEof 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 1 54'-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 roompressurization 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 for15.5-3515.5-35Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATEelemental 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 therecirculation 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 in leakage into the control room during Mode 1 and Mode 4 is fid4inassumed to be 70 cfm Table 15.5 86 and (includes 10 cfm for inleakage due toingress/egress inekaebased on the guidance provided in SRP 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. Thiis 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.15.5-3615.5-36Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe 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.informatio pre...iousl.. h in, thi section" has bec mo. vc..d, to Section 15.5.8.1.15.5.10 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 1-00.41-50.67, 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 t-he-two,, hours. any 2-hour period following the onset of the postulatedfission product release shall not receive a total-radiation dose *".. ,,,,,,,in. e..cess of 25re o a... total1 radiation,, dose in e..ce.. of 300 remn to, the thyroidfrom.iodin e, ...uren 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 ...hole, body, in excess. of 25 rem, or atotaexcess of 0.025 Sv (2.5 rem) TEDE.15.5-3715.5-37Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEControl 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.15.5.10.2 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 Condition15.5-3815. 5-38Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEII events, the scenario outlined below for the LOL analysis is based on the conservativeassumptions outlined in Regulatory Guide 1.1 83 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.Computer 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 p#Ci/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 j..Ci/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 pCi/gm DE 1-131 coolant concentration. The durationof the assumed spike is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 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 1iCi/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 and15.5-3915.5-39Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEiodine 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% elementaland 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 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months , at which time shutdown cooling is initiated viaoperation of the Residual Heat Removal (RHR) system, and environmental releasesare terminated.15.5.10.2.30Offsite 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 z/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 Assessment_The 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 control-room 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 Factors15.5-4015.5-40Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEDue to the proximity of the MSSVs/1 0% ADVs to the control room normal intake of theaffected unit, and because the releases from the MSSVs/IO% 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/10% ADVs (refer toSection 2.3.5.2.2 for detail).The 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 z/Q values presented in Table 15.5-98 take into considerationthe various release points-receptors applicable to the LOL to identify the bounding z/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 2.3-1 47 and 2.3-1 48.The bounding Control Room dose following a LOL event at either unit is presented inTable 15.5-9.and the iodine ..oncentratio;n in the stea generator water.. prior. to the accden. of$ thesem keyt parameters; the rmeult~ presented] in Figuvrme "15. 2 t lhroug",h155.As hown^n on the figu.r~e, the potential thyroid doses aehigher w^ith inc-reaingstea releases and.. iodine concent.ratio~ns. Fiues 15... 5 2 15.5 3 are result t+ hatfasum R -]egulator/_,; Guide A, 1, assumptfions fo pest. acc-r-t'ident meerology.I..andbrathngrats DesgnBass seAsumpios) ,, s shown rles in,'Figure T, 15.5,2,roxmt ! 1 .6 Ibm o~f steam is the ma,.ximm sta.rl.e..ete o aflcooldown.. without an... codese availability;H, and asemrlaeo prxmtlIbm would result from. releasing only... the .. contents,-,of one' steam generator. due,, Cond~ition II events. T~he hig~hest antiripafed doses would r~esult fro~m an event uhalOSS of elecricar-l load, andl the ptntian[l thyroid and w.hole body, doses fromtnfhis..reeae to. t he atmospher the.. first 2. .hourS,ananditol1,300IbII eve.,nt rel,'eaes.% assumptio-ns;used ,,,fo r meteorology, parra aphs.nn that;, thet preceding, stemnn "Freleas e~e quantities are assocriatd ihr horiginal steam gnenrator {(OSG)' loss of loadJP (LOL nalysis which providles the basis for,15.5-4115.5-41Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE1,023,000 Ibm, respectively) and are thercfore bounding since total dose is propo~ionaIto total steam release.For the design basis case, it was assumed that the plant had been operatingcontinuously with 1 percent fuel cladding defects and 1 gpm primary to secondaryleakage. For the expected case calculation, operation at 0.2 percent defects and20 gallons per day to the secondary was assumed. In both cases, leakage of waterfrom primary to secondary was assumed to continue during cooldown at 75 percent ofthe pre accident rate during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at 50 percent of the pre accident rateduring the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These values were derived from primary to secondarypressure differentials during cooldown.It was also consen'at~vely assumed for both cases that the iodine padition factor in thesteam generators releasing steam was 0.01, on a mass basis. In addition, to accountfor the effect of iodine spiking, fuel escape rate coefficients for iodincs of 30 times thenormal operation values given in Table 11.1 8 were used for a period of 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months sfollowing the stan of the accident. Other detailed and less significant modelingass umpt~ons are presented in Reference 1.T he resulting potential exposures from this type of accident are summarized inTable 15.5 9 and are consistent with the parametric analyses presented inFigures 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 w..hole, body, and to" the thyoi o.. an indviualocated at any point on the boundary of the exclusion area for the twe-hea-rsany 2-hour period kneiaeyfollowing the onset of the postulatedfission product release is within 0.025 Sv (2.5 rem) TEDE ...... o g, ;, ioarii,--, ......asshown in Table 15.5-9.(2) The radiation dose to the w..hole, body4 ..nd to the, tkhyroid;, 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.. e-.......... 4as shown in Table 15.5-9.(2(3) The radiation dose to an individual in the control room for the duration ofthe accident is within 0.05 Sv (5 remn) TEDE as shown in Table 15.5-9.15.5-4215.5-42Revision 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 400.&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.product,,. release .hall not, recei. a,= radiaton, dose to, the ...hole body infromiodie exp.ure! n indiviodua locate at+ an point. on... the boundar ofhe ÷ .. .,15.5-4315.5-43Revision 19 May 2010 DCPP UNITS 1 & 2FSAR UPDATEbody or its equivalentskin, Reference 51) fctoayp~o h oy (i 'e., 30,,rem thyri betatI-~H~uuwuun ut w ..... .... .... ...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.de.iled. escr.pion. o the... .model..... used.i calculating the fromSectio,-n 15.5.1"7 o-f this .o... The specifi assumption.. used..., in the analysi r.... aS-15.5.7, resectvel. common.assumptions ..re. described in thepreviou of 15.5.(2) It haso been ... assme that* all of* the .ater cont..aie in RCS is releasedto th containment. For the, desig basi ce the reactor colantpercent deeciv claddng" were..... use., These.. acivities and concentrationsi I II II I I Im used in aetermining tnese values are aescriDedi n section 11 .1.(3) Of the amounts of noble gases contained in the primar,' coolant100 percent is assumed to be released to the containment atmosphere atthe time of the accident. For the iodines, it is assumed that only 10 percentof the dissolved iodine in the coolant is released to the containmentatmosphere, due to tho solubility of the iodine. It is assumed that the15.5-44 Revision 19 May2010 DCPP UNITS 1 & 2 FSAR UPDATEamount. of" "odine;' in chemical fom t..hat are not ffected by the ..p..yreleased..'" .. , from the.,. fue,,'l, up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , after,, the accident,. i.";4" s assumed to." -""4* bereleased to t he containment.. Of the amount... of,, noble ..ases released to*-10 percen of; the. iodine relea sed....' tothe onainen are. reeae to theSection 15.5.17.(6) The containment lea.ag.rate n..this anlyisar also assume..d to be thes.me.as for thearg break F~ -,LOC (A and4 arc discus..ed i n Section 15.5.1"7.The resultin potenti ra*;l e..posures arc lIsted, in Table "15.5, 10 and demonstra,-te allca-lculatedl doses are well1 below the alues,, in, 10C 011 .SD the activity- relases.. from this typ of.. evn wil. ,, be significantly,,h low:er than thoe,, f.-romlarge break L-OA, any cntnrol room evxposure which might would1, be we~ll within15.5.11.3 ConclusionsThea nahlysisdemonstrates that the aceptarfnce criteria aJre met ase follows:e(1) The radiation dose to the whole body and to the thyroid of an indiv'iduallocated any point ; on the CF boud 0 of th exclusion T-area forth !0hurfollowin **t he Wonseto the,,4 postu~lated"*h fhrission producti relesar\Aelwtinteds imits of 10 CFR 100.11 as shown in Table 15.5 10.(2)The kradatIo doseA to the1 whole body, andn to~ rthea thyroid o an h indivIduaOnthreleases (durhing thnerentire operiodof itsprassge, arei wenlludthd thatte dose15.5-45 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEconsequences 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 100.11! asoulnd cox10 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.(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.15.5-46 Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEAn individual located at n, point o'n the boundalr, of the. exc,'lusio;n -are- for the An, indiidu4al located- at÷ any point~ on the o,,ter bounda, of the low, population -zone, w*ho(during the entire period of its passage), shall not receive a tota! radiation dose to thewhole body in excess of 25 rem, or a total radiation dose in excess of 200 rem to thethyroid from iodine exposure.wV15.5.12.2- 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 15.51-_84.2, 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 (ON BR) 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 1 5.5-34, the dose consequences at the EAB and LPZfollowing a MSLB is within the acceptance criteria applicable to the minor secondary linebreak.Th4,e-p si ble consequences. of this event, due to the releas of so..e steaOn 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 15.5.12.1.15.5-4715.5-47Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEO'n the. ba-'sis of the discus..ed it can be, concluded that th, potetia e .po.urefollowing.a.mi.or...co.d.......tem pie, pur ol be...n...gnificant.3 ....The, radi4ation,, dose to the bod,, and , to the thyoi of. an4 individual;4"' loc"ted at an..The.3. rdAtincoeptoathe wholerbdiandt h hri fa niiullctda npimlmntonted outrin boudr; lofdn. the loounltion zoeen whot is exosedin tror thersradiactises cludportnesutiongro the3. pstuated cofirm prtoductntreleas(duing theailoiaeontieqperiooeis passage),ra areisigu fcat oflasin erhownsi15.5.13.1 Acepticatnce f CrtiaussadAcdn ecitoFuel assembly loading errors suhall benprventedtby ladministatie prmoedfurlasmlesinoimplemenedduringiore, loading. In thel rounlikel eventutactur ait loadngerror mocreplts,analyses supongertingSchtion 15ad3n3 safll cofirm tatsml noevntraing mauatouradwiohplogicaonstequrngenrcshallnoccurlasead reutof lcrasdin herroruxs. i h ro eslsi cnfuel and core loitoncading erors such aflsse niheth inadvertentl loading oneofoefelasmleone 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.15.5.13.3 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.15.5-4815.5-48Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.14 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 a s outlincd beowo::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 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 EA8 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.,An individu`,,l at, an... pint on the, bounda`4 of. the exclsio area.... for, the. h,who~le body in excess# of 25 rem,. or a tota~l radiatio~n dosea in exces of 1300 remn to thetnytroia tor~n iodilne exposuc-ire.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 pumps15.5-4915.5-49Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE(RC Ps). 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 accidcnt has nqo significantts,-7The 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.15.5.1 4.3 ConclusionsOn the basis of this comparison approach, it is concluded that the dlose 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 .-edescribed finn Sletion 1h5.3.h1l demntrate thatnr ther arenosinif ,aefeth s o~f nthe CompeteLoss~f ofi n Colnt lweei nt.f Tereorethon the boundar' he excluso area lfor the immediately folownradi~oac ftiv cloudr resulting from the pnostulated fission producl~t relcase (during the15.5-5015.5-50Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.15 RADIOLOGICAL CONSEQUENCES OF AN UNDERFREQUENCYACCIDENT15.5.1 5.1 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 IlI 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 remn) 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) TEDEF.15.5-5115. -5 1Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEA .- " ,, I -I I --I- l *f II I............... ........ atan point o.n tne, couter; o'"r, .... exclus. are. whotlewhlebdyineces or a total' radiation, dose in excess of 300 remn to the.,.J15.5.15.2- Identification of Causes and Accident DescriptionA transient analysis for this unlikely event has been carried o'-tis 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 drt~ilred nf the~ nnfrnti2I ren,-irnnmrentn! of ... ....ini,,-Ih,;nn c{,'nm rm,,r,-nni-r ;Ic. nracra-far ;rn Qar',finr g fl I kaE',,' .......h...--exposures. it can, be concludeda tha,f alt~houg..h very.. ulikely, the occurrence.. of this15.5.1 5.3 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. Onv, the,,, of" ... .......... -.. ... ..-... .. .f ... .... .. ..AL A tI t -) ~ l I II .. .................i*.. J" I.. .IIVlIVV gl....... .... ............ ..... ...... F ......Additionally, the analysis demonstrates that the acceotance criteria arc met as follows:15.5-5215.5-52Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEThe radiation do"se to the ,whole, and to, thyri of. an ,- locatod,, at ..n.The radiatonoia donseqtoethes hofasnle bodoad tousthethroi cofnto indieiduly woctedrata san"ponot onee the douter boindas of the low popu !atio zo'-ne, ..ho w.50.67, ad to l theethrdoseaccetine cldreiuteing fom thegpostulateude fision prouct re00eased(durlingd thelentirperuodtofy Gitspsae, 1.83res inotsign ificant y a ddhrniTbes 15ndti5 1I. naisHwvr15..1 ReuaDIoLOGuiCAL CONSEUENCSecn121 OF Aul SiNGlemetto RoD CLSTERlos15.5.16e1 Aouiieteds cceptance C riteria of1CF506inaldsTh ailgclconsequencesaaye adtofna Seciongl r.4od clsergcontroly assdembly83withdrawaleshlntha o exedt wthe dos limitsrbait of 10cCRu10.11ce otlin toelsed ieo:0 nd willmet thefdsacetnecieiofRegulatory Guide 1.183, Julyostl2000EA and LZosshoutlinoteced bhelwcriegi aulatoed Gide 1.13bde hs, nth dspeiial addrseqecs Codtio Ih A ndwll scnrosboevrpiier Rgltor Gh oesaui e 1.83reotedion 1.2.1 , e, a ful smplementaction of0% AS lows ah iiicesetouiizph osedacetncbrieiao 10 CFR 50.67.i lldsthat foAvnt wniithual highter probabilpinty ofncurec tha ondayo thoe exlstdion Tablea 6foanReglaoury Guriode 1.183,heoneto the postulated EAfnLZdssshoudnpodut rexceaed thelcrtrioabltredeine Tabl 6.dithus, thse doenoneqence o 05Sv at. them TEABaDELZwilb(1) An individual located at any point on the bouteondary of the exluson araporulatinyzone, 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.15.5-5315.5-53Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEAn individual located at any point on the bounda~ of thc exclusion area for the ~ohours immediately following 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 total radiationdose in excess of 300 rem to the thyroid from iodine exposure.A~n individual located at any point on the outer boundar; of the low population zone, whc~s 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. in excess of 25 rem, or a total radiation dose in excess of 300 rem to thethyroid from 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 (RCCA) 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 of potential radiological" conse..uence. of accidents15.5-5415.5-54Revision 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 he ,6fo the potent hia exposure dicun d t h h,, it can be concuded,,. that++ the ...occurrence,, ofthis -a., cidenrt, ., would not,,Ic.au,, e undue #,-ri+kto the afety, of,-,th.!5.5 !2.ana.7R I LysIsA CONSEUENCE tht Fh cAJ RiRUTeRia are met MAs RYtheAN accetanc15.5.17.1 Acceptance Criteriaof a large break loss of coolant+The 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.15.5-5515.5-55Revision 19 May 2010 DCPP UNITSI1 & 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) Thc rad4;-iological con..equence. of a major... ruptur of primry pipecontanmen tn post LOCA' rec ...Ircultin LoopInL leakg in the Auviliary BuIdinglof. a residual heatJ reoam,-l (RHR) pu~mp seaol failu~re resulti+ng in a 50gp~m leak, for sta.rting. at T-24i hrs post+ LOCA),^ and containment.shall not e.ceed the dos limits of 10 CFR 100..1 a s outlined belo...:for the fl'"o hounrs immediately folwnthe onset of the: posulaed~lsfissio p.. ,roduc rele..se.shall not receiv a,= tota, radiatio;n dl'ose to the30 rem to the thyro'idflro~m iodnelh ii. An1 individulH O at any.. point{ o+nthe oute fr bound.r ol...f the lowppulationm-+ ..one, who is expose.d to the, radioactive.+ cloud resulting. r,-o~f its pasage,=-r, shall notf receive*+ ar totl dos tr he ,,ho..le.body4, in exces.. of 25 rem, or a total drose n excess. o-f 300room operator.. , under faccien conditions shall not be0 in excess.of 5rem ,whole or,, its equivah'.lent to+ any part of the body; 30' remthyri a..;,nd beta ski,n Reference. 51) fo~r the duration of the acideont.(1) In the eeant corntrolled4 ve+nting of the co~ntainment is implemented4 postcapabnhi;lt fo+r co~ntrol to the hydroge~n frecombiners,r an loatednf at anyl point on the bou"llil4Jndar of' the exclusoin whno is exosed15.5-5615.5-56Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATErelease (during the entire period of its passage), shall not receive a totalradiation dose to the whole body in excess of 0.5 rem/year in accordancewith 10 CFR Part 20.15.5.17.2 Identification of Causes and Accident Description15.5.17.2.1III~asc Ientsandkeicec -ract Release PathwaysThe accidental rupture of a main coolant pipe is the event assumed to initiate a -L--largebreak 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. Because.. of the..... prsuization, of the cotimn.ulin..te.UII wiu ~dJiyiimm iuuu~ *urii ... ui v,,iuui, iu,,, i.+ k:U~l ,,m-+ U +h-i .f;.,,., ,,+,.,,,-,t.,m ,II; mI d rsf ,-~,r ," t.,;n ii-.,s mci ",,d ,.,i Iikir.,+cIi A i ri m _ t;hcpx,z...rm.,ntc. r'on,1,i,'.f h"_ th,= .t.~. Thci frn.-,r.ticr ~.f th.-' frf.*"4iless, since the rate of thermal radiolytic decomDosition would exceed the rate ofOrganic compounds of iodine can be formed by reaction of absorbed elemental iodineon su~aces of the containment vessel. Experiments have shown that the rate offormation is dependent on specific conditions such as the concentration of iodine,concentration of impurities, radiation level, pressure, temperature, and relative humidity.The rate of conversion of airborne iodine is proportional to the su~ace to volume ratio ofthe enclosure, whether the process L limited to diffusion to the su~aoc or by thereaction rate of the absorbed iodine. The obsered yields of organic iodine as afunction of aging time in various test enclosures, with various volume to su~ace arearatios, were extrapolated to determine the values for the DCPP containment vessel.The iodine conversion rates predicted in this manner did not exceed 0.0005 percent ofthe atmospheric iodine per hour.The potential exposures following the postulated sequence of events in LBLOCAs have~tcicip .,~.,.,k,-,,..,4 fe-~~ h~in r1c~~~~- In +P~~ '~"n'~trd n'~'~ it ho" hn~~r~ g'~"ii'~'r'd th-~d thci p~tiri-~15.5-5715.5-57Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATEcons...a... s;,nei,:. the E-COS is designed to pre...ent gross cladding damage. Inzeo.n The paoicl-a-.,te~ fractioln of iodinei also assumed--,P~n to be zeor for the expected"case. since-. this. fraction-,- is small and the spra r..mova,,l rates. fo ....,-+,-.,at+. is large.ashown in Reference 10.Fofh einbssLCi a enasmdta 25 pecnffteqiiruRadilative Giodie in.ento,' inpteni cor isdmedtiaisthel larailbleforeakageC fro the dsgreactor caeontainent Ninetyu one perent oiesfo thiuat25 peretorssmaned tof beinatefroeementgatio iodines an phercntaotismnt 25d percentcisity stheg fomorglanie tordiooscandofP theolas idntvfentr sin atheitcoreleise pashsumedtown be immedAtlreasdothc .Rlae i h ontainment buligPsdicsenerlessre p/ragaphs, Relefasesa tof theemgiueaenvrnoet execedto l ocurheen ifnthinen ECsdoestnot palesr ase coexpce.Anaayiusingntthesenassumptionsoishpresentedmbecausetherevler contimnisoaidered accieptablfor. dSign basis eanalysi ino REgFssesthtrcrulator'uGuiew1terRvistond1basi easlue of the spectrump ofbeak size fo4rs revauating perorance ofm rleakfsemiigtonytem CPdsg nlds andhonanent, andmoshre failtyrstiong relatiem to fradoogicarglatrcosequncpin es. tnadRve lnScin1.65 pedxB(eeec 7,awePP has idenutifie s uix e activit releas 0 icuin tislaae paths flloindaseC1.oeleseqviacthesonotarqinment Prwessuer acuumte Rntefoloief patwaytions theRHenionetuni5hecnaimn 5ioaio5 ave8r cloed. in1 My21 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 mains. A review of the equipment in theRHR system loop and the 0S8 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 as~sumptions / 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.15.5-5915.5-59Revision 19 May 2010 DCPP UNITS 1 & 2 FSAR UPDATE15.5.17.2.2 Activity Release Transport Moe..pra .... ,,din Remo,.aleshe... conaincn ..pr.y syte (CSS) is, desied;.... in detail ..ong wi... "th a spe~rmayne...rate, for.organic ;,ddes was...assumed' to be 0.058 per hour*has° also bee assumed.., for the design' basis case, that the CSS has. no effect on theAILtho'-Ih a s'ubsen'-ent ssfe÷,- eX'-l!'Iat!Ol s"hr;-ed that the De~sign Case enefficient nf.... .------... .-"-.. .. ....J --. ---------.. .... ... .... .. ------.. .... ... .31 per" hour/ (f'"or260 ,-pm., spray. header flow) should, be,- reduce to. appro....imately,15.5.17.2.2.1 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 the* primary 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 acts release of containment air to the15.5-6015.5-60Revision 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 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (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 0.0005Lanthan ides 0.000215.5-6115.5-61Revision 19 May 2010

v t e Letter Sequence Response to RAI
TAC:MF6399, (Open) TAC:MF6400, (Open)
Initiation Request Acceptance, Acceptance, Acceptance Supplement Administration Questions 22 September 2015 23 September 2015 1 October 2015 7 October 2015 20 October 2015 2 December 2015 11 January 2016 17 February 2016 Answers 17 December 2015 10 February 2016 21 April 2016 11 December 2015 17 December 2015 17 December 2015 17 December 2015 17 December 2015 Results Approval... Other:
MONTHYEAR2015-08-31 [Table View]

ML16004A361

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