ML16004A361

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
Download: ML16004A361 (89)

Text

Enclosure Attachment 4 PG&E Letter DCL-1 5-1 52 License Amendment Request 15-03, Attachment 4 Diablo Canyon Power Plant Updated Final Safety Analysis Report Markup (For Information Only), Revision 1

DCPP UNITS 1 & 2FSAR UPDATE releases of radioactive materials to the atmosphere and (2) coping with radiological emergencies.

2.3.1.4 Safety Guide 23, February 1972 - Onsite Meteorological Programs An onsite meteorological monitoring program that is capable of providing meteorological data needed to estimate potential radiation doses to the public as a result of routine or accidental release of radioactive material to the atmosphere and to asses other environmental 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 Conditions During and Following an Accident Control room display instrumentation for use in determining the magnitude of the release of radioactive materials and in continuously assessing such releases during and following an accident is provided.

2.3.1.6 Regulatory Guide 1.111, March 1976 - Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors Annual average relative concentration values are used during the postulated accident to estimate the long-term atmospheric transport and dispersion of gaseous effluents in routine releases.

2.3.1.7 Regulatory Guide 1.111, Revision 1, July 1977 - Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors In accordance with the requirement of Regulatory Guide 1.145, Revision 1 annual average relative concentration values are developed for each sector, at the outer low population zone (LPZ) boundary distance for that sector, using the method described in Regulatory Position 0.1 .c of Regulatory Guide 1.111, Revision 1. This information is used as input to develop the design basis radiological analysis 7./ values at the LPZ using Regulatory Guide 1.145, Revision 1 methodology.

2.3.1.8 Regulatory Guide 1.145, Revision 1, February 1983 -Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants The method outlined in Regulatory Guide 1 .145, Revision 1, (with the exception of methodology associated with elevated or stack releases, i.e., Regulatory Positions C.1.3.2, 0.2.1.2 and 0.2.2.2), is used for calculating short-term atmospheric dispersion factors for off-site locations such as the exclusion area boundary or the low population zone for design basis radiological analysis dispersion factors.

2.3-2 2.3-2Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE 2.3.1.9 Regulatory Guide 1.194, June 2003 - Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants The method outlined in Regulatory Positions C.1 through C.3, and the adjustment factor for vertically orientated energetic releases from steam relief valves and atmospheric dump valves allowed by Regulatory Position C.6 of Regulatory Guide 1.194, June 2003 is used to determine short-term on-site atmospheric dispersion factors in support of design basis radiological habitability assessments.

2.3.1.710 NUREG-0737 (Item III.A.2), November 1980 - Clarification of TMl Action Plan Requirements Item Ill.A.2 - Improving Licensee Emergency Preparedness-Long-Term:

Reasonable assurance is provided that adequate protective measures can and will be taken in the event of a radiological emergency. The requirements of NUREG-0654, Revision 1, November 1980, which provides meteorological criteria to ensure that the methods, systems and equipment for monitoring and assessing the consequences of radiological emergencies are in use, is implemented.

Item III.A.2.2 - Meteorological Data: NUREG-0737, Supplement 1, January 1983 provides 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 Problems of Meteorological Measurements Programs Meteorological data that are climatically representative, of high quality, and reliable in providing credible dose calculations and recommendations for protective actions in an emergency situation, and for doses calculated to assess the impact of routine releases of radioactive material to the atmosphere are available.

2.3.2 REGIONAL CLIMATOLOGY HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.

2.3.2.1 Data Sources The information used in determining the regional meteorological characteristicsof Diablo Canyon Power Plant (DCPP)site consists of climatologicalsummaries, technical studies, and reports by Dye (Reference 2), Edinger (Reference 3), Elford (Reference 4),

2.3-3 2.3-3Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE 22.5° interval. The 1-year gap (April 1971 through March 1972) in the period of record, October 1970 through September 1972, resulted from an unauthorized bivane modification.

Frequency distributions of wind speed and wind direction classified into seven stability classes as defined by the vertical temperature gradient are shown in Tables 2.3-21 through 2.3-28. The column headings are labeled in terms of mean hourly wind speed in 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 22.50 intervals. Table 2.3-28 shows the number of observations in each of the seven stability classes (Pasquill A through G) for the period of record July 1, 1967, through October31, 1969, when the mean hourly wind speed is less than 1 mph. The wind data were measured at the 76 meter level, and the vertical temperature difference measurements are the 76 meter level minus the 10 meter level.

The radius of the low population zone (LPZ) at DCPPhas been established to be 6 miles. Cumulative frequency distributionsof atmospheric dilution factors at each 22.50 intersection with a 10,0O00-meter radius (slightly greaterthan 6 miles) for the period May 1973 through April 1975 are presented in Table 2.3-4 1, Sheets 7, 8, 9 and 10. Each data set used to compile the frequency distribution is comprised of averages taken over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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 a result, a single hourly measurement may influence the value of over 5 percent of the observations. Since overlapping means are used in the distributions, the data are not independent and no assumption of normality can be made. These data show z/Q estimates from the 25th through the 100th percentile levels for each of the averaging periods.

2.3.5.2 Design Basis Radiological Analysis Dispersion Factors 2.3.5.2.1 Exclusion Area Boundary and Low Population Zone Atmospheric Dispersion Factors Atmospheric 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 are applicable to all dose consequence analyses documented in Section 15.5 with the exception of the tank rupture events. The methodology used for the tank rupture in accidents is discussed in Section 15.5.5.2 and the associated ylQs are reported Table 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 computer a

program "Atmospheric Dispersion Factors" EN-i113 (Refer to Section 15.5.8.10 for 2.3-28 2.3-28Revision 21 September 2013

DCPP UNITS 1 & 2FSAR UPDATE description of computer program EN-i113) using a continuous temporally representative 5-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 the various averaging periods using hourly meteorological data related to wind speed, wind direction, and stability class.

Equations used to determine the %/Q'sare as follows:

X/Qi = .{(u)[(7t)(ay)(c*z) + (A/2)]}1 (2.3-7)

Z/Q2 = [(u)(3*t)(a'y)('z)]- 1 (2.3-8)

Z/Q3 = [ (u) (7;)(X,) (Oz)]- 1 (2.3-9) where:

/Q = relative concentration (sec/in 3);

• yz= horizontal and vertical dispersion coefficients, respectively, based on stability class and horizontal downwind distance (in);

u = wind speed at the 10-meter elevation (m/sec);

A = cross-sectional building area (in 2);

= (M)(ay) for distances of 800 meters or less; and

=[(M-1)(ary800rn) + a*y] for distances greater than 800 meters with 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 value calculated 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 from Figure 2.1-2, taking into consideration a 45-degree azimuth sector centered on each 22.5°-azimuth sector as described in Regulatory Guide 1.145, Revision 1, Regulatory Position C.1 .2. The EAB x/Q values for the radiological releases from each unit are conservatively based on the EAB distances from the outer edge of each containment building.

An LPZ distance of 6 miles (9,654 meters) is used in the analysis. The use of one LPZ distance in all downwind directions from the center of the site for all release points is reasonable given the magnitude of this distance relative to the separation of the release point locations from one another.

The containment building cross-sectional area along with the containment building height is used for the annual average x/Q calculations (used as input to develop the accident x/Q values at the LPZ using Regulatory Guide 1.145 methodology). The applicable methodology for the annual average %/ calculations is identified in 2.3-29 2.3-29Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE Regulatory Guide 1.111, Revision 1, Regulatory Position C.1.c (Reference 28). These annual average x/Q values are used to calculate the intermediate averaging time %/Q values for the periods of 2-8 hours, 8-24 hours, 1-4 days, and 4-30 days by logarithmic interpolation.

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 that are sufficiently high to escape the aerodynamic effects of the plant buildings;

• The distances from the Unit 1 and Unit 2 releases are determined from the closest edge of the containment buildings to the EAB;

• The plume centerline from each release is transported directly over the receptor; and.

• A terrain recirculation factor of 4 is used in the calculation of the annual average x/Q values o anid-ne-Rfadioactive decay or plume depletion due to deposition is not considered.

The highest EAB and LP7 x/Q values from among all 22.5 0 -downwind sectors for each release/receptor combination and accident period are summarized in Table 2.3-1 45.

EAB %/Qvalues are presented for releases from Unit 1 and Unit 2, while the LPZ ;(/Q values are applicable to both units. The 0.5% sector dependent z/Q values are presented with the worst case downwind sector indicated in parentheses.

2.3.5.2.2 On-Site Atmospheric Dispersion Factors The control room and technical support center %IQvalues for radiological releases from Unit 1 and Unit 2 are calculated using the NRC "Atmospheric Relative CONcentrations in 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); and various receptor parameters (e.g., distance and direction from release to control room air intake and intake height). Refer to Section 15.5.8.11 for a description of computer program ARCON96). -

A continuous temporally representative 5-year period of hourly on-site meteorological data from the DCPP onsite meteorological tower (i.e., January 1, 2007 through December 31, 2011) is used for the ARCON96 analysis. Each hour of data, at a minimum, has a validated wind speed and direction at the 10-meter level and a temperature difference between the 76- and 10-meter levels. This period of data is temporally representative and meets the requirements of Safety Guide 23, February 1972 (Reference 21).

The ARCON96 modeling follows the ground level release requirements of Regulatory Position C.3 of Regulatory Guide 1 .194, June 2003 (Reference 30) relative to 2.3-30 2.3-30Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE determination 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 release and receptor locations, initial diffusion coefficients etc.).

Releases are assumed to be ground-level as none of the release points meet the definition of an elevated release as required by Regulatory Position C.3.2.2 of Regulatory Guide 1.194, June 2003 (i.e., do not meet the requirement to be at a minimum 2.5 times the height of plant buildings).

Only the containment building edge releases are treated as diffuse sources as the releases occur from the entire surface of the building. In these cases, initial values of the diffusion coefficients (sigma y, sigma z) are determined in accordance with the requirements 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-site distances.

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; and Sector 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 control room or technical support center air intake/receptor (conservative);

oThe distances from the Unit 1 and Unit 2 containment building surfaces to the receptors are determined from the closest edge of the containment buildings and the release/receptor elevation differences are set to zero (conservative);

• The applicable structure relative to quantifying building wake effects on the dispersion of the releases is based on release/receptor orientation relative to the plant structures;

• The releases from the Unit 1 and Unit 2 containment building surfaces are treated as diffuse sources; 2.3-31 2.3-31Revision 21 September 2013

DCPP UNITS 1 &2 FSAR UPDATE 0All releases are treated as ground level as there are no release conditions that merit categorization as an elevated release (i.e., 2.5 times containment building height) at this site (conservative); and The x/Q value from the accident release point to the center of the control room boundary at roof level is utilized for control room in-leakage since the above %/Q can be considered an average value for in-leakage locations around the control room envelope. The y/Q from the accident release point to the center of the control room boundary at roof level is also utilized for control room.

ingress/egress. The outer doors to the control room are located at approximately the middle of a) the east side (i.e., auxiliary building side) wall of the control room and 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 is utilized for TSC in-leakage since the above 7J can be considered an average value for in-leakage locations around the TSC building envelope.

Summarized below are some of the other salient aspects of the control room and technical support center %/ analyses, as applicable.

Control Room Receptors within 10-meters of Release Regulatory Guide 1.194, June 2003, Regulatory Position C.3.4 recommends that ARCON96 methodology not be used for analysis at distances less than about 10 meters. However, as an exception to Regulatory Guide 1.194, June 2003, Regulatory Position C.3.4 the ARCON96 methodology has been applied for two cases 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 Unit 1 control room normal intake and 7.8 meters for Unit 2 containment building to Unit 2 control room normal intake) is considered acceptable since the dominating factors 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 Release Since the Unit 1 and Unit 2 MSSVs, 10% ADVs, and MSLB release points are located within 1.5 meters line-of-sight distance from the affected unit's control room normal intake, this near-field distance is considered outside of the ARCON96 application domain. Although ARCO N96 is capable of estimating near-field dispersion, the 1.5-meter line-of-sight distance from the releases to the receptors is much less than the 10-meter distance recommended as the minimum applicable distance in Regulatory Position 0.3.4 of Regulatory Guide 1.194, June 2003. Thus no z/Qs are developed for the above release point /

receptor combinations.

Enernqetic Releases 2.3-32 2.3-32Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE The largervertical velocity than the of the MSSV 95th percentile windand 10%ofADVs speed releases 1 rn/sec is at least 95 times and approximately 5 times larger thanthe highest observed 10-meter wind speed (i.e., 18.9 m/sec) within the 5-year meteorological data base. The large vertical velocities of the MSSV and 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 buildings such 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 not contaminated given that the horizontal distance is only 1.5 meters. Moreover, this short distance precludes the releases from reaching the control room normal intakes 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 normal intakes (i.e., 22 meters). Plume rise calculations indicate that the MSSV and ADV release heights will be enhanced by 11 meters at the 95th percentile wind speed of 1 rn/sec due to the large vertical velocities of the releases. Thus, for purposes of estimating dose consequences, it is appropriate to use the x/Q associated with the normal control room intake of the opposite unit for releases from the MSSVs / 10% ADVs as the worst case control room intake location.

Vertically-Oriented Enerqetic Releases Regulatory Position C.6 of Regulatory Guide 1.194, June 2003 establishes the use of a deterministic reduction factor of 5 applied to ARCON96 7./Q values for energetic releases from steam relief valves or atmospheric dump valves. These valves must be uncapped and vertically-oriented and the time-dependent vertical velocity must exceed the 95th-percentile wind speed at the release point height by 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 the first 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> of the accident, the reduction factor of 5 is clearly applicable to the DCPP MSSV and 10% ADVs releases. Note that since %!Q values are averaged 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 <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />, application of the factor of 5 reduction is not appropriate for %/Q values applicable to averaging periods .beyond the 2-8 hours averaging period. For assessment of an environmental release between 8 to 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />, continued use of the 2-8 hour

%IQ,with the factor of 5 reduction, is acceptable and conservative.

Dual Intakes The Unit 1 and Unit 2 control room pressurization air intakes which also serve the technical support center, may be considered dual intakes for the purpose of providing a low contamination intake regardless of wind direction for any of the release points since the two control room pressurization air intakes are never within the same wind direction window; defined as a wedge centered on the line of sight between the release and the receptor with the vertex located at the release point. The size of the wedge for each release-receptor combination is 90 2.3-33 2.3-33Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE degrees azimuth with the use of ARCON96, as described in Regulatory Position 0.3.3.2 of Regulatory Guide 1.194, June 2003.

Redundant Radiation Monitors Per Regulatory Guide 1.194, June 2003, Regulatory Position 0.3.3.2.3, based on the dual intake design of the control room pressurization intakes, and the availability of redundant PG&E Design Class I radiation monitors at each pressurization intake (which provide the capability of initial selection of the cleaner intake and support the expectation that the operator will manually make the proper intake selection throughout the event), allows the x/Q values applicable to the more favorable control room pressurization intake c-anto be reduced 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 Vent The 16 inch PG&E Design Class 11gland seal steam exhauster line connects to the PG&E Design Class I plant vent. In addition, the plant vent expansion joint may experience a tear during a seismic event, however the plant vent will remain intact and functional.

a) The gland seal steam 16 inch exhauster line connects to the plant vent at El 144'-6" (Centerline) on the North-East side / South-East side of the Unit 1 and Unit 2 containments, respectively. It has been determined that should a failure occur due to a seismic event, it would occur at the interface of this line and the plant vent.

b) The plant vent expansion joint is located at El 155.83' North-East side I South-East side of the Unit 1 and Unit 2 containments, respectively. As discussed earlier, the plant vent expansion joint may experience a tear during a seismic event.

An assessment of the potential release locations identified above indicates that the %/Qvalues developed for the plant vent are either conservative or representative of these potential release points.

Release points and receptor locations are provided in Figure 2.3-5, while Table 2.3-1 46 provides the release point I receptor combinations that were evaluated. Tables 2.3-147 and 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 to support bounding analyses for an accident that occurs at either unit. They take into consideration the various release points-receptors applicable to each accident in order to identify the bounding x/Q values and reflect the allowable adjustments and reductions in the values as discussed earlier and further summarized in the notes of Tables 2.3-1 47 and 2.3-1 48.

2.3-34 2.3-34Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE the 7J values 49 presents for for the individualtopost-LOCA release point TSC Table 2.3-1combinations Unit 1 and Unit 2 applicable the TSC normal intake and the receptor value for potential TSC center of the TSC boundary at roof level (considered an average and Unit 2 control unfiltered in-leakage locations around the envelope). The Unit 1 mode. Thus, room pressurization air intakes also serve the TSC during the emergency room pressurization the 7J~ presented in Tables 2.3-1 47 and 2.3-148 for the control intake design and ability to select the more intakes inclusive of the credit for dual favorable intake are also applicable to the TSC.

2.3.6 LONG-TERM (ROUTINE) DIFFUSION ESTIMATES TO BE REVISED.

HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED 2.3.6.1 Objective kilometers Annual relative concentrations (z/Q) were estimated for distances out to 80 April 1975. These from on site meteorological data for the period May 1973 through the relative concentrationsare presented in Table 2.3-2; they were estimated using models described in Reference 18. The same program also produces cumulative means having frequency distributionsfor selected averagingperiods using overlapping of hourly updates. Forcritical offsite locations, measured lateralstandarddeviations wind direction, GrA, and bulk Richardson number, Ri, were used as the stability at the parameters in the computations. The meteorological input data were measured averaged relative 10 meter level of the meteorological tower at DCPPsite. Annual concentrations calculated by the above methods are presented in Table 2.3-4.

2.3.6.2 Calculations for the The meteorologicalinstrumentation that was used to obtain the input data site is described in previously discussed relative concentration calculationsat DCPP are Section 2.3.4. Proceduresfor obtaining annual averagedrelative concentrations describedin detail in Reference 15.

2.3.6.3 Meteorological Parameters input The following assumptions were used in developing the meteorological parametersrequired 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 law function where the exponent, F, varies with stability class and is assigned the following values:

Pasquill Stability Class Exponent (P) 2.3-35 2.3-35Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE 2.3.8.4 Safety Guide 23, February 1972 - Onsite Meteorological Programs As discussed in Section 2.3.4, the pfeepefatipna-meteorological data collection program was designed and has been updated continually to meet the requirements of Safety 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 Conditions During and Following an Accident Wind speed, wind direction, and estimation of atmospheric stability indication in the of control room provide information for use in determining the magnitude of the release radioactive materials and in continuously assessing such releases during and following an accident (refer to Table 7.5-6 for a summary of compliance to Regulatory Guide 1 .97, Revision 3).

2.3.8.6 Regulatory Guide 1.111, March 1976- Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Re leases from Light-Water-Cooled Reactors The pre-operational values of dilution factor and deposition factor used in the calculation of of annual average offsite radiation dose are discussed in Section 11 .3.7. The values deposition rate were derived from Figure 7 of Regulatory Guide 1.111, March 1976, for a ground-level release.

2.3.8.7 Regulatory Guide 1.111, Revision 1, July 1977 - Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors The annual average relative concentration values are developed for each sector, at the outer LPZ boundary distance for that sector, using the method described in Regulatory Position C.1 .c of Regulatory Guide 1.111, Revision 1. These values are used to 2-8 calculate the intermediate averaging time 7/0 values at the LPZ for the periods of hours, 8-24 hours, 1-4 days, and 4-30 days following the postulated accident. This information is used as input to develop the accident x/O values at the LPZ using Regulatory 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 - Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants The short-term atmospheric dispersion factors applicable to the exclusion area boundary and the low population zone for post-accident accident releases from Unit 1 and Unit 2 are calculated using methodology applicable to "ground level" releases provided in Regulatory Guide 1.145, Revision 1. Refer to Section 2.3.5.2.

2.3-38 2.3-38Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE 2.3.8.9 Regulatory Guide 1.194, June 2003 - Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants The control room and technical support center atmospheric dispersion factors for radiological releases from Unit 1 and Unit 2 are calculated using methodology outlined in Regulatory Positions 0.1 through C.3, and the adjustment factor for vertically orientated energetic releases from steam relief valves and atmospheric dump valves allowed by Regulatory Position C.6, and NRC ARCON96 methodology as documented in 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 TMI Action Plan Requirements Item llII.A.2 - Improving Licensee Emergency Preparedness-Long-Term:

As discussed in Section 2.3.4, the primary and backup meteorological data are available in the control room and emergency response facilities via the TRS servers and EARS, in accordance with NUREG-0654, Revision 1, November 1980.

As discussed in Section 2.3.4, the measurement subsystems consist of a primary meteorological tower and a backup meteorological tower. The primary meteorological computer and the backup meteorological computer communicate with each other, the EARS and also with the TRS server. Primary and backup meteorological data are available on the PPCs via the TRS servers and thus in the control room and emergency response 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 Guide 1 .97, Revision 3. Wind direction, wind speed, and estimation of atmospheric stability are categorized as Type E variables, based on Regulatory Guide 1 .97, Revision 3. The PPC is used as the indicating device to display meteorological instrument signals. In addition, Type E, Category 3, recorders are located in the meteorological towers.

2.3.8.8---1 IE Information Notice 84-91, December 1984- Quality Control Problems of Meteorological Measurements Programs In addition to the primary meteorological towers, a supplemental meteorological measurement system is provided in the vicinity of the plant site in order to meet IE Information Notice 84-91. As discussed in Section 2.3.4.5, this supplemental measurement system consists of three Doppler SODAR and seven tower sites located as indicated in Figure 2.3-4. The primary and secondary meteorological towers in conjunction with the supplemental system adequately predict the meteorological conditions at the site boundary (800 meters) and beyond.

2.3-39 2.3-39Revision 21 September 2013

DCPP UNITS 1 & 2 FSAR UPDATE

24. ANSI/ANS 2.5, American National Standard for Determininq Meteoroloqical Information at Nuclear Power Sites, American Nuclear Society, 1984.

25. National Oceanic and Atmospheric Administration, An Evaluation of Wind Measurements 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-F of the FSAR Update, Revision 0 through Revision 10 (Currently maintained at PG&E Nuclear Power Generation Licensing office files).

28. Regulatory Guide 1.111, Revision 1, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors, USNRC.

29. Ramsdell, J. V. Jr. and C. A. Simonen, Atmospheric Relative Concentrations in Building Wakes. Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, PNL-10521, NUREG/CR-6331, Revision 1, May 1997.

30. Regulatory Guide 1.194, June 2003, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, USNRC.

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DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.1-1 Sheet 8 off112 Set ITITLE APPLICABILITY CRITERIA Auxiliary Containment Containment Containment Combustible Emergency Control Technical EnierdSafety Fetue Containment Support Feedwater IFunctional Heat Removal Air Purification Isolation Gas Control in Core Cooling Room Section [6.2.1 Design Systems 6.2.2 and Cleanup 6.2.3 System 6.2.4 Containment 6.2.5 System 6.3

_ I Habitability 6,4.1 Center System 6.4.2 System 6.5

5. Recqulatory Guides (contd.)

Regulatory Guide Performance-Based 1.163, Containment Leak- X X September 1995 Test Program Alternative Radiological Source Regulatory Guide Terms for Evaluating 1.163, July 2000 Design Basis Accidents at Nuclear

______________Power Reactors Demonstrating Regulatory Guide Control Room 1.197, Revision 0, Envelope Integrity atX May 2003 Nuclear Power Reactors

6. NRC NUREG Clarification of TMI X X X X X X X NUREG-0737, Action Plan November 1980 Rqieet

7. NRC Generic Letters Generic Letter 89-10, Safety-Relatedt Motor-Operated I I x June 1989 Valve Testing andI Surveillance Revision 22

DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-32 LOSS-OF-COOLANT ACCIDENT TOTAL ELEMENTAL IODINE & PARTICULATE REMOVAL COEFFICIENTS Elemental Iodine Removal Particulate Removal Coefficient Coefficient From To Time (hr Note 1 (hr1)

Time (sec) Sprayed Unsprayed Sprayed Unsprayed (sec) Region Region Region Region 0 30 N/A N/A 30 111 272.45.89 0.0062 111 1,800 2.24 0.0071 1,800 3,798 20.57 (Note 2) 9.35 0.1144 3,798 4,518 0.00 (Note 3) __1.02_0.122 4,518 5,030 7.50 0.1239 5,030 6,4806.0.13 6,480 7,200 19.91 (Note 2) 0.00 4.74____ 0.1236__

7,200 8,004 3.39____ 0.1222__

8,004 22,1521.3000 22,152 22,518 00 22,518 720 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 for elemental iodine by sprays is eliminated after a DF=200 is reached in the containment atmosphere.

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 sprayed region after termination of recirculation spray

DCPP UNIT 1 & 2FSAR UPDATE TABLE 6.2-33 Sheet 1 of 3 LOSS-OF-COOLANT ACCIDENT CONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATA Post-LOCA Time Containment Containment Containment -RH Pressure Temperature 0

Seconds psia F%

0.00 16.00 120 18 0.52 18.81 145.9 72.3 1.04 21.45 163.8 88.2 1.54 23.76 176.78 93.5 2.04 -25.88 186.8 96 2.54 27.80 194.83 97.4 3.04 29.44 201.03 98.1 3.54 30.86 205.97 98.5 4.04 32.10 210.01 98.9 4.54 33.21 213.43 99.1 5.04 34.24 216.47 99.2 7.04 38.04 226.75 99.5 7.54 38.96 229.03 99.5 8.54 40.68 233.06. 99.6 10.04 43.06 238.34 99.7 10.54 43.81 239.9 99.7 11.54 45.21 242.76 99.8 13.04 47.12 246.47 99.8 14.54 48.62 249.65 99.8 16.04 50.36 252.4 99.9 17.54 51.70 254.7 99.9 19.04 52.87 256.67 99.9 20.54 53.88 258.34 99.9 21.54 54.24 258.9 99.9 22.04 54.36 259.08 100 23.54 54.48 259.25 100 25.04 54.40 259.12 100 29.54 53.87 258.24 100 32.54 53.62 257.85 100 48.54 53.14 257.02 100 54.54 53.37 257.03 100 68.04 53.70 256.79 100 86.54 53.18 255.93 100 144.18 50.88 251.95 100 158.18 50.44 251.15 100

DCPP UNITi1 & 2FSAR UPDATE TABLE 6.2-33 Sheet 2 of 3 LOSS-OF-COOLANT ACCIDENT CONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATA Post-LOCA Time Containment Containment Containment -RH Pressure Temperature 0

Seconds psia F%

188.18 49.70 249.82 100 200.18 49.48 249.41 100 212.18 49.33 249.15 100 266.18 49.21 248.92 100 333.18 49.37 249.2 100 400.18 49.70 249.82 99.9 534.18 50.56 251.76 99.1 668.18 51.49 254.34 97.4 803.18 52.60 256.32 97.3 816.18 52.43 255.22- 98.7 857.18 52.08 254.17 99.5 912.18 51.73 253.6 99.4 1021.19 51.21 252.76 99.3 1131.19 50.83 252.11 99.2 1240.19 50.55 251.61 99.1 1458.19 50.16 .250.91 99 1677.19 . 49.94 250.49 98.9 1730.19 50.43 251.61 98.5 1746.19 50.26 250.56 99.9 1859.19 49.41 248.91 100 1988.19 48.57 247.32 100 2247.19 47.07 244.4 100 2505.19 45.75 241.75 100 2764.19 44.56 239.29 100 3022.19 43.47 236.96 99.9 3281.19 42.45 234.71 99.8 3604.24 41.26 231.94 99.9 3798.24 40.10 229.1 100 3888.29 40.52 230.99 98.2 3978.29 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 UPDATE TABLE 6.2-33 Sheet 3 of 3 LOSS-OF-COOLANT ACCIDENT CONTAINMENT PRESSURE, TEMPERATURE & RELATIVE HUMIDITY DATA Post-LOCA Time Containment Containment Containment -RH Pressure Temperature 0

Seconds psia F%

4573.29 41.87 234.69 97.2 4592.29 41.82 233.97 98.4 4666.29 41.74 233.23 99.5 5110.73 41.37 232.3 99.6 5700.73 40.79 230.94 99.6 6890.73 39.55 227.94 99.6 8080.73 38.36 224.96 99.5 10000.80 36.64 220.36 99.4 11001.50 35.82 218.08 99.4 12001.50 35.08 215.93 99.4

• 13001.50 34.39 213.85 99.4 14001.50 33.74 211.85 99.4 15001.50 33.13 209.91 99.4 16001.50 32.57 208.07 99.3 18001.50 31.59 204.69 99.3 20001.50 30.71 201.53 99.3 21001.50 30.31 200.06 99.3 22518.00 29.78 198.01 99.4

DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-34 Sheet 1 of 4 LOSS-OF-COOLANT ACCIDENT CONTAI NMENT STEAM CONDENSTION DATA Post- Steam Condensation Rate LOCA Thermal Containment Injection Recirculation Total Steam Time Conductor Fan Coolers Spray Spray Condensation Rat __

Seconds Ibm/sec Ibm/sec Ibm/sec Ibm/sec Ibm/sec g/se 0.00 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. i7 1.04 49.71 0.00 0.00 0.00 49.71 22548 08 2.54 204.71 0.00 0.00 0.00 204.71 92854,90 3.04 250.47 0.00 0.00 0.00 250.47 .113611.29 3.54 290.16 0.00 0.00 0.00 290.16 131614.37 4.04 325.28 0.00 0.00 0.00 325.28 147544.54 5.04 384.15 0.00 0.00 0.00 384.15 174247.52 7.54 509.97 0.00 0.00 0.00 509.97 23131 .52 10.04 611.01 0.00 0.00 0.00 611.01 27714£.49 12.54 684.86 0.00 0.00 0.00 684.86 310647.29 15.04 734.83 0.00 0.00 0.00 734.83 33331 .30 17.54 766.16 0.00 0.00 0.00 766.16 347524.35 20.54 783.18 0.00 0.00 0.00 783.18 355244.50 24.54 742.58 0.00 0.00 0.00 742.58 33682 .64 28.54 684.40 0.00 0.00 0.00 684.40 31043 .64 32.54. 644.28 0.00 0.00 0.00 644.28 29224 .51 37.04 623.09 0.00 0.00 0.00 623.09 28262 .89 53.54 538.96 0.00 0.00 0.00 538.96 24446 .16 70.04 469.30 0.00 0.00 0.00 469.30 21287 .91 87.04 412.48 0.00 0.00 0.00 412.48 187097.79 87.57 410.75 44.28 0.00 0.00 . 455.03 20639 .15 88.07 409.15 45.29 0.00 0.00 454.44 20613 .53 107.14 354.70 44.87 13.61 0.00 413.18 18741 .31 124.18 312.19 44.28 72.76 0.00 429.23 19469 .47 146.18 267.97 43.51 70.37 0.00 381.85 17320 .26 169.18 231.50 42.78 68.46 0.00 342.74 155464.26 197.18 197.26 41.97 66.66 0.00 305.89 13874! .38 234.18 166.95. 41.32 49.25 0.00 257.52 11680£.11 262.18 149.95 41.00 48.48 0.00 239.43 10860* .63 327.18 121.98 40.50 47.28 0.00 209.76 95145 54 403.18 100.43 40.23 46.58 0.00 187.24 849306B4 449.18 91.02 40.17 46.66 0.00 177.85 80671 41 502.18 82.32 40.11 47.29 0.00 169.72 76983 70 Revision 19 May 2010

DCPP UNITI1 & 2FSAR UPDATE TABLE 6.2-34 Sheet 2 of 4 LOSS-OF-COOLANT ACCIDENT CONTAINMENT STEAM CONDENSTION DATA Post- Steam Condensation Rate _________

LOCA Thermal Containment Injection Recirculation Total Steam Time 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 43 560.18 74.55 40.08 48.66 0.00 163.29 74067 10 629.18 67.45 40.12 47.87 0.00 155.44 70506 40 683.18 63.43 40.22 47.23 0.00 150.88 68438 02 754.18 59.41 40.45 47.01 0.00 146.87 66619 12 802.18 57.28 40.64 46.99 0.00 144.91 65730 07 832.18 49.92 40.37 57.62 0.00 147.91 67090 85 876.18 43.40 40.00 58.41 0.00 141.81 64323 94 937.18 37.29 39.57 "57.66 0.00 134.52 61017 25 1013.19 32.26 39.16 56.94 0.00 128.36 58223 12 1094.19 28.58 38.82 56.39 0.00 123.79 56150 20 1148.19 26.76 .38.64 56.09 0.00 121.49 55106 94 1243.19 24.16 38.39 55.68 0.00 118.23 53628 23 1341.19 22.03 38.19 55.34 0.00 115.56 52417 14 1423.19 20.55 38.07 55.11 0.00 113.73 51587 06 1492.19 19.48 37.98 54.94 0.00 112.40 50983 79 1564.19 18.50 37.90 54.79 0.00 111.19 50434 94 1607.19 17.97 37.86 54.72 0.00 110.55 50144634 1644.19 17.54 37.83 54.66 0.00 110.03 49908 77 1672.19 17.23 37.81 54.61 0.00 109.65 49736 41 1678.19 17.19 37.71 54.73 0.00 109.63 49727 33 1730.19 19.93 36.37 52.72 0.00 109.02 49450654 1794.19 16.03 35.05 60.34 0.00 111.42 50539 27 1859.19 14.01 33.89 60.13 0.00 108.03 49001 59 1985.19 11.88 32.19 59.74 0.00 103.81 47087 43 2052.19 11.05 31.50 59.53 0.00 102.08 46302 71 2116.19 10.33 30.96 59.34 0.00 100.63 45645 00 2244.19 9.09 30.09 58.96 0.00 98.14 44515 56 2311.19 8.50 29.72 58.76 0.00 96.98 43989 39 2439.19 7.47 29.12 58.40 0.00 94.99 43086 74 2567.19 6.53 28.62 57.93 0.00 93.08 42220 38 2695.19 5.66 28.19 57.37 0.00 91.22 41376 70 2763.19 5.23 27.94 57.08 0.00 90.25 40936 71 2890.19 4.47 27.51 56.53 0.00 88.51 40147146 3018.19 3.76 27.11 56.00 0.00 86.87 39403 57 Revision 19 May 2010

DCPP UNIT 1 & 2 ESAR UPDATE TABLE 6.2-34 Sheet 3 of 4 LOSS-OF-COOLANT ACCIDENT CONTAINMENT STEAM CON DENSTION DATA Post- ______Steam Condensation Rate _________

LOCA Thermal Containment Injection Recirculation Total Steam Time 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 77 3210.19 2.80 26.56 55.23 0.00 84.59 38369 38 3338.19 2.21 26.22 54.74 0.00 83.17 37725 28 3466.19 1.64 25.89 54.25 0.00 81.78 37094179 3594.19 1.12 25.58 53.78 0.00 80.48 36505 12 3722.24 0.24 25.02 54.13 0.00 79.39 36010,701 3796.24 0.02 24.72 53.61 0.00 78.35 35538 *96 3843.29 2.24 24.91 0.00 0.00 27.15 12315 03 3901.29 4.39 25.13 0.00 0.00 29.52 13390 05 3995.29 6.72 25.42 0.00 0.00 32.14 14578 46 4105.29 8.33 25.70 0.00 0.00 34.03 15435175 4189.29 9.21 25.87 0.00 0.00 35.08 15912 032 4291.29 10.03 26.06 0.00 0.00 36.09 16370 15 4383.29 10.63 26.21 0.00 0.00 36.84 16710 34 4463.29 11.06 26.33 0.00 0.00 37.39 16959182 4515.29 11.25 26.41 0.00 0.00 37.66 1708229 4518.29 11.26 26.41 0.00 0.92 38.59 17504 13 4584.29 10.17 26.48 0.00 8.26 44.91 20370 83 4592.29 10.15 26.49 0.00 9.02 45.66 20711033 4654.29 10.18 26.55 0.00 11.12 47.85 2170440 4698.29 10.19 26.59 0.00 11.33 48.11 21822 33 4734.29 10.21 26.60 0.00 11.39 48.20 21863 15 4785.29 10.19 26.62 0.00 11.43 48.24 21881 30 4807.29 10.17 26.63 0.00 11.44 48.24 21881 30 4843.29 10.14 26.63 0.00 11.46 48.23 2187676 4851.29 10.13 26.64 0.00 11.46 48.23 2187676 4895.29 10.09 26.64 0.00 11.48 48.21 2186769 4926.29 10.05 26.64 0.00 11.49 48.18 21854108 4932.29 10.06 26.63 0.00 11.49 48.18 21854i08 4988.29 9.99 26.63 0.00 11.51 48.13 21831 40 6120.73 8.70 25.98 0.00 11.43 46.11 20915 15 7400.73 7.52 25.02 0.00 11.13 43.67 1980838 8680.73 6.51 24.04 0.00 10.86 41.41 1878326 9510.73 6.06 23.47 0.00 10.70 40.23 18248 02 14001.50 4.42 20.91 0.00 9.98 35.31 16016 35 Revision 19 May 2010

DCPP UNIT 1 & 2 FSAR UPDATE TABLE 6.2-34 Sheet 4 of 4 LOSS-OF-COOLANT ACCIDENT CONTAINMENT STEAM CONDENSTION DATA Post- Condensation Rate _________

T Total Steam

____________Steam LOCA Thermal Containment Injection Recirculation Time Conductor Fan Coolers S ray Spray Condensation Rat Seconds 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 28 22518.00 3.23 17.41 0.00 9.27 1 29.91 13566 95 Revision 19 May 2010

DCPP UNIT 1 &2 FSAR UPDATE TABLE 6.2-35 PARAMETERS FOR FISSION PRODUCT REMOVAL ANALYSIS Parameter Value Total Containment Volume, ft3 2.55 x 106 .

Containment Spray Coverage 0.825 Fraction Average Spray Fall Height, ft 116 (Note 1)

Spray Flow Rate, gpm 2,456 for 111 sec-< t - 3,798 sec 0 for 3,798 sec <t-<4,518 sec

______________________1,211 for 4,518sec<t-<22,518sec Spray Droplet Radius, cm 500 x i 04 Note 1: The average fall height is conservatively approximated as the distance from the lowest spray header to the operating deck as follows:

Elevation of Header #1 =256'-0" Elevation of Deck = 140'-0' Fall Height to the Deck = 116'-0" Note that this fall height is more conservative than the area-weighted average drop fall height of 128 ft shown in Table 6.2-37.

DCPP UNIT 1 & 2 FSAR UPDATE

-ABLE462-36 PARDAMEII:TERSI ANI*D RESU l:q TS FORh SpRAY IOD'MINE: DREMOVt'\AL ANkALYISVl DURIIING"* INJCION-fTf'k PHIASEI- OPERAION*-I*'

Mitnmtur Estimate Poara m tert Containment free ;'ol, fi 2. x 10'. -- p4 Unaprayedve'lume, % -4 !0

-t--

-- 4

--- 4 -7 Spray pump flo,-'iate, gppm C...nta...inet p~ressure, psig

-- _-2(b R-esults Ex*ponential removal -- 9 - -46 _*

eonstxi~t for the spray system (lrl )

ssntai. *nms nt i~ntsri anaysi rs pr.... s.. in.. pp. ndi.:..

potential. offsic, . !nra ii- dsc IA,thi s1;hangs is exreel . s...

mall a,.d san' be

... ns.dercd, insignifisant (Refr~sneo~0O

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-13 CORE FIS .... PRODUCT ENERG ATER OPEDR-ATIONM WAITH- EXVTENDI'EDr FEl CYCLVPrES Trime After Reactor Trip, Energy Re!ease Rate, Intcgrated Energy Release, watts!MWt 10~ watts days!MWt x 1p4 0496 8A4 8 2_28

-10 20 2*00 40 642, 143 404 4402 44*4 8O 0*488 424 400 (a) As..u.e. 50o/ coere halogens +99%oo/cther fission pr,,,-*odut and4 n*o nobl-*) e Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE F4SSION PRODUCT DECAY DEPOSTION IN SUMP SOLUTION f-aA bump ,l-ISn

ss o autEnr '---

1/2 w nt-tgas/Mt x 0 P¶eester-Tnp-watts/M\At x 10 1/2 2-&6 546 40) 5 20 2g06

&44$

40 t4*6 (a} -onsioers reiease 0ot u percent or core natogens,-noenoble-gasesan 1 ecnaroto ~inpodcst n sm ouin Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE 15.5 RADIOLOGICAL CONSEQUENCES OF PLANT ACCIDENTS The purposes of this section are: (a) to identify accidental events that could cause radiological consequences, (b) to provide an assessment of the consequences of these accidents, and (c) to demonstrate that the potential consequences of these occurrences are 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 single event. In the analyses reported in this section, the basic events involved in various possible plant accidents are identified and studied with regard to the performance of the engineered safety features (ESF). The full spectrum of plant conditions has been divided into four categories in accordance with their anticipated frequency of occurrence and risk to the public. The four categories as defined above are as follows:

Condition I: Normal Operation and Operational Transients Condition 11: Faults of Moderate Frequency Condition Ill: Infrequent Faults Condition IV: Limiting Faults The basic principle applied in relating design requirements to each of these conditions is that 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 shall be 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 guide to the Preparation of Environmental Reports (Reference 3). While some differences exist 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 the Condition I category, and some of those in the Condition 11category, fall in the range of normal operation. For this reason, the radioactive releases and radiological exposures associated with these conditions are analyzed in Chapter 11 and are not discussed separately in this chapter. The analyses of the variations in system parameters associated with Condition I occurrences or operating modes are discussed in Chapter 7 since these states are not accident conditions. In addition, some of the events identified as potential accidents in Regulatory Guide 1.70, Revision 1 (Reference 2), have no significant radiological consequences, or result in minor releases within the range of normal releases, and are thus not analyzed separately in this chapter.

15.5-1 15.5-IRevision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.1 DESIGN BASES The following regulatory requirements, including Code of,"Federal .. *;,on (Cp*.,

• • ,., Regulati.D FR) 10 CFR Par.t 100, Genr- De...igr,,n Critria; (DC), Safety Guide,* and. Regul...ator,,

in this Gudsare applicable to the DCPP radiological consequence analyses presented GIa*I. They form the bases of the acceptance criteria and methodologies as described 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 the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors" (5) Safet Gu.dc 7, March 1971, ,Conro.l of C~,ombusibl Gas Co..... rtios in Gentaknmen4*

(6~)(5) Safety Guide 24, March 1972, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure" Radio;logical,-,v*v, ,v*of, a Fuel, Handl,,,ing,A^ccident in,,t,;he,,

Consequences Fuel.* Handling and Storg F..a*cility,, for Boilin and. Pressuri..ed Water ** Rea,-tors"

{8)(6) Regulatory Guide 1.183, July 2000, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (9) Reguato,,,,-,,'_,Guide !.!9,,5,Ma 20,--,"'°n "Methods,4, and* Assumption for,.,E,.a-luating,-

15.5.1.1 List of Analyzed Accidents for The following table summarizes the accident events that have been evaluated UFSAR Section radiological consequences. The table identifies the applicable and describing the analysis and results for each event, the offsite/onsite locations codes applicable dose limits, and the radiological analysis and isotopic core inventory used.

15.5-2 15.5-2Revision 19 May 2010

DCPP UNITS 1 & 2 ESAR UPDATE FSRRadiological Isotopic Core Section Boundary Dose Limit Analysis Inventory AcietEet Code(s)

Code(s) 15.5.10 EAB and LPZ RADTRAD SAS2 /

Loss of Electrical Control *O¢m3.03EMERALD ORIGEN-Load (LOL)

RoomE-A8- 2*ie2.5 rem SEMEP.ALD

~TEDE

-T* 5 rem TEDE CONDITION Ill 15.5.11 EAB and LPZ 2.5 remn TEDE N/A N/A Small Break Refer to LOCA (SBLOCA) Control  ;=04mRefer to RoomE-AB- 25-84m Section Section a~4~5 rem TEDE 15.5.23E-ME-RA 15.523E-ME-RA Trhyre4 L-bL 15.5-3 15.5-3Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE FSRRadiological Isotopic Core AcietEet Section Boundary Dose Limit Analysis Inventory Code(s) Code(s) 15.5.12 EAB and LPZ 2.5 rem TEDE N/A N/A Minor Secondary System Pipe *A-i-P Refer to Refer to Breaks :T-Ayred OQ-re* Section Section WhGle-Beety 25re 15.5.18N/A 15.5.18NIA Refe-re R4ef-e~4 Scin55.2 Section 15.5.12 Inadvertent 15.5.13 EAB and LPZ 2.5 rem TEDE N/A N/A Loading of a Fuel E-AS-ani-L-P- Refer to Refer to T*y~4 Section 15.5.13 Section 15.5.13 Assembly Complete Loss of 15.5.14 LAB and LPZ 2.5 rem TEDE N/A N/A

• A-.4P Refer to Refer to Forced Reactor Coolant Flow Thri Section Section W^h,,e,,,,.-p,,ody ,?,0e-re 15.5.4410 15.5.1410 25 rem 15.5.15 LAB and LPZ 2.5 rem TEDE N/A N/A Under-Frequency EA-n-PZRefer to Refer to T*hyroid Section Section WholeBedy 3004en4 15.5.10E-ME-RA 15.5.10E-ME-RA 2-84emL-L-15.5.16 LAB and LPZ 2.5 rem TEDE N/A N/A Single Rod Cluster Control *A-i4lP Refer to Refer to Assembly Thri 3G4ei Section Section Withdrawal Whl-Bd 2,5--rein 15.5.23E-MER#A 15.5.23E-ME-RA CONDITION IV Large 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-RAI4 LZ300 rem L-QGADOSE OP4GEN-2 Whole-Bod' Control Room 30 rem 15.5-4 15.5-4Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE FSRRadiological Isotopic Core AcietEet Section Boundary Dose Limit Analysis Inventory Code(s) Code(s)

Main Steam Line 15.5.18 EAB and LPZ RADTRAD SAS2 /

Break (MSLB) 3.03tGA*O ORIGEN-Pre-Accident E-SORlGEN-2 Iodine Spike Thyid0O-25 rem Whele-Bedy TEDE 25 ,remn Accident-initiated Iodine Spike Woe-ey 2.5 rem TEDE Control Room Whee-*-~ em 5 rem TEDE Main Feedwater 15.5.19 EAB and LPZ N/A N/A Line Break Refer to Refer to (FWVLB) Pre-Accident 25 rem TEDE Section Section Iodine Spike 15.5.189 15.5.198 Accident- 2,5 rem TEDE initiated Iodine Spike E-AB nd LPZ 30rm 15.5-5 15.5-5Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE FSRRadiological Isotopic Core AcietEet Section Boundary Dose Limit Analysis Inventory Code(s) Code(s)

Steam Generator 15.5.20 EAB and LPZ RADTRAD SAS2 /

3.03RADTR-A, ORIGEN-Tube Rupture Pre-Accident 25 rem TEDE SEME-RAbQ--

(SGTR)

Iodine Spike NO'RMA4=

Accident- 2.5 rem TEDE initiated Iodine Spike Control Room 5 remn TEDE Pr-hre-A*~~

z-300remn Th~4 25-rem Accident (LIA) ACcidntrl 5rmT E 3.E-RAD OIGN initiAted ;30 -remSEM-L-

_ _-LP- 2 5 em 30-remn Control Room e 15.5.21. EAB and LPZ 2.5 rem TEDE RADTRAD SAS2 /

Loced Rotorng Control 5 rem TEDE 3.03L-ORADLB ORIGEN-Accident (LRA)

F-e,4a*#*RoomE-Ag- .03O S'-EDEE- SOMRAGEN Araad-L-P-Z ~

Gontb!Room~~e 155-hRvsin 9Mai21

DCPP UNITS 1 &2 FSAR UPDATE FSRRadiological Isotopic Core AcietEet Section Boundary Dose Limit Analysis Inventory

__________Code(s)

Code(s) 462* ,AB-aJd-bJP-_ L nGAnnS OR!GEN-2 Fu*el Hand!.n

.Inside- Trhyr4 75 .rem Control Room Control Rod 15.5.23 EAB and LPZ 6.3 rem TEDE RADTRAD SAS2 I Ejection Accident Control 5 rem TEDE 3.03EMERALD ORIGEN-(CREA) RoomF=AB-  ;=380em S=ME-RAL-D Whoe~eody Control Room S-rom Waste Gas Decay 15.5.24 EAB and LPZ EMERALD EMERALD Tank Rupture Thyroid 300 rem Whole Body 25 rem Liquid Holdup 15.5.25 EAB and LPZ LOCADOSE EMERALD Tank Rupture Thyroid 300 rem Whole Body 25 rem Volume Control 15.5.26 EAB and LPZ EMERALD EMERALD Tank Rupture Thyroid 300 rem Whole Body 25 rem 15.5.1.2 Assumptions associated with Loss of Offsite Power The assumptions regarding the occurrence and timing of a Loss of Offsite Power (LOOP) during an accident are selected with the intent of maximizing the dose consequences. A LOOP is assumed for events that have the potential to cause grid perturbation.

event

i. The dose consequences of the LOCA, MSLB, SGTR, LRA, CREA and LOL are evaluated with the assumption of a LOOP concurrent with reactor trip.

ii. The assumption of a LOOP related to a postulated design basis accident which leads to a reactor trip does not directly correlate to an FHA. Specifically, a FHA does not directly cause a reactor trip and a subsequent LOOP due to grid instability; nor can a LOOP be the initiator of a FHA. Thus the FHA dose consequence analyses are evaluated without the assumption of a LOOP.

unit is In addition, in accordance with current DCPP licensing basis, the non-accident assumed unaffected by the LOOP.

15.5-7 15.5-7Revision 19 May 2010

DCPP UNITS 1 & 2 ESAR UPDATE 15.5.2 APPROACH TO ANALYSES OF RADIOLOGICAL EFFECTS OF ACCIDENTS 15.5.2.1 Introduction The potential radiological effects of plant accidents are analyzed by the evaluation of all physical factors involved in each chain of events which might result in radiation exposures to humans. These factors include the meteorological conditions existing at the 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 are identified and evaluated, and uncertainties in their values are discussed. Because some degree of uncertainty always exists in the prediction of these factors, it has become general practice to assume conservative values in making calculated estimates of radiation doses. For example, it is customarily assumed that the accident occurs at a time when very unfavorable weather conditions exist, and that the performance of the plant engineered safety systems is degraded by unexpected failures. The use of these unfavorable values for the various factors involved in the analysis provides assurance that each safety system has been designed adequately; that is, with sufficient capacity to cover the full range of effects to which each system could be subjected. For this reason, 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 are coincidentally assumed to occur has been called a design basis accident (DBA). The calculated doses for the DBA, provide a basis for determination of the design adequacy of the plant safety systems. In the process of safety review and licensing, the radiation exposure 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 criteria proposed in regulatory guidance, and if these calculated exposures fall below the regulatory 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... prvde a...relsi do....not... enosesngteepce in tho ,ernidPnft which *re osfimafos of tho 2ntuaI ,J~hme *nt*'. efnnd to oc'ir if tho accidcnt took place. The resulting doses were close to the doses expected to result use tIhe* customary',..,,

from an .accident of this* tpe. The, second',r,,, -,-case, tlhe* DBlA,,,

e-stimate÷ of expected*,. doses, can* pr,*,ovid a1bais o r deemiaio÷*m n,+,, of the desig~rn 15.5-8 15.5-8Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE As 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 the radionuclides, and the timing of the release of these nuclides), b) replaces the previous accident source term used in all design basis radiological analyses, and c) incorporates the Total Effective Dose Equivalent (TEDE) criteria of 10 CFR 50.67, and Section II of Standard Review Plan 15.0.1, Revision 0.

The dose consequences of the following accidents have been re-evaluated using AST in accordance with Regulatory Guide 1.183, July 2000.

1. Loss of Coolant Accident (LOCA) - Section 15.5.17

2. Fuel Handling Accident (FHA) - Section 15.5.22

3. Locked Rotor Accident (LRA) -Section 15.5.21

4. Control Rod Ejection Accident (CREA) - Section 15.5.23

5. Main Steam Line Break (MSLB) - Section 15.5.18

6. Steam Generator Tube Rupture (SGTR) - Section 1.5.5.20

7. Loss-of Load (LOL) Event -Section 15.5.10 The 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 tanks resulting from normal plant operations, thus the source term characteristics of AST are not applicable to these events.

The dose consequences for the remaining accidents are addressed by qualitative comparison to the seven accidents listed above (with the exception of the tank rupture events).

Note reference to Regulatory Guide 1.183, July 2000 is used extensively within this section, as a result any reference to "Regulatory Guide 1.183" within Section 15.5 refers to Regulatory Guide 1.183, July 2000.

The methodology used to assess the dose consequences of the DBAs, including the specific values of all important parameters, data, and assumptions used in the radiological exposure calculations are listed in the following sections. The computer programs used to assess the dose consequences of the DBAs are described briefly in Section 15.5.8.

As discussed previously, certain- radiological source terms for accidents and some of the releases resulting from Condition I and Condition II events have been included in Chapter 11.

15.5.2.3 Dose Acceptance Criteria EAB and LPZ Dose 15.5-9 15.5-9Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE The dose acceptance criteria presented below for the EAB and LPZ reflect use of AST and are applicable to all accidents with the exception of the tank rupture events. The tank rupture events are evaluated against 100 CFR100.11 (refer to Sections 15.5.1.1 and 15.5.24 through 15.5.26 for detail)

The acceptance criteria for the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) Dose are based on 10 CFR 50.67, and Section 4.4, Table 6 of Regulatory Guide 1.183:

(1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, shall not receive a radiation dose in excess of the accident-specific TEDE value noted in Reference 55, Table 6.

(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a radiation dose in excess of the accident-specific TEDE value noted in Reference 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 Ill scenarios. However, per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST allows a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all dose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of Regulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. Thus, the dose consequences at the EAB and 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 Dose The 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 the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

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 Dose 15.5-10 15.5-10Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE The 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 CER 50.67. The dose to an operator in the TSC should not exceed 5 remn TEDE for the duration of the accident.

15.5.2.4 Dose Calculation Methodology The dose calculation methodology presented below reflects use of AST and is applicable to all accidents with the exception of the tank rupture events. The methodology used for the tank rupture events are discussed in the accident specific sections, i.e., Sections 15.5.24 through 15.5.26.

15.5.2.4.1 Inhalation and Submersion Doses from Airborne Radioactivity Computer Code RADTRAD 3.03 is used to calculate the committed effective dose equivalent (CEDE) from inhalation and the effective dose equivalent (EDE) from submersion 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 Section 4.1 .4 of Regulatory Guide 1.183.

The CEDE is calculated using the inhalation dose conversion factors provided in Table 2.1 of Federal Guidance Report 11 (Reference 41).

The submersion EDE is calculated using the air submersion dose coefficients provided in Table 111.1 of Federal Guidance Report 12 (Reference 42). The dose coefficients are derived based on a semi-infinite cloud model. The submersion EDE is reported as the whole body dose in the RADTRAD 3.03 output.

RADTRAD 3.03 includes models for a variety of processes that can attenuate and/or transport radionuclides. It can model the effect of sprays and natural deposition that reduce the quantity of radionuclides suspended in the containment or other compartments. In addition, it can model the flow of radionuclides between compartments within a building, from buildings into the environment, and from the environment into a control room. 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 whole body dose, the thyroid dose, and the TEDE dose (rem) to the public located offsite, and to onsite personnel located in the control room due to inhalation and submersion in airborne radioactivity based on user specified, fuel inventory, nuclear data, dispersion coefficients, and dose conversion factors. Note that the code uses a numerical solution approach to solve coupled ordinary differential equations. The basic equation for radionuclide transport and removal is the same for all compartments. The program breaks its processing into 2 parts a) radioactive transport and b) radioactive decay and daughter in-growth.

Computer Code PERC2 is used to calculate the CEDE from inhalation and the EDE from submersion due to airborne radioactivity in the TSC. PERC2 is a multiple 15.5-11 15.5-11Revision 19 May2010

DCPP UNITS 1 & 2 FSAR UPDATE compartment activity transport code with the dose model consistent with Regulatory Guide 1.183. The decay and daughter build-up during the activity transport among compartments and the various cleanup mechanisms are included. The CEDE is calculated using the Federal Guidance Report No.11 (Reference 41) dose conversion factors. The EDE in the TSC is based on a finite cloud model that addresses buildup and attenuation in air. The dose equation is based on the assumption that the dose point is at the center of a hemisphere of the same volume as the TSC. The dose rate at that 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 energy release rates per energy group from activity airborne in the TSC, and the ANSI/ANS 6.1.1-1991 neutron and gamma-ray fluence-to-dose factors. (Reference 84)

Offsite Dose In accordance with Regulatory Guide 1.183, for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of the public located offsite is assumed to be 3 .5xl0A m 3/sec. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is assumed to be 1 .8x1 0- m 3 /sec. After that and until the end of the accident, the rate is assumed to be 2.3x10-4 m 3 /sec. The maximum EAB TEDE for any two-hour period following the start of the radioactivity release is calculated and used in determining compliance with the dose criteria in 10 CFR 50.67. The LPZ TEDE is determined for the most limiting receptor at-the outer boundary of the low population zone and is calculated for the entire accident duration.

Control Room Dose The control room inhalation CEDE is calculated assuming a breathing rate of 3.5x1 0 4 m 3 /sec for the duration of the event. The following occupancy factors are credited in determining the control room TEDE: 1 .0 during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 0.6 between 1 and 4 days, and 0.4 from 4 days to 30 days. The submersion EDE is corrected for the difference in the finite cloud geometry in the control room and the semi-infinite cloud model used in calculating the dose coefficients. The following expression obtained from Regulatory Guide 1.183 is used in RADTRAD 3.03 to correct the semi-infinite cloud dose, EDEnO, to a finite cloud dose, EDEi'inite, where the control 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 ED~y,.** - 1173 Technical Support Center Dose The TSC inhalation CEDE is calculated by computer code PERC2 assuming the same breathing rate and occupancy factors as those used in determining the control room dose. The submersion EDE developed by PERC2 (which computes the photon fluence at the center of TSC and utilizes the ANSI/ANS 6.1 .1-1991 fluence to effective dose conversion factors), is a close approximation of the dose determined using Table 111.1 of Federal Guidance Report No. 12 (Reference 42) (refer to Section 4.1.4, 15.5-12 15.5-12Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE factor given in Regulatory Guide 1.183) and adjusted by the finite volume correction Regulatory Guide 1.183, Section 4.2.7.

15.5.2.4.2 Direct Shine Dose from External and Contained Sources dose equivalent Computer program SW-QADCGGP is used to calculate the deep to external and contained sources (DDE) in the control room, TSC and at the EAB due (CEDE) and the following a LOCA. The calculated DDE is added to the inhalation the final TEDE.

submersion (EDE) dose due to airborne radioactivity to develop models are prepared to Conservative build-up factors are used and the geometry The dose ensure that un-accounted streaming/scattering paths were eliminated.

the scatter dose in albedo method with conservative albedo values is used to estimate situations where the scattering contributions are potentially significant.

flux to the dose ANSI/ANS 6.1.1-1977 (Reference 83) is used to convert the gamma equivalent rate.

The.pecifi ....

value.. o...f all impo,, ant parameters ,*r data, and,, a...umptions used, in the er{rrc Tkhe ,~4.-',i;l *{ +kR v,'M ,I ra r'.,I',,lt* ,I', r* erc I~cfA  ;,4 +kn {'a*,ll-*,n rcAr,Ir*l*;,-ii .............~........

implementation of the equations, models, the original licensing basis computer code Snec 'I) and4 the EMERALD Nr M/RAL the EMERALD computer program (Refere :re described briefl,. in Section -1558+:L computer program (Reference 5), which a 15.5.3 ACTIVITY INVENTORIES IN THE PLANT PRIOR TO ACCIDENTS 15.5.3.1 Design Basis Accidents Excluding Tank Ruptures gaps, and the primary The fission product inventories in the reactor core, the fuel rod based on plant coolant prior to an accident have been conservatively calculated of 3411 MWth, with operation at 105% of the current licensed rated thermal power current licensed values of fuel enrichment and fuel burnup. ,,usk.,,*+*,._h...e,-,,,..

1 1.1 12 b the= EMRALDA[ NO*RMAL code, except for slight` d4iffrences- in some in* the*: accdet~pP nucidesdue lto di,4frferent initial co,-re inve*ntories and irradPia{tion times*

secondar,' syste inventories are, lis.ed in{a , Tablea 11.23 I_' t shou"d be noted that+these-15.5-13 19 May 2010

15. 5-13Revision

DCPP UNITS 1 & 2 FSAR UPDATE stem,,,e f...........e..

andmasses ..... lum1,pcd vaue,* u..ed for

,pproximat .., acti+it, ma.se.. WAh;ie thes. amdequate for activty balance..,

alues re.. the s..hould!, not+be 15.5.3.1.1 Core Activity Inventory In accordance with Section 3.1 of Regulatory Guide 1.183, the inventory of fission products in the reactor core available for release to the containment following an accident should reflect maximum full power operation of the core with the current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty in the 10OCFR50 Appendix K analysis (typically 1.02).

The equilibrium core inventory is calculated using computer code ORIGEN-S. The calculation is performed using the Control Module SAS2 of the SCALE 4.3 computer code package. The SAS2 control module provides a sequence to calculate the nuclide inventory in a fuel assembly by calling various neutron cross section treatment modules and 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 for all major nuclear interactions including fission, activation, and various neutron absorption reactions with materials in the core. It calculates the neutron-activated products, the actinides and the fission products in a reactor core.

The reactor core consists of 193 fuel assemblies with various Uranium-235 enrichments. Per control imposed by DCPP core-reload design documentation, the peak 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 for variation of U-235 enrichment in fresh fuel, the radionuclide inventories were calculated for a 4.2% average enriched core (representing minimum enrichment at DCPP), and 5%

average enriched core (representing maximum enrichment). The higher activity for each isotope from the above two enrichment cases is chosen to represent the inventory of 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 fuel cycle, 1/3 of the core to two fuel cycles and 1/3 of the core to three fuel cycles. This approach has been demonstrated to develop an isotopic core inventory that is a reasonable and conservative approximation of a core inventory developed using DCPP specific fuel management history data. Minor variations in fuel irradiation time and duration of refueling outages will have a slight impact on the estimated inventory of long-lived isotopes in the core. However, these inventory changes will have an insignificant impact on the radiological consequences of postulated accidents. A 4%

margin has been included in the final isotopic radioactive inventories in support of bounding analyses and to address minor changes in future fuel management schemes.

15.5-14 15.5-14Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE A 19 month fuel cycle length was utilized in the analysis. The 19-month average fuel cycle is an artifact of the current DCPP fuel management scheme which specifies 3 fuel cycles every 5 years 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 with lower enrichment is a common industry practice when replacing assemblies previously irradiated but proven unsuitable for continued irradiation. As these assemblies are designed to replace higher enrichment assemblies with ones of similar reactivity for the remainder of the fuel cycle, their inventory is enveloped by the isotopic core average inventory developed to support the dose consequence analyses.

iii. A maximum core average burnup of 50 GWD/MTU.

The core inventory developed by ORIGEN-S using the above methodology includes over 800 isotopes. The DCPP equilibrium core fission product inventory of dose significant isotopes relative to LWR accidents is presented in Table 15.5-77.

15.5.3.1.2 Coolant Activity Inventory

1. Desig~n Basis Primary and Secondary Coolant Activity Concentrations Computer code, ACTIVITY.2, is used to calculate the design basis primary coolant activity concentrations for both DCPP Unit 1 and Unit 2 based on the core inventory developed using ORIGEN-S and discussed in Section 15.5.3.1. The source terms for the primary coolant fission product activity include leakage from 1% fuel defects and the decay of parent and second parent isotopes. The depletion terms of the primary coolant fission product activity include radioactive decay, purification of the letdown flow and neutron absorption when the coolant passes the reactor core. The nuclear library includes 3 rd order decay chains of approximately 200 isotopes.

Computer code, IONEXCHANGER, is used to calculate the design basis halogen and remainder activity concentrations in the secondary side liquid. The source terms for the secondary side activity include the primary-to-secondary leakage in steam generators and the decay products of parent and second parent isotopes. The depletion terms of the secondary side liquid activity include radioactive decay, and purification due to the steam generator blowdown flow, and continuous condensate polishing.

The design basis noble gas concentrations in the secondary steam are calculated by dividing the appearance rate (pJCi/sec) by the steam flow rate (gm/sec). The noble gas appearance rate in the steam generator steam space includes the primary-to-secondary leak contribution and the noble gas generation due to decay of halogens in the SG liquid. The activity concentrations of the other isotopes in the steam are determined by 15.5-15 15.5-15Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE the SG liquid concentrations and the partition coefficients recommended in NUREG 0017, Revision 1 (Reference 56).

2. Technical .Specification Primary and Secondary Coolant Activity Concentrations In accordance with Technical Specifications the primary coolant Technical Specification activities 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 equilibrium concentrations at 1% fuel defects.

b. Iodine concentrations based on the thyroid inhalation weighting factors for 1-131, 1-132, 1-133, 1-134, and 1-135 obtained from Federal Guidance Report 11 (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 obtained from Federal Guidance Report 12 (Reference 42)

The Technical Specification 1 pCi/gm DE 1-131 concentrations per nuclide in the primary coolant are calculated with the following equation:

D~

()(u~

11 i)=C(i)x C1,o, (15.5-1)

=,T__Z{FQ0xC(i)}

Where:

F(i) = DCF(i) / DCF E-131 DCF(i)= Federal Guidance Report-il, Table 2-1 (Reference 41) Thyroid Dose Conversion Factor 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 Technical Specification limit for full power operation), or 60 times the primary coolant total iodine Technical Specification concentration.

The accident initiated iodine spike activities are based on an accident dependent multiplier, times the equilibrium iodine appearance rate. The equilibrium appearance rates are conservatively calculated based on the technical specification reactor coolant activities, along with the maximum design letdown rate, maximum Technical Specification based allowed primary coolant leakage, and an assumed ion-exchanger iodine efficiency of 100%.

The Technical Specification secondary liquid iodine concentration is determined using methodology similar to that described above for the primary coolant where CTtot iS 15.5-16 15.5-16Revision 19 May 2010

DCPP UNITS I & 2 FSAR UPDATE 0.1 pCi/gm DE 1-131, and C(i) is the design basis secondary coolant equilibrium concentrations per nuclide.

The Technical Specification noble gas concentrations for the primary coolant are based on 270 pci/gm DE Xe-133. The DE Xe-I133 for noble gases is calculated as follows:

DEX 133 .=2{F(i) x C(i)} (15.5-2)

Where:

F(i) = DCF(i) / DCF Xe-133 DCF(i) = EPA Federal Guidance Report No. 12 (Reference 42) Table II1.1, Dose Coefficient per Nuclide [(rem-m 3)/(Ci-sec)I C(i) =design basis primary coolant equilibrium noble gas concentration per nuclide (pJCi/gm)

The noble gas and halogen primary and secondary coolant Technical Specification Activity 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 15.5-79 15.5.3.1.3 Gap Fractions for Non-LOCA Events Regulatory Guide 1.183, July 2000, Table 3 provides the gap fractions for Non-LOCA events that are postulated to result in fuel damage for AST applications. The referenced gap fractions are contingent upon meeting Note 11 of Table 3 of Regulatory Guide 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 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU." The burnup criterion associated with the maximum allowable linear heat generation rate is applicable to the peak rod average burnup in any assembly and is not limited to assemblies with an average burnup that exceeds 54 GWD/MTU.

DCPP has three design basis non-LOCA accidents that are postulated to result in fuel damage, i.e., the Locked Rotor Accident (LRA), the Fuel Handling Accident (FHA) and the Control Rod Ejection Accident (CREA)

To support flexibility of fuel management, and establish dose consequences that take into consideration fuel rods that may exceed the Regulatory Guide 1.183, Table 3, Note 11 linear heat generation criteria, the fuel gap fractions provided in Table 3 of Draft Guide (DG)-1 199 (Reference 62) for all No n-LOCA events that are postulated to result in fuel damage with the exception of the CREA. This approach is acceptable (i.e., in lieu of developing plant specific fission gas release calculations using NRC approved methods and bounding power history to establish the gap fractions), since DCPP falls within, and intends to operate within, the maximum allowable power operating envelop 15.5-17 15 .5-17Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE for PWRs shown in Figure 1 of DG-1199.

consequences of the FHA In summary, the fuel gap activity fractions used to assess the dose and LRA are as follows:

FHA /LRA Nuclide Group (based on DG-1199)

I-131 0.08 1-132 0.23 Kr-85 0.35 Other Noble Gases 0.04 Other Halogens 0.05 Alkali Metals 0.46 In 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: 10%

Halogens: 10%

Refer to Tables 15.5-80 for the isotopic concentrations in the gap assumed for the LRA and CREA. The isotopic concentrations assumed for the FHA are presented in Table 15.5-47C.

15.5.3.2 Tank Rupture Events Activity inventories,, in.. variou.s ... ,waste,..' .... ystem*j,.,,÷" tanksused for the tank rupture events witi-beare cross-are alse-4istedprovided in sections of Chapters 11 and 12 and releases from these referenced in the sections of this chapter dealing with accidental tanks.

Drtr*I IIDEr'* TC'* DE Dr\I[C'rr I 11c'TrDI' 'BtAI l*IrI*D!.A ATIf'KI IlI ITAI IC'O DEIl CxAI kIrT during Refueling shutdown studios at operating Westin gho uce P WRs indicate that, i-L~IEd h

-I........ a b 1....... -i nf pS.......... -,c-i 1441 .. . .

.5 ............

.I SI.~1.nc-i .. . .... 4.1.....

O......... .... . nrc..................................

I

... .... ................ c " ~,V I, 4

~ r~ vr ~ r'c~d i- c-,s-~,c-c-nc-id! .f nn-,nn ;An ni- anti c-Mci, IA c-ni-inn cf i-Mn DC' 0 c-sc- i-Mn A, ,rrr. i-I, a 5 4 5 4 4 7 c-nc-c., ,r, Ann IEIS. I s.~ 54.4 LII'.4 I 541.41.411. 54I 541 I ,.454.,I5454I IL, I.AI ISA ,.41i54L4I1.A f 54.41.1541 I~.LJtI.dIl 541 sALI I EI~LI, of primary therefore be taken into account in the calculation of post accident re!~a~~~

coolant to the environment.

Table 15.5 1 illuotra toe the ant'~ipat~d coolant acti~'ity pc--i-A increases of ceveral isotopes for

,,~i-;nc' A, ,c-hr, c-i-anti,, c-i-ni-n F~iDO A. Ic-icI.o c'M~ li-An,,,,., TMc- #obIa Ic-.,Cc, I I ISA LI 5411.1 Il1.Jt1.J i-Mn avc-sac-.i-nA LI~ 54 54fl .454541.541.j 1.1541.1WILl 5454 541.541.5.41.54 I 54LIIII I~J 4111.41.1.454ill I. 1 ncn!,nnc- TA-sac-n An!-,

nnnnnnnAnn#n~n..,l-.,sA nnnInn,,,i-mnc-A,,nnnl..,ni-nnnlti-.,,,,n OIA/IZ i/'-"J ics c-u-i~l-sc ;n HAc,;nn i-n i-M c-c-. .i -- s . -- ssc c-i--finn-sm .-scn ,snc--shn,- 11.1 III SAl..'lJI~j II 1.54 (II r*m* tAll S..If.JI..lI 1.4LII I~I **IS LII 1.41. 154 541111111.41 1.4I 54 I.J1.A1..ll.IS4 *.JII III 545454I4I541II54I 11.54 11541$I 15.5-18 15. 5-18Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE

"'sigonificant÷f,,ol defects. "The me......dact,-iit. leel orf.

with* f..

DCPPand'has oper...

thc opratingflr p~lant are* also inc~hrluded4 in Tale*~l 15.5 1.

dep-rnss,,r;-,at,-,, is 131t.! Tlho a,.tivity* lvel in t he coor.lant÷ wasc obso.,n'or to be hig.her" system*h ther purificat.*ionratec varyng betwn approx,'~nc',imatly,'*Innlu~ap The ......o soc.... ,- ,-;,; .. leso tou . in.. agni...doter othor*-m onb prod"uct* p... ÷.

fsir,-

domir~'rrnoraizor dulring ln olo ngL p~ nts tindiat ha maxrrml m ino;~rctrthas of~c approximatlyt r--Frsengat~rl*f dt ;romoprati dprsurzai5pocdue coldwnan Alhuhastaysaevisio 133 a 21

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.4 EFFECTS OFl PILUiTONllIUM INVIENJTORv rON~ PO*TENTLq AC'IEn~kT DOSESDELETED BE REVISED.

HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO thrml isios nPu 239, estvt td aodce to d 2er3n th" osi 15,5.51 Deofthsig Bffsit Anpotentsa (Ecluding donsRptue.

This EtAd dand ther7atmosptht erc~d disperso fartonlys!g (t/Q atilzed by theds listingi TabulaoGie 15.542, ievisaon tha pluonieumenceantories. Thae resutngdiferenesd irraefom4t eprally eprcsentandv wholea berody dosf geourlly mthyoiddosesgende otnos from2t prete assumingt mteoraoiogita tocred Jatur 1,20.

mecreorosogcdat locathios study, toas core fisont yelromads wer avb byaueswihing ofhe23 calculated fissionspheric Reandv PuO23enfissionyils.i Becusedthe cores(AON6 messhofdo235gi consieranly g1.Reater the thectr 2.3.d5arcclo2.taso u29 oa oefsion U23 of thereleas . Thinmassesepof Uoc238n anepovddu21ta isin aigre extremewhly smabll, and14 thusd 2.3816 adpur41hvie essenmtialynoeft on the teesoitalecepore fissionationds.

15hat. DesignauaedBasies Accident (Ecldng Tank14 prupdetures)nrlromQvle frtheEAindivda thelPZatmsepheictdispeptrscmiaionsfacors QUtilizead inith dorspeciey ory-LGuide1e15asevisoint 1S consequence4analysens thae been dvelueopted uingiRegulat Untempoally retapresabentative5-yea perodma ofak hourlye meepthooogy andbiatontinuous Usingeo the TSam meteorologicavldataidthed alue appliabevlue tor ponstentaS houndrly locations aronhnelpovied. ineFnigur 2.3-5, while2cotable Allnfilthered nleasaepitadrcpo combsuinations for Uintak1 ando Unieth2 applCableito the TSeomalgintak ande Thes reporm 15.5-20 15.5-20Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE the z/Q presented in Tables 2.3-1 47 and 2.3-148 for the control room pressurization intakes inclusive of the credit for dual intake design and ability to select the more favorable 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-dose specific tables presented in Chapter 15.5. The x/Q values selected for use in the accident that consequence analyses are intended to support bounding analyses for an occurs at either unit. They take into consideration the various release points-receptors applicable to each accident in order to identify the bounding y/Q values and reflect the allowable adjustments and reductions in the values as discussed earlier and further summarized in the notes of Tables 2.3-147 through 2.3-149.

15.5.5.2 Tank Rupture Events For the analyses of offsite doses from the DBletank rupture events, the rare and Guide unfavorable set of atmospheric dilution factors assumed in the NRC-Regulatory 1 .4, Revision 1 (Reference 6) was used. On the basis of meteorological data collected at the DCPP site, these unfavorable dilution factors, assumed for the design bases of the cases, are not expected to exist for onshore wind directions more than 5 percent time. The particular values used for this site are given in Table 15.5-3.

Eforect anarlyeaseo odurateoo dowenwfrd theropcnd cavelaccientratinhae asuedn ofd amoephuricdireution facddtorermistedthoeincabley 15.5 4.norwtedge cates 10preonta bsspectrse nfumbuerscerohteu bsed.Onthatis ofstd o theor siterdatlyage tat thvestignl Becaushe hofizothelo coprobabilts ofocurruence nassgoiatd ihtesel ot~ assme diluti~cnt pmossible emergenyieracuatione tand in conenutrtonsy largew saratondss due atoedwnin 15.5-21 19 May2010 15.5-2 1Revision

DCPP UNITS 1 & 2 FSAR UPDATE dimension of the cloud need be modified for concentration estimates for noncontinuous releases. Slade (Reference 7) using the approach recommended by Cramer, gives a time-dependent adjustment of the lateral component of turbulence to be:

• 0(T) = 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 minutes Ge (To) = lateral intensity of turbulence measured over a time period T0 , where To is on the order of 10 minutes Near a source there is a direct linear relationship between GO and the plume crosswind dimension G-y so that the Gyversus distance curves presented by Slade can be directly scaled by the factor (T/To) " to provide estimates of a reference Gy at about 100 meters downwind from the source. Beyond this distance, the lateral expansion rates for continuous and noncontinuous point source releases are approximately the same, and thus the ratio of short-term release concentration to continuous release concentration for 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 volume of the source increases.

Using the above scaling concept, the dilution equation in Regulatory Guide 1.4, and the cloud dimension curves given by Slade, the ratio of short-term release concentration to continuous release concentration was calculated for several different release durations (Figure 15.5-1). For a 10-second duration, the short-term dilution factor is only 2.3 higher than the continuous release dilution factor, and thus the appropriate short-term release correction is within the uncertainty limits of the continuous release dilution factor.

The variou.s plntacidnt consdere in*,.*. 15.2,o15.

Sections,. !3, and 15.1 may resu.,lt.in.

acuvity release lnrougn various painways: conia:nmentsystem leaKage, seconoary sieam dumping, vontilation discharge, and radioactive waste discharge.

Post accident containment leakage is a slow continuous process, and thus continuous release 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 release significant quantities of activity only through relief valves. Relief valve flow limitations combined with large steam release result in activity releases of long duration. Thus continuous release dilution factors apply for these cases.

15.5-22 15.5-22Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE and the of-The release duration for liquid holdup tank rupture, gas decay tank ruptureT over inless than 10 volume control tank ruptureadfe handling...... are accident... are all for these minutesT-,-. As discussed above, continuous release dilution factors apply cases=

Contnuou,release* dIlution factor- hav ben app~lied to a1l Conditions II, IIl, and I1 10 CFR Part 100 limits Short ter releas dilutionh,. factors ar.... only about twice as.high .... continuou.. relea.e dilutiorn factors+r in the contnuousn~ releaset dilutio~n factrsf*-

Furt,-hermore, theaboe- reason... indic-*at that and a more sophisticated or complex short-term release dilution model is not justified.

Thtopei diserio fatr for praessurization and in~filtration air flows._to the:

• /Q methodology*, whic-h is Co*ntrol rooxm are analyzedq using the modifierd Halitskyx

• o^a result* of the TM!, accident the NRC, in NURE 073_n7 Secion III 1"3 .A, asked all1 roonm habitability designs usi~ng nucflear powe*r plants"to review* their post LOCA contfrol was ov*,e

• rly* ronseRantiv;a nd (7 /Q) nmethodology reco~mmended in the NI C paer The 5/I C equat-io~nsae brashed primarily on inappnropriate for most of the plant design.

ar va...lid onl,,

the Hali+s,, data for round topped. EBR-!! (PWP.tye c...ntinment. and building wake

.. I,*us wasL,. based..*

• , . a series on . of;'wi'"nd*

Historically, the, preliminap' work..*o.,n*,

7). In 19'71 K." Mrphy and and Atomic- Energ' 1 ' 198 D. H.1Slade Editor (Reference*

15.5-23 Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE actual bull4 ding. wake */Q measurements+,r haveg been co.-.nducted*r at' Rancho Seco*

cylindr,',cal containment... Inilrtin.i buildings*.,n,,-+;. into,. the contro room,..would come from abov t.. ,he highest:- roof elvaion. +' of the .... ilar,',builing*,- Pres....urization,,. air. for.,, the buldn... roofI and a.,portion,, of the,+turbi;ne builing;,, Wall. f aci~ng ..... a n*,d" the wall facing"'

2.-/ * "/Au ..... ... (15.5 2)

A- cross....* sectionalara, -~hgo#,.*a-t,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 ignored field tests 1 hr do.

Tvoicaltests. account for olume meander effects. while 3 to 5 minute. wind.

tu nl no.. .. i -

15.5-24 15.5-24Revision 19 May 2010

DCPP UNITS 1 & 2 ESAR UPDATE

-/ K x fxf _x-f,*,-fd~*f--(sec-m*) (15.5..3)

This, ,.,,dificd, H-alitsky,, methodolog isv:inhcr.... cos.......... be,, cause... the.ind is,4; as.umed to be blowh* in the cont,,rol room dur"ing the* first or..wor.t part of the t..o;ward....

acc idn,,- and. becas eretwndsed aeue;rte ha 0prcn.

ar alwas biaed toard te adjstmen additon, factrs the ini rdcinta fTs-fchc,-f. d e to"+'< tnhei uncert.inty, wr compared* toth 1hor.ied e, x<.at masured- ;int some cas-esigtht was-! rsigiicnly higheorsecn.~ n hnawn !t 1

The .hoicof Kators and.t*÷;,he suggste modifying factors t÷ fl*, ,; etc. are. dsussed,,

be~Ow, K- a s*r Jl; . .. .. ;ool.. ... k ae th!xQ estimate..r to. bevaid... iThky in Reference,. ... ha severa sets, ofrm*

K iopethfr run tppe.cntanmnt (frRev)an lokbidiongs9 (for01

DCPP UNITS 1 & 2 FSAR UPDATE K 1-t ,,*,rlmcte- .,eonr,*l\ Bas *., *,("Q eee,*',l*

Ca.se Pressu riza:tio n 4 1 3690 1.084x! 04 Infiltration 5 1 1661 3.0!x!04 tr7 -ind-peed r'nntiinm~nt Ar hiuilclrir-i

,..~. T rafv-L~ fl~a Nil C' nr~u-r'ant I VVI I uuuir~ eu-.~a~rI .~

IA ka 4aA +a +k.~ ',,4, al A a4 fka -. ,-~-.+

4A rna+,-~.- I.uak4 rd.au

• .....~,..'

n a r.1u ~ .~ ;p,ti 4k a ~r. in,.,~.. In4~r.

ar rala',ea n~nG Tha C nr. in, r.

• i,,, A n r,,-*r*A

• j*v* *,,,* *,,* ,*,,,.*,*,*,,

t(,,w Il X'*%*'..*11,~t.'*.. LAL- *AL t~-,,L tIILAW*1..

ujp .,.*,l%. uL,,.r,., L¥ y

• r Vll um.,Jul I T

- (z I~ZRef I

"~ ,~, -X I',t IC l *./ . '.3 A\

whe rc:

-u-n-- wind speed at height -z~

-- Kel T *'~ a , ,t a ks:A a l~,~ A A~'~t,nnI h ,, -r~,m.. . h . .. InI' aatL pi-

,,, Jt. A

,T -ha ... ,r r.+

meteorological data for a 10 year period of record.

. 0 1--.-0 0 8hhrs1 24 hrs 0.83 ---- 0.92 8

0.48 0.---Q7 96 720 hrs f_ in f"rh!.r' -. - .

his Wilson in Reference 30 and field tests confirm Halitsky's statement that of 5 to 10 too conservative due to not accounting for K icoeleths are a factor rt'ha **,irnr i .,nnnr~ro.hrmn tha= he 1 IAt;nn ILJ1m ul Lli'u*1 Tharafarar u~lt~utL -, far-fa~r LLfl ,--,u',Oi rnn~r'irn thtfuufu,-+ranc.n i iiT--..i*iLiiT--.. ,* *,,.;bi;Ji b';

• ,..*l.ij.iiidG,*iiiii*.i *,,v *i.liiL.,,*,ui.

[.rLii iii'.., iiiii*.*

0.2 vv~as u.sed,+,,.. foa t*.

f fllfltfl#flA raIn an4'ani~

ii (RAf~rAnm 31Yinc1in~it~ thM thr~rA r~rn nfl tA 10 null ~"ind PAl PAJm~C*t~tI~r r4 flu irnn *haea narnAr. +ka c.raarl ,.,-,nA;+;anc- ,-Juurnn ., n hau ur ni A.,f-, r'rIIar'tianS. L..LAI Euu,.4 LE u....J*.. L.'St~S.L4VJ Lu iSp S4%Ai i* SI LAu S I SttAu Vu'.AVLtAL*.JilS..~..LiS II uJS.sA'.IStSAiISA ILi*tt I i II + iI I*

I enct a etmoeu,m pume rise and* D,,,a,,y .. ou...' result in m1e-radiactie*

15.5-26 15.5-26Revision 19 May 2010

DCPP UNITS 1 & 2 ESAR UPDATE th wk 2ca ny,A reduction facto-ra *,., C_

of 1' wa.. used* *t fen&teringCp*

AE NHLAIO (15.55) 1556 15.5.6.1D sig - protti A p cci entratxluiongTn Rutrs brathin The re moed ntelcn cntations fihlto oe r itdi otes prt vause amplingd time aeaediybetigrtspo e Tabl15.-7A inSetormn wind3tunnegldatar Gisdaen for83.t 0mnt ape.Tufra1hu asviumed19for 201 CmA1aueo.55-a27neraivl

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.6.2 Tank Rupture Events The breathing rates used in the calculations of inhalation doses are listed in Table 15.5-7* These values are based on the average daily breathing rates assumed in ICRP Publication 2 (Reference 8) which are also used in Regulatory Guide 1.4, Revision 15.5.7 DELETED POPULATION DISTRIBUTION dist.ribu tio te u...ed, is I;-*-in! Table 15.5 8. The actual post accident÷ population;,

,re weI*

dit~i*ributl ionlr cold*l bet' s'l!ignif'ican'lfr.3rtly llowel l~r if* a'nyl evacuatlion' p,*}lanr~ imlementedr.*÷ 15.5.8 RADIOLOGICAL ANALYSIS PROGRAMS 15.5.

8.1 DESCRIPTION

of the EMERALD (Revision I) and EMERALD-NORMAL (Tank Rupture Events)-P-!egfam EMERALD is used to develop the source term for the tank rupture events and assess the dose consequences at the EAB and LPZ following a waste gas decay tank rupture and a volume control tank rupture.

The EMERALD program (Reference 4) is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large PWR. The approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant that contains a radioactive material is represented by a subroutine, which keeps track of the production, transfer, decay, and absorption of radioactivity in that volume. During the course of the analysis of an accident, activity is transferred from subroutine to subroutine in the program as it would be transferred from place to place in the plant. Fo e...mple, in the ca,,-,lculation of the* dose r.sulting from*,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 at various distances for individual isotopes and sums of isotopes. The program contains a library of physical data for 25 isotopes of most interest in licensing calculations, and other isotopes can be added or substituted. Because of the flexible nature of the simulation approach, the EMERALD program can be used for most calculations involving the production and release of radioactive materials, including design, 15.5-28 15.5-28Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE operational and licensing studies. The complete description of the program, including models and equations, is contained in Reference 4.

The EMERALD-NORMAL program (Reference 5) is a program incorporating the features of EMERALD, but designed specifically for releases from normal and near-normal operating conditions. It contains an expanded library of isotopes, including all those of interest in gaseous and liquid environmental exposures. Models for a radwaste system are included, using the specific configuration of radwaste system components in the DCPP. The program contains a subroutine for doses via liquid release pathways developed by the Bechtel Corporation and a tritium subroutine. The code calculates activity inventories in various radwaste tanks and plant components which are used for the initial conditions for accidents involving these tasks. In addition, it is used in some near-normal plant conditions classified in this document as Condition I and Condition II and discussed in Chapter 11.

15.5.8.2,.,D,.o.cc of, thc LOCADOSE-P-Fegram

.,rip...i'on*

The LOCADOSE program (Reference 47) is designed to calculate radionuclide activities, integrated activities, and releases from a number of arbitrarily specified regions. One region is specified as the environment. Doses and dose rates for five organs (thyroid, lung, bone, beta skin, and whole body) can be calculated for each region, and for a number of offsite locations with specified atmospheric dispersion factors. The control room can be specified as a special region for convenience in modeling airborne doses to the control room operators.

LOCADOSE is also used to assess the dose consequences at the EAB and LPZ following a liquid holdup tank rupture.

15.5.8.3 nELETED:nn....ript,., of,the* ORIGEN_2 Program...

The core inventor; and gamma ray energy spectra of post accident fission products for selected accidents (See Section 15.5.1) were computed using the ORIGEN 2 computer program. ORIGEN 2 (Reference 50) is a versatile point depletion and decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions of materials contained therein. This code represents a revision and update of the original ORIGEN computer code which has been distribut ed world ~A~ide beginning in the early 1970s. Included in it arc provisions for incorporating data generated by more

~ophisticated reactor physics codes, free format input, the abili~' to simulate a wide variety of fuel cycle flowshcets, and more flexible and controllable output features 15.5.8.4 DELETEDDescription of the ISOSHLD Program ISOSHLD (Reference 9) is a computer code used to pe~orm gamma ray shielding calculations for isotope sources in a wide variety of source and shield configurations.

tt~nuation calculations arc performed by point kernel integration; for most geometries 15.5-29 15.5-29Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE this is done by Simpson's rule numerical integration. Source strength in uniform or (where ~nnlicable~ m~v he ealctthted by the linked fis~on e~nonential distribution \~rE.I~-----------

a**pa-rt*;.,icular..= .. ,.,sh;ied points....;"<, the effective a*tomic,,,. number of*,,..,'* (t*/he region**"

sour...ce..... and detector*+*"

unes lasote~isechsen, nd hepoit iotopi Ncler eveopentAsocite 15.5.8.5 Description of the ISOSHLD IIProgram ISOSHLD II (Reference 11) is a shielding code that is principally intended for use in calculating the radiation dose, at a field point, from bremsstrahlung and/or decay gamma rays emitted by radioisotope sources. This program, with the newly-added bremsstrahlung mode, is an extension of the earlier version (ISOSHLD). Five shield regions can be handled with up to twenty materials per shield; the source is considered to be the first shield region, i.e., bremsstrahlung and decay gamma rays are produced only in the source. Point kernel integration (over the source region) is used to calculate the radiation dose at a field point.

ISOSHLD II is used to determine the dose to the control room operator due to direct shine from the airborne activity inside the containment following a LOCA during daily ingress / egress for the duration of the accident.

15.5.8.6 DrLE/TEDpescri-pt-in of, the RADTRAD Pogr...m ORIDTR-D (Referencte 52 uses L a somintioof tables andch numericalmodels byofig theCtierome dependarientdospuerataluser o soreN emaeutionalphenomenarto(determine spliedsn evlocations.fo SAS give accidntroscnro.ul thals providesah invuenetorcadecae chainucanddoe covnversion factorl taseblesb neededg foriteouse calulaton.os Theto cotrolaromn aouls elas sthe bxouentialdoeandi to-esltimat thmosue attGENuatondu 15587SAS2 / ORIGEN-S(Rfrne6)aluaethtmeeenetnurnlxanth NationldLabofislrato-ryOnLum forlthesNRCto perfoprmy sacondadie oamputer anaclyesafr It interactions including fission, activation, and various neutron absorption reactions.

15.5-30 15.5-30Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE can calculate accurately the neutron-activated products, the actinides and the fission products in a reactor core.

SAS2/ORIGEN-S is used to develop the equilibrium core activity inventory and the decayed fuel inventories after shutdown utilized to assess the design basis accidents excluding the tank ruptures.

15.5.8.8 ACTIVITY2 ACTIVITY2 (Reference 65) calculates the concentration of fission products in the fuel, coolant, waste gas decay tanks, ion exchangers, miscellaneous tanks, and release lines to the atmosphere for a pressurized water reactor system. The program uses a library of properties of more than 100 significant fission products and may be modified to include as many as 200 nuclides. The program output presents the activity and energy spectrum at the selected part of the system for any specified operating time ACTIVITY2 is used to develop the reactor coolant activity inventory (design and as limited by the plant Technical Specifications) utilized to assess the design basis accidents excluding the tank ruptures.

15.5.8.9 IONEXCHANGER ION EXCHANGER (Reference 66) calculates the activity of nuclides in an ion exchanger 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 with known radionuclide activities, it calculates the activity of each nuclide and its products in the ion exchanger or tank at some later time. The program also calculates the specific gamma activity for each of the seven fixed energy groups.

1ONEXCHANGER is used to develop the secondary coolant activity inventory (design and as limited by the plant Technical Specifications) utilized to assess the design basis accidents excluding the tank ruptures.

15.5.8.10 EN 113, Atmospheric Dispersion Factors EN-i113 Atmospheric Dispersion Factors (Reference 73) calculates z/ values at the EAB 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 specified time period and set of site-specific and plant-specific parameters. A release point can be identified as either of two types of release (i.e., ground or elevated), time periods for which sliding averages are calculated (i.e., 1 to 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> and/or annual average),

applicable short-term building wake effect, meandering plume, long-term building height wake effect, and a wind speed value to be assigned to calm conditions. Downwind distances can be assigned for each of the sixteen 22.5-degree sectors for two irregular boundaries and for ten additional concentric boundaries used only in the annual average calculation. EN-i113 performs the same calculations as the NRC PAVAN code 15.5-31 15.5-31Revision 19 May 2010

DCPP UNITS 1 & 2 FSARUPDATE except that EN-I113 using hourly calculates meteorological x/Qwhereas data values for the various PAVAN uses aaveraging periods joint frequency directly of distribution wind speed, wind direction, and stability class.

EN-i113 is used to develop the DCPP site boundary atmospheric dispersion factors utilized to assess the design basis accidents excluding the tank ruptures.

15.5.8.11 AROON96 ARCON96 (Reference 74) was developed by Pacific Northwest National Laboratory (PNNL) for the NRC to calculate relative concentrations in plumes from nuclear power plants at control room air intakes in the vicinity of the release point. ARCON96 has the ability to evaluate ground-level, vent, and elevated stack releases; it implements a straight-line Gaussian dispersion model with dispersion coefficients that are modified to account for low wind meander and building wake effects. The methodology is also able to 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 of the diffuse or area source. Hourly, normalized concentrations (x/Q) are calculated from hourly meteorological data. The hourly values are averaged to form x/Qs for periods ranging from 2 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> in duration. The calculated values for each period are used to form cumulative frequency distributions.

ARCON96 is used to develop the control room and TSC atmospheric dispersion factors utilized to assess the design basis accidents excluding the tank ruptures.

15.5.8.12 SWNAUA SWNAUA (Reference 67) is a derivative of industry computer code NAUN/Mod 4 which was originally developed in Germany and was based on experimental data. NAUA/Mod 4 addressed particulate aerosol transport and removal following a LOCA at an LWR. It developed removal coefficients to address physical phenomena such as gravitational settling (also called gravitational sedimentation), diffusion, particle growth due to agglomeration, etc using time-dependent airborne aerosol mass. NAUA4 (included in the NRC Source Term Code Package) was used by NRC during the initial evaluations of post-TMI data. NAUA/Mod 4 was modified to include spray removal and diffusiophoretic effects suitable for design basis accident analyses. A version of SWNAUA (SWNAUA-HYGRO) was proven to be the most reliable of more than a dozen international entries, in making predictions of aerosol removal for the LWR Aerosol Containment Experiments (LACE) series.

SWNAUA is used to develop the time dependent post LOCA particulate aerosol removal coefficients in the sprayed and unsprayed regions of containment.

15.5-32 15 .-32Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.8.13 RADTRAD 3.03 RADTRAD 3.03 (Reference 68) is a NRC sponsored program, developed by Sandia National Labs (SNL). It can be used to calculate radiological doses to the public, plant operators and emergency personnel due to environmental releases that resulting from postulated design basis accidents at light water reactor (LWR) power plants. The RADTRAD 3.03 (GUI Interface Mode) includes models for a variety of processes that can attenuate and/or transport radionuclides. It can model sprays and natural deposition that reduce the quantity of radionuclides suspended in the containment or other compartments. It can model the flow of radionuclides between compartments within a building, from buildings into the environment, and from the environment into a control room). 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 public located offsite and to onsite personnel located in the control room due to inhalation and submersion in airborne radioactivity based on user specified, fuel inventory, nuclear data, dispersion coefficients, and dose conversion factors.

RADTRAD is used to develop the TEDE dose to the public located offsite and to onsite personnel located in the control room due to inhalation and submersion in airborne radioactivity following design basis accidents excluding tank ruptures 15.5.8.14 PERC2 PERC2 (Reference 69) is a multi-region activity transport and radiological dose consequence program. It includes the following major features:

(1) Provision of time-dependent releases from the reactor coolant system to the containment atmosphere.

(2) Provision for airborne radionuclides for both TID and AST release assumptions, including daughter in growth.

(3) Provision for calculating the CEDE to individual organs as well as EDE from inhalation, DDE and beta from submersion, and TEDE.

(4) Provisions for tracking time-dependent inventories of all radionuclides in all control regions of the plant model.

(5) Provision for calculating instantaneous and integrated gamma radiation source strengths as well as activities for the inventoried radionuclides to permit direct assessment of the dose from contained / or external sources for equipment qualification, vital area access and control room and EAB direct shine dose estimates.

15.5-33 15.5-33Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE PERC2 is used to calculate the accident energy release rates and integrated gamma energy releases versus time for the various post-LOCA external and contained radiation sources. This source term information is input into SWV_QADCGGP to develop the direct shine dose to the control room. PERC2 is also used to develop the decay heat in the RWST and MEDT and develop the TEDE dose to personnel located in the TSC due to inhalation and submersion in airborne radioactivity following LOCA.

15.5.8.15 SW-QADCGGP SW-QADCGGP (Reference 70) is a variant of the QAD point kernel shielding program originally written at the Los Alamos Scientific Laboratory by R. E. Malenfant. The QADCGGP version implements combinatorial geometry and the geometric progression build-up factor algorithm. The SW-QADCGGP implements a graphical indication of the status of the computation process.

SW-QADCGGP is used to develop the direct shine dose to the operator in the control room, TSC and EAB.

15.5.8.16 GOTHIC GOTHIC (Reference 71) is developed and maintained by Numerical Applications Incorporated (NAI) and an integrated, general purpose thermal-hydraulics software package for design, licensing, safety and operating analysis of nuclear power plant containments and other confinement buildings. GOTHIC solves the conservation equations for mass, momentum and energy for multicomponent, multi-phase flow in lumped parameter and/or nmulti-dimensional geometries. The phase balance equations are coupled by mechanistic models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow, as well as single phase flows. The interfac:e models allow for the possibility of thermal non equilibrium between phases and unequal phase velocities, including countercurrent flow. Other phenomena include models for commonly available safety equipment, heat transfer to structures, hydrogen burn and isotope transport.

GOTHIC is used to estimate the containment and sump pressure and temperature response with recirculation spray, the temperature transient in the RWST / MEDT gas and liquid due to incoming sump water leakage / inflow / decay heat from the RWST /

MEDT fission product inventory, and the volumetric release fraction transient from the RWST /MEDT gas space to the environment.

15.5.9 CONTROL ROOM DESIGN AND TRANSPORT MODEL The 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) are made of 3'-0" concrete, whereas as the control room east and west walls are made up 15.5-34 15.5-34Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE of 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 HVAC room is located adjacent to the control room (east side), at El 154'-6".

The control room has a normal intake per unit (each located on opposite sides the auxiliary building; i.e. north and south), and a pressurization flow intake per unit (each located on either side of the turbine building; i.e. north and south). The control room pressurization air intakes have dual ventilation outside air intake design as defined by Regulatory Position C.3.3.2 of Regulatory Guide 1.194,. June 2003 (refer to Section 2.3.5.2.2)

During normal operation (CRVS Mode 1), both control room normal intakes are operational. Redundant PG&E Design Class I radiation monitors located at each control room normal intake have the capability of isolating the control room normal intakes on detection of high radiation and switching the control room ventilation system (CRVS) to Mode 4 operation (i.e., control room filtered intake and pressurization).

CRVS Mode 4 operation utilizes redundant PG&E Design Class I radiation monitors located at each control room pressurization air intake and the provisions of acceptable control logic to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident in accordance with Regulatory Guide 1.194, June 2003. Thus, during Mode 4 operation the dose consequence analyses can utilize the x/Q values for the more favorable 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 CRVS Mode 4, however, it initiates Containment Isolation Phase A which initiates Mode 4 operation.

During normal operations, unfiltered air is drawn into the control room envelope (refer to Table 15.5-81) from the Unit 1 and Unit 2 normal intakes. In response to a control room radiation monitor or SIS, the control room switches to CRVS Mode 4 operation, and control logic ensures that the CRVS pressurization fan of the non-accident unit is initiated and air is taken from the less contaminated of the Unit 1 or Unit 2 control room pressurization air intakes. The control room pressurization flowrate used in the dose consequence 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, all pressurization flow is filtered.

The allowable methyl iodide penetration and filter bypass for the CRVS Mode 4 Charcoal 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 factor of 2 is used in determining the charcoal filter efficiency for use in safety analyses (refer to Section 9.4.1 and Table 9.4-2. Thus the control room charcoal filter efficiency for 15.5-35 15.5-35Revision 19 May 2010

DCPP UNITS 1 & 2FSAR UPDATE elemental 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 efficiency particulate air (HEPA) filters in Technical Specifications is a "penetration plus system bypass" < 1.0%. Similar to the charcoal filters, the HEPA filter efficiency for particulates used 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 the recirculation flow filtered through the same filtration unit as the pressurization flow.

Refer to Table 15.5-81 for a summary of recirculation flow rates.

Unfiltered inleakage into the control room during Mode 1 and Mode 4 is fid4 inassumed to be 70 cfm Table 15.5 86 and (includes 10 cfm for inleakage due to ingress/egress inekaebased on the guidance provided in SRP 6.4).

For purposes of estimating the post-accident dose consequences, the control room is modeled as a single region. When in CRVS Mode 4, the Mode 1 intakes are isolated and outside air is a) drawn into the control room through the filtered emergency intakes; b) enters the control room as infiltration, c) enters the control room during operator egress/ingress, and d) enters the control room as unfiltered leakage via the emergency intake back draft dampers. The direction of flow uncertainty on the CRVS ventilation intake flowrates (normal as well as accident), are selected to maximize the dose consequence in the control room.

The dose consequence analyses for the LOCA, MSLB, SGTR and the CREA, assume a LOOP concurrent with reactor trip.

In addition, and as noted in Section 15.5.1.2, in accordance with current licensing basis the non-accident unit is assumed unaffected by the LOOP. Thus, to address the effect of a LOOP, and taking into consideration the fact that the time of receipt of the signal 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 second damper closure time.

b. Automatic isolation of the control room normal intake of the accident unit, and credit for CRVS Mode 4 operation is delayed by 38.2 seconds from receipt of the signal to switch to CRVS Mode 4. Thiis delay takes into account a) 28.2 seconds for the diesel generator to become fully operational including sequencing delays, and b) 10 seconds for the control room ventilation dampers to re-align. The 2 second SIS processing time occurs in parallel with diesel generator sequencing and is therefore not included as part of the delay. In addition, and as discussed earlier, the CRVS system design ensures that upon receipt of a signal to switch to Mode 4, the control room pressurization fans of the non-accident unit is initiated; thus fan ramp-up is assumed to occur well within the 38.2 seconds delay discussed above, unhampered by a LOOP.

15.5-36 15.5-36Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE analyses for the LRA and the LOL event assume that the The dose consequence in normal operation mode and do not credit CRVS Mode 4 control room remains operation.

design.

Table 15.5-81 lists key assumptions / parameters associated with control room The* 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 FAULTS 15.5.10.1 Acceptance Criteria The radiological consequences of accidents analyzed in Section 15.2 (or from other events 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 dose acceptance criteria of Regulatory Guide 1.183, July 2000 as outlined below:

EAB and LPZ Dose Criteria Regulatory 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 a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all dose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of the Regulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed be criteria tabulated in Table 6. Thus, the dose consequences at the EAB and LPZ will limited to the lowest value reported in Table 6, i.e., a small fraction (10%) of the limit imposed by 10 CFR 50.67.

(1) An individual located at any point on the boundary of the exclusion area for t-he-two,, hours. i,*mmediatel,, any 2-hour period following the onset of the postulated fission 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 thyroid from.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 population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the... hole, body,in excess.of 25 rem, or atota excess of 0.025 Sv (2.5 rem) TEDE.

15.5-37 15.5-37Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Control Room Dose Criteria of (3) Adequate radiation protection is provided to permit access and occupancy the control room under accident conditions without personnel receiving of the radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration accident.

15.5.10.2 Identification of Causes and Accident Description 15.5.10.2.1 Activity Release Pathways breach of any As reported in Section 15.2, Condition IIfaults are not expected to cause from the core or of the fission product barriers, thus preventing fission product release isotopes could be plant. Under some conditions, however, small amounts of radioactive of atmospheric released to the atmosphere following Condition II events as a result IIevents that are steam dumps required for plant cooldown. The particular Condition expected 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 system time relief valves The amount of steam released following these events depends on the remain open and the availability of condenser bypass cooling capacity.

bound all The mass of environmental steam releases for the Loss of Load Event Condition II events.

condition, in A LOL event is different from the Loss of Alternating Current (AC) power (e.g., reactor that offsite AC power remains available to support station auxiliaries being coolant pumps). The Loss of AC power condition results in the condenser steam releases from the SG unavailable and reactor cooldown being achieved using MSSVs and 10% ADVs until initiation of shutdown cooling.

all Condition II In-keeping with the concept of developing steam releases that bound to determine the events and encompass the LRA and CREA, the analysis performed of Loss of mass of steam released following a LOL event incorporates the assumption offsite power to the station auxiliaries.

with respect to Although Regulatory Guide 1.183 does not provide specific guidance from Condition scenarios to be assumed to determine radiological dose consequences 15.5-38 19 May 2010

15. 5-38Revision

DCPP UNITS 1 & 2 FSAR UPDATE II events, the scenario outlined below for the LOL analysis is based on the conservative assumptions outlined in Regulatory Guide 1.1 83 for the MSLB, and was analyzed to bound all Condition II events that result in environmental releases.

Table 15.5-9A lists the key assumptions / parameters utilized to develop the radiological consequences following a LOL event. The conservative assumptions utilized to assess the dose consequences ensure that it represents the Limiting Condition II event.

Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LOL event.

15.5.10.2.2 Activity Release Transport Model No melt or clad breach is postulated for the LOL (refer to Section 15.2.7). Thus, and in accordance with Regulatory Guide 1.183, Appendix E, item 2, the activity released is based on the maximum coolant activity allowed by the plant Technical Specifications, which focus on the noble gases and iodines. In accordance with Regulatory Guide 1.183, two scenarios are addressed, i.e., a) a pre-accident iodine spike and b) an accident-initiated iodine spike.

a. Pre-accident Iodine Spike - the initial primary coolant iodine activity is assumed to be 60 p#Ci/gm of DE 1-131 which is the transient Technical Specification limit for full power operation. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.

b. Accident-Initiated Iodine Spike - the initial primary coolant iodine activity is assumed to be at Technical Specification of 1 j..Ci/gm DE 1-131 (equilibrium Technical Specification limit for full power operation). Immediately following the accident the iodine appearance rate from the fuel to the primary coolant is assumed to increase to 500 times the equilibrium appearance rate corresponding to the 1 pCi/gm DE 1-131 coolant concentration. The duration of the assumed spike is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The initial primary coolant noble gas activity is assumed to be at Technical Specification levels.

The initial secondary coolant iodine activity is the Technical Specification limit of 0.1 1iCi/gm DE 1-131.

Plant Technical Specification limits primary to secondary steam generator (SG) tube leakage to 150 gpd per steam generator for a total of 600 gpd in all 4 SGs. To accommodate any potential accident induced, leakage, the LOL dose consequence analysis 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 STP conditions; total for all 4 SGs) is leaked into an effective SG. In accordance with Regulatory Guide 1.183, the pre-existing iodine activity in the secondary coolant and 15.5-39 15.5-39Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE iodine activity due to reactor coolant leakage into the 4 SGs is assumed to be homogeneously mixed in the bulk secondary coolant. The effect of SG tube uncovery in intact SGs (for SGTR and non-SGTR events) has been evaluated for potential impact on dose consequences as part of a WOG Program and demonstrated to be insignificant. Therefore, per Regulatory Guide 1.183, the iodines are released to the environment via the via the main steam safety valves (MSSVs) and 10% atmospheric dump valves (ADVs) in proportion to the steaming rate and the inverse of a partition coefficient of 100. The iodine releases from the SG are assumed to be 97% elemental and 3% organic. The noble gases are released freely to the environment without retention 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 the environment from the steam generators via the MSSVs / 10% ADVs. The SG releases continue for 10.73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br />, at which time shutdown cooling is initiated via operation of the Residual Heat Removal (RHR) system, and environmental releases are terminated.

15.5.10.2.30Offsite Dose Assessment AST methodology requires that the worst case dose to an individual located at any point on the boundary at the EAB, for any 2-hr period following the onset of the accident be reported as the EAB dose. For the LOL event, the worst two hour period can occur either during the 0-2 hr period when the noble gas release rate is the highest, or during the t=8.73 hr to 10.73 hr period when the iodine level in the SG liquid peaks (SG releases are terminated at T=10.73 hrs). Regardless of the starting point of the worst 2 hr window, the 0-2 hr EAB z/Q is utilized.

The bounding EAB and LPZ dose following a LOL event at either unit is presented in Table 15.5-9.

15.5.10.2.4 Control Room Dose Assessment_

The parameter values utilized for the control room in the accident dose transport model are discussed in Section 15.5.9. A summary of the critical assumptions associated with control room response and activity transport for the LOL event is provided below:

Control Room Ventilation The LOL event does not initiate any signal which could automatically start the control

-room pressurization air ventilation. Thus the dose consequence analysis for the LOL event assumes that the control room remains in normal operation mode.

Control Room Atmospheric Dispersion Factors 15.5-40 15.5-40Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE Due to the proximity of the MSSVs/1 0% ADVs to the control room normal intake of the affected unit, and because the releases from the MSSVs/IO% ADVs have a vertically upward discharge, it is expected that the concentrations near the normal operation control room intake of the affected unit (closest to the release point) will be insignificant. Therefore, only the unaffected unit's control room normal intake is assumed to be contaminated by releases from the MSSVs/10% ADVs (refer to Section 2.3.5.2.2 for detail).

The bounding atmospheric dispersion factors applicable to the radioactivity release points / control room receptors applicable to an LOL event at either unit are provided in Table 15.5-9B. The z/Q values presented in Table 15.5-98 take into consideration the various release points-receptors applicable to the LOL to identify the bounding z/Q values 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 the notes 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 in Table 15.5-9.

and the iodine .. oncentratio;n in the stea generator water..prior. to the accden. ;An*,,

of$thesem *^o keyt parameters; the rmeult~ a*re presented] in Figuvrme "15. 2 tlhroug",h 155.As hown^n on the figu.r~e, the potential thyroid doses aehigher w^ith inc-reaing stea releases and.. iodine concent.ratio~ns. Fiues 15... 5 2 and* 15.5 3 are result t+hatf asum R Guide A, R*,evision 1, assumptfions fo pest. acc-r-t'ident meerology.I..

- ]egulator/_,;

andbrathngrats DesgnBass seAsumpios) ,, s rles in,'Figure 15.5,2, shown T, a*pp roxmt ely"*h ! . 1 6 Ibm o~f steam is the ma,.ximm sta.rl.e..ete o afl cooldown.. without an... codese availability;H, and asemrlaeo prxmtl 4-x-1--* Ibm would result from. releasing only... the .. contents,-,of one' steam generator. due,,to*,-

Cond~ition II events. T~he hig~hest antiripafed doses would r~esult fro~m an event uha lOSS of elecricar-l load, andl the ptntian[l thyroid and w.hole body, doses fromtnfhis..

reeae to. the atmospher du*,ring, the..first 2. .hourS,ananditol1,300Ib Co*"nd~ition II eve.,nt steam*.' rel,'eaes.% The* assumptio-ns;used ,,,fo r meteorology, breat*--Jhing parra aphs.nn Nte*in that;, thet preceding, stemnn "Freleas e~e quantities are assocriatd ihr h original steam gnenrator {(OSG)' loss of loadJP (LOL a*%nalysis which providles the basis for, 15.5-41 15.5-41Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 1,023,000 Ibm, respectively) and are thercfore bounding since total dose is propo~ionaI to total steam release.

For the design basis case, it was assumed that the plant had been operating continuously with 1 percent fuel cladding defects and 1 gpm primary to secondary leakage. For the expected case calculation, operation at 0.2 percent defects and 20 gallons per day to the secondary was assumed. In both cases, leakage of water from primary to secondary was assumed to continue during cooldown at 75 percent of the pre accident rate during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at 50 percent of the pre accident rate during the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These values were derived from primary to secondary pressure differentials during cooldown.

It was also consen'at~vely assumed for both cases that the iodine padition factor in the steam generators releasing steam was 0.01, on a mass basis. In addition, to account for the effect of iodine spiking, fuel escape rate coefficients for iodincs of 30 times the normal operation values given in Table 11.1 8 were used for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the stan of the accident. Other detailed and less significant modeling ass umpt~ons are presented in Reference 1.

The resulting potential exposures from this type of accident are summarized in Table 15.5 9 and are consistent with the parametric analyses presented in Figures 15.5 2 through 15.5 5.

15.5.10.3 Conclusions It can be concluded from the results discussed that the occurrence of any of the events analyzed in Section 15.2 (or from other events involving insignificant core damage, but requiring atmospheric steam releases) will result in insignificant radiation exposures and are 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 indviua located at any point on the boundary of the exclusion area for the twe-hea-rsany 2-hour period kneiaeyfollowing the onset of the postulated fission product release is within 0.025 Sv (2.5 rem) TEDE o......;, g, ioarii,--, ......as shown in Table 15.5-9.

(2) The radiation dose to the w..hole, body4 .. nd to the, tkhyroid;, of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), is within 0.025 Sv (2.5 rem) e-..........

TEDE.. 4as shown in Table 15.5-9.

(2(3) The radiation dose to an individual in the control room for the duration of the accident is within 0.05 Sv (5 remn) TEDE as shown in Table 15.5-9.

15.5-42 15.5-42Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE LOCA 15.5.11 RADIOLOGICAL CONSEQUENCES OF A SMALL-BREAK 15.5.11.1 Acceptance Criteria (SBLOCA)

The radiological consequences of a small-break loss-of-coolant-accident the dose shall not exceed the dose limits of 10 CFR 400.&4 50.67, and will meet acceptance criteria of Regulatory Guide 1.183, July 2000 as outlined below:

However, Regulatory Guide 1.183 does not specifically address Condition III scenarios.

allows a per Regulatory Guide 1.183, Section 1.2.1, a full implementation of AST 50.67 in all dose licensee to utilize the dose acceptance criteria of 10 CFR 1.183 indicates consequence analyses. In addition, Section 4.4 of Regulatory Guide in Table 6 of that for events with a higher probability of occurrence than those listed should not exceed the Regulatory Guide 1.183, the postulated EAB and LPZ doses and LPZ will be criteria tabulated in Table 6. Thus, the dose consequences at the EAB (10%) of the limit limited to the lowest value reported in Table 6, i.e., a small fraction imposed by 10 CFR 50.67.

EAB and LPZ Dose Criteria for any (1) An individual located at any point on the boundary of the exclusion area product release shall 2-hour period following the onset of the postulated fission 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 population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose -in excess of 0.025 Sv (2.5 rem) TEDE.

product,,. release .hall not, recei. total* radiaton, dose to, the... hole body in a,=

fromiodie exp.ure theouter*0boundar ofhe ÷ ..

low<*,{- .,

! n indiviodua locate at+an point.on...

15.5-43 15.5-43Revision 19 May 2010

DCPP UNITS 1 & 2FSAR UPDATE body or its equivalent toayp~o h oy (i 30,,rem thyri and* beta

'e.,

skin, Reference 51) fctI-~H~ uuwuun ..... ut.... w .... ...

15.5.11.2- Identification of Causes and Accident Description As discussed in Section 15.3.1, a SBLOCA (defined in UFSAR Chapter 15.3.1 as a break that is large enough to actuate the emergency core cooling system), is not expected to cause fuel cladding failure. For this reason, the only activity release to the containment will be the dissolved noble gases and iodine in the reactor coolant water expelled from the pipe rupture. Some of this activity could be released to the containment atmosphere as the water flashes, and some of this amount could leak from the 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 CREA since the CREA is postulated to result in 10% fuel damage, whereas the SBLOCA has no fuel damage.

As demonstrated in Table 15.5-52, the dose consequences at the EAB and LPZ following a CREA is within the acceptance criteria applicable to the SBLOCA.

The.de.iled. escr.pion. o the... .model..... used.i calculating the potential* e..........pos*ures from Sectio,-n 15.5.1"7 o-f this. o... The specifi assumption.. used..., in the analysi r....aS-15.5.7, resectvel. Othe"r*÷common.assumptions .. re. described in the previou sections';'*r, of 15.5.

(2) It haso been assme... that*all of*the.ater cont..aie in the* RCS is released to th containment. For the, desig basi ce the reactor colant percent deeciv claddng"were..... use., These.. acivities and concentrations i

• I II I I *1 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,' coolant 100 percent is assumed to be released to the containment atmosphere at the time of the accident. For the iodines, it is assumed that only 10 percent of the dissolved iodine in the coolant is released to the containment atmosphere, due to tho solubility of the iodine. It is assumed that the 15.5-44 Revision 19 May2010

DCPP UNITS 1 & 2 FSAR UPDATE amount. of" "odine;' in chemical fom t..hat are not ffected by the.. p..y

.. , from the.,.

released..'" fue,,'l, up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, after,, -""4*to."be s assumed

• ," the accident,. i.";4" released to t he containment.. Of the amount... of,,

noble .. ases released to*-

10 percen of;the. iodine released....' tothe onainen are. reeae to the Section 15.5.17.

(6) The containment lea.ag.rate n..this anlyisaralso assume..d to be the s.me.as for thearg break F~ -,LOC (A and4 arc discus..ed i n Section 15.5.1"7.

The resultin potentira*;l e..posures arc lIsted, in Table "15.5,10 and demonstra,-te that* all ca-lculatedl doses are well1 below the guide*line alues,, sec*ified4 in, 10C 011 . SD ince*

the activity- relases.. from this typ of..evn wil. ,, be significantly,,h low:er than thoe,, f.-rom large breakL-OA, any cntnrol room evxposure which might occu*r would1, be we~ll within 15.5.11.3 Conclusions Thea 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'idual located a"*;*thany point *'  ; boud the on CF of 0th exclusion forth T-area !0hur immediately;*÷;* followin**t he Wonseto the,,4 postu~lated"*h fhrission producti relesar

\Aelwtinteds CFRof imits 10 as 100.11 shown in Table 15.5 10.

(2)The kradatIo doseA to the1 whole body, andn to~ thyroid o an indivIdua rthea h Onthreleases (durhing thnerentire operiodof itsprassge, areiwenlludthd thatte dose 15.5-45 Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE consequences at the EAB and LPZ following a SBLOCA will remain within the acceptance criteria listed in Section 15.5.11.1.

15.5.12 RADIOLOGICAL CONSEQUENCES OF MINOR SECONDARY SYSTEM PIPE BREAKS 15.5.12.1- Acceptance Criteria The radiological consequences of accidents analyzed in Section 15.3 such as minor secondary system pipe breaks shall not exceed the dose limits of 10 CFR 100.11! as oulnd cox10 CFR 50.67, and will meet the dose acceptance 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 a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all dose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of Regulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. Thus, the dose consequences at the EAB and 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.

EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.

(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.

15.5-46 Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE An individual located at n, point o'n the boundalr, of the. exc,'lusio;n -are-for the *t*o An, indiidu4al located- at÷ any point~ on the o,,ter bounda, of the low, population -zone,w*ho wV receive a tota! radiation dose to the period of its passage), shall not (during the entire dose in excess of 200 rem to the in excess of 25 rem, or a total radiation whole body thyroid from iodine exposure.

15.5.12.2- Identification of Causes and Accident Description system caused by The effects on the core of sudden depressurization of the secondary were described in an accidental opening of a steam dump, relief or safety valve pipe breaks. As Section 15.2 and apply also to the case of minor secondary system to occur. In shown in that analysis, no core damage or fuel rod failure is expected on the core of a major Section 15.51-_84.2, analyses are presented that show the effects expected to occur.

steam line break, and, in this case also, no fuel rod failures are from nucleate The analyses presented in Section 15.3.2 demonstrate that a departure occur anywhere in the boiling ratio (ON BR) of less than the safety analysis limit will not core in the event of a minor secondary system pipe rupture.

to be The steam releases following a minor secondary line break is expected significantly less than that associated with a main steam line break.

and LPZ As demonstrated in Table 1 5.5-34, the dose consequences at the EAB criteria applicable to the minor secondary line following a MSLB is within the acceptance break.

consequences. of this event, due to the releas of so..e stea Th4,e-p si ble radiological,..*

that the dose On the basis of this conservative comparison approach, it is concluded pipe rupture will consequences at the EAB and LPZ following a minor secondary system remain within the acceptance criteria listed in Section 15.5.12.1.

15.5-47Revision 19 May 2010 15.5-47

DCPP UNITS 1 & 2 FSAR UPDATE O'n the. ba-'sis of the discus..ed result..*,- it can be, concluded that th, potetia e .po.ure following.a.mi.or...co.d.......tem pie, pur ol be...n...gnificant.3 * ....

The, radi4ation,, dose to the whole,* bod,, and, to the thyoi of.an4individual;4"' loc"ted at an..

rdAtincoeptoathe wholerbdiandt h hri fa niiullctda n The.3.

pimlmntontedoutrinboudr; lofdn. theloounltion zoeen whot is exosedin tror thers radiactises cludportnesutiongro the3. cofirm pstuated prtoductntreleas(duing theailoia eontieqperiooeis passage),ra areisigu oflasin fcat erhownsi 15.5.13.1 Acepticatnce f CrtiaussadAcdn ecito Fuel assembly loading errors suhall benprventedtby ladministatie prmoedfurlasmles inoimplemenedduringiore, loading. In thel rounlikel eventutacturaitloadngerror mocreplts, analyses supongertingSchtion 15ad3n3 safll cofirm tatsmlnoevntraing mauatouradwiohplogica onstequrngenrcshallnoccurlaseadreutof lcrasdin herroruxs. i h ro eslsi cn fuel andcore erors such aflsse loitoncading inadvertentl loading oneofoefelasmle niheth one or more fuel assemblies requiring burnable poison rods into a new core without of burnable poison rods is also included among possible core loading errors. Because margins present, as discussed in detail in Section 15.3.3, no events leading to radiological consequences are expected as a result of loading errors.

15.5.13.3 Conclusions Because of margins present, as discussed in detail in Section 15.3.3, no events leading to radiological consequences are expected as a result of loading errors.

15.5-48 15.5-48Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.14 RADIOLOGICAL CONSEQUENCES OF COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.5.14.1 Acceptance Criteria The radiological consequences of small amounts of radioactive isotopes that could be released to the atmosphere as a result of atmospheric steam dumping required for plant cooldown following a complete loss of forced reactor coolant flow shall not exceed the dose limits of 10 CFR 100.11 a s outlincd beowo::50.67, and will meet the dose acceptance 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 a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all dose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of Regulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. Thus, the dose consequences at the EA8 and 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.

EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.

(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.

,An individu`,,l located*, at,an... pint on the, bounda`4 of.the exclsio area....for,the. h, to the who~le body in excess#of 25 rem,. or a tota~l radiatio~n dosea in exces of 1300 remn tnytroia tor~n iodilne exposuc-ire.

15.5.14.2 Identification of Causes and Accident Description As discussed in Section 15.3.4, a complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps 15.5-49 15.5-49Revision 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 of loss 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 the complete loss of forced reactor coolant flow, the DNBR does not decrease below the safety analysis limit during the transient, and thus there is no cladding damage or release of fission products to the RCS. For this reason,, this accidcnt has nqo significant r*4 -ef-,-,h-* ts,- 7 The possible radiological consequence of a complete loss of forced reactor coolant flow is expected to be bounded by the conservative Loss-of-Load scenario with a coincident Loss of offsite power described in Section 15.5.10.

As demonstrated in Table 15.5-9, the dose consequences at the EAB and LPZ following a Loss of Load is within the acceptance criteria applicable to the complete loss of forced reactor coolant flow.

15.5.1 4.3 Conclusions On the basis of this comparison approach, it is concluded that the dlose consequences at the EAB and LPZ following a complete loss of forced reactor coolant flow will remain within the acceptance criteria listed in Section 15.5.14.1 .- nay* e demntrate thatnrther fnicnt;,rl,,* enirnmenta!*

arenosinif ,a described finn Sletion 1h5.3.h1l efeth s o~f nthe CompeteLoss~f ofi Forced,*

n rloeator* Colnt lweei nt.f Tereoreth po*int on the boundar' excluso ofllt,*tt* he area lfor the *ohours immediately folown radi~oac ftiv cloudr resulting from the pnostulated fission producl~t relcase (during the 15.5-50 15.5-50Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.15 RADIOLOGICAL CONSEQUENCES OF AN UNDERFREQUENCY ACCIDENT 15.5.1 5.1 Acceptance Criteria The radiological consequences of small amounts of radioactive isotopes that could be released to the atmosphere as a result of atmospheric steam dumping required for plant cooldown following an underfrequency accident shall not exceed the dose limits of 10 CFR 100.11!as outlined below: 50.67, and will meet the dose acceptance criteria of Regulatory 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 a licensee to utilize the dose acceptance criteria of 10 CFR 50.67 in all dose consequence analyses. In addition, Section 4.4 of Regulatory Guide 1.183 indicates that for events with a higher probability of occurrence than those listed in Table 6 of Regulatory Guide 1.183, the postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6. Thus, the dose consequences at the EAB and 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.

EAB and LPZ Dose Criteria (1) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release shall not receive a radiation dose in excess of 0.025 Sv (2.5 remn) TEDE.

(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDEF.

15.5-51 19 May 2010

15. -5 1Revision

DCPP UNITS 1 & 2 FSAR UPDATE A .- " I=* ,, I - I I - -I- l

• f II I

............... ........atan point o.n tne, couter; o'"r, .... exclus. are. whotle whlebdyineces or a total' radiation, dose in excess of 300 remn to the.

.f25rm,-*

,.J 15.5.15.2- Identification of Causes and Accident Description A transient analysis for this unlikely event has been carried o'-tis discussed in Section 15.3.4. The analysis demonstrates that for an underfrequency accident, the DNBR does not decrease below the safety analysis limit during the transient, and thus there is no cladding damage or release of fission products to the RCS. However, small amounts of radioactive isotopes could be released to the atmosphere as a result of atmospheric steam dumping required for plant cooldown.

The possible radiological consequence of this event is expected to be bounded by the conservative Loss-of-Load scenario with a coincident Loss of offsite power described in Section 15.5.10.

As demonstrated in Table 15.5-9, the dose consequences at the EAB and LPZ following a Loss of Load is within the acceptance criteria applicable to an underfrequency accident.

A drt~ilred dkr*i,,.£ien nf the~nnfrnti2I ren,-irnnmrentn! cnen*,,ienne'* of *ncecde~ntp...........----..--.

ini,,-Ih,;nn ,afrnnc'*nhari- c{,'nm rm,,r,-nni-r

....... ;Ic. nracra-far ;rn Qar',finr 4*g fl I kaE',,'

h...--

exposures. it can, be concludeda tha,f alt~houg..h very.. ulikely, the occurrence.. of this 15.5.1 5.3 Conclusions On the basis of this comparison approach, it is concluded that the dose consequences at the EAB and LPZ following a complete loss of forced reactor coolant flow will remain within the acceptance criteria listed in Section 15.5.15.1. Onv, the,,, b,,*,asis of"the*,, potential'*;*

.f .. .. -) ..AL A. . tI . .t . . . . . .~ .l .I . . . II. . i*.. J" I..

t*1 IVlIVV

.

• I gl

....... .... ............ ..... * ...... F ......

Additionally, the analysis demonstrates that the acceotance criteria arc met as follows:

15.5-52 15.5-52Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE The radiation do"se to the ,whole,body*, and to, the* thyri ,- of.an ndividual;*,. locatod,, at ..n.

The radiatonoia donseqtoethes bodoad hofasnle tousthethroi cofnto indieiduly woctedrata san" ponot onee the douter boindas of the low popu !atio zo'-ne, ..ho w.50.67, adto theeth l rdoseaccetine cldreiuteing fom thegpostulateude fision prouct re00eased(durlingd thelentir peruodtofy Gitspsae,1.83res inotsign ificant a y ddhrniTbes 15ndti5 1I. naisHwvr 15..1 ReuaDIoLOGuiCAL CONSEUENCSecn121 OF Aul SiNGlemettoRoD CLSTERlos 15.5.16e1 C cceptance Aouiieteds riteria of1CF506inalds Th ailgclconsequencesaaye adtofna Seciongl r.4od clsergcontrolyassdembly83withdrawaleshl ntha oexedt wthe dos limitsrbait10cCRu10.11ce of otlin toelsed nd willmet thef ieo:0 dsacetnecieiofRegulatory Guide 1.183, LZosshoutlinotecedbhelw Julyostl2000EA and criegi aulatoed Gide 1.13bde nth hs, dspeiial addrseqecs Codtio Ih scnrosboevr A ndwll piier Rgltor oesaui Gh 1.83reotedion 1.2.1 e ful

, ae, AS smplementaction of0% lows ah ii icesetouiizph osedacetncbrieiao 10 CFR 50.67.i llds that foAvnt wniithual highter probabilpinty ofncurec ondayo thoe exlstdion Tablea 6foan tha Reglaoury Guriode 1.183,heoneto EAfnLZdssshoudnpodut rexceaed thel the postulated crtrioabltredeine Tabl 6.dithus, thsedoenoneqence o 05Sv at. them TEABaDELZwilb (1) An individual located at any point on the bouteondary of the exluson araporulatiny zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.025 Sv (2.5 rem) TEDE.

15.5-53 15.5-53Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE An individual located at any point on the bounda~ of thc exclusion area for the ~o hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

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 the whole body. in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

15.5.16.2- Identification of Causes and Accident Description A complete transient analysis of this accident is presented in Section 15.3.5. For the condition of one rod cluster control assembly (RCCA) fully withdrawn with the rest of the bank fully inserted, at full power, an upper bound of the number of fuel rods experiencing DNBR less than the safety analysis limit is 5 percent of the total fuel rods in the core.

The possible radiological consequence of this event is expected to be bounded by the CREA discussed in Section 15.5.23.

The dose consequences following a single rod cluster control assembly withdrawal will be 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 rest of 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 LPZ rod following a CREA is within the acceptance criteria applicable to the condition of one cluster control assembly fully withdrawn with the rest of the bank fully inserted, at full power. A detailed*' discussion*' of the* potential radiological" conse..uence. of accidents 15.5-54 15.5-54Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.16.3 Conclusions On the basis of this comparison approach, it is concluded that the dose consequences at the EAB and LPZ following the condition of one rod cluster control assembly fully withdrawn with the rest of the bank fully inserted, at full power will remain within the acceptance criteria listed in Section 15.5.16.1.

the potent hia

,6fo exposure dicun t dh h,,it can be concuded,,. that++ the ...

On he basi+s*,

ofthis

- a., cidenrt,

., undue would not,,Ic.au,,e #,-ri+kto the healh,-.,-and,*r**i*l+ afety,of,-,th.

occurrence,,

!5.5 !2.

I ana.7R LysIsA CONSEUENCE tht Fh cAJ are met RiRUTeRia MAsRY theAN accetanc of a large break loss of coolant +

15.5.17.1 Acceptance Criteria The radiological consequencesacceptance of a LOCA shall not exceed the dose limits of 10 CFR 50.67, and will meet the dose criteria of Regulatory Guide 1.183, July 2000 and outlined below:

EAB and LPZ Dose Criteria (1) An individual located at any the point on the boundary of the exclusion area for any 2-hour period following onset of the postulated fission product release shall 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 population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.25 Sv (25 rem) TEDE.

15.5-55 15.5-55Revision 19 May 2010

DCPP UNITSI1 & 2 FSAR UPDATE Control Room Dose Criteria Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) TEDE for the duration of the accident.

Technical Support Center Dose Criteria The 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 CFR 50.67. The dose to an operator in the TSC should not exceed 5 rem TEDE for the duration of the accident.

(1) Thc rad4;-iological con..equence. of a major... ruptur of primry coolan.t,* pipe contanmen tn post LOCA' rec ...Ircultin LoopInL leakg in the Auviliary BuIdingl (inclushiveiL* of.a residual heatJ reoam,-l (RHR) pu~mp seaol failu~re resulti+ng in a 50 gp~m leak, for 30mnutes.,+* 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~ls fissio p.. ,roducrele..se.shall not receiv tota, radiatio;n dl'ose to the a,=

30 rem to the thyro'idflro~m iodnelh exposre.*

ii. An1 individulH O l/'+*oate at any..point{ o+nthe oute fr bound.r ol...f the low ppulationm-+ .. one, who is expose.d to the, radioactive.+ cloud resulting. r,-

o~f its pasage,=-r, shall notf receive*+ ar totl radiat*ion dos to*0trhe ,,ho..le.

body4, in exces.. of 25 rem, or a total rad*,iation drose n excess. o-f 300 co*ntrol room operator.. ,

under faccien conditions shall not be0 in excess.of 5 rem ,wholebodyl* or,, its equivah'.lent to+any part of the body; (i;e,* 30' rem thyri 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 post capabnhi;lt fo+r hydroge*n co~ntrol to the hydroge~n frecombiners,r an indivdul,,*

loatednf at anyl point on the bou"llil4Jndar of'the exclusoin a*rea whnois exosed 15.5-56 15.5-56Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 0.5 rem/year in accordance with 10 CFR Part 20.

15.5.17.2 Identification of Causes and Accident Description II 15.5.17.2.1 I~asc Ientsandkeicec - ract on*ACtlVlty Release Pathways The accidental rupture of a main coolant pipe is the event assumed to initiate a -L--large break LOCA. Analyses of the response of the reactor system, including the emergency core cooling system (ECCS), to ruptures of various sizes have been presented in Sections 15.3.1 and 15.4.1. As demonstrated in these analyses, the ECCS, using emergency power, is designed to keep cladding temperatures well below melting and to limit zirconium-water reactions to an insignificant level. As a result of the increase in cladding temperature and the rapid depressurization of the core, however, some cladding failure may occur in the hottest regions of the core. Following the cladding failure, some activity would be released to the primary coolant and subsequently to the inside of the containment building. Active mechanisms include radioactive particulate and iodine removal by the containment sprays inclusive of the containment air mixing provided by the CFCUs. Section 6.2 describes the design and operation of the CSS and the CFCUs. wiu Because.. of the.....

~dJiyiimm prsuization, iuuu~ of the cotimn.ulin..te.UII

• urii...ui v,,iuui, iu,,, *,

f;.,,., ,,+,.,,,-,t.,m ,II; mId mci insi,* ",,d ,.,i Iikir.,+cIi Ai _ri m t;hc i.+ k:U~l,,m-+ U +h-i . rsf ,-~,r ," t.,;n ii-.,s px,z...rm.,ntc. r'on,1,i,'.f h"_ th,= .VL* .t.~. Nt*J~fjinnts tsIqh.*rgcr-,fnr Thci frn.-,r.ticr ~.f th.-' frf.*"4

• • i ÷F*f i*.. *-P.I, less, since the rate of thermal radiolytic decomDosition would exceed the rate of Organic compounds of iodine can be formed by reaction of absorbed elemental iodine on su~aces of the containment vessel. Experiments have shown that the rate of formation 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 of the enclosure, whether the process L limited to diffusion to the su~aoc or by the reaction rate of the absorbed iodine. The obsered yields of organic iodine as a function of aging time in various test enclosures, with various volume to su~ace area ratios, 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 of the atmospheric iodine per hour.

The potential exposures following the postulated sequence hn~~r~ of events in LBLOCAs have In +P~~ '~"n'~trd n'~'~ it ho" g'~"ii'~'r'd th-~d thci p~tiri-~

~tcicip .,~.,.,k,-,,..,4 fe-~~ h~in r1c~~~~-

15.5-57 15.5-57Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE cons...a... s;,nei,:.the E-COS is designed to pre...ent gross cladding damage. In zeo.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+. islarge.a shown in Reference 10.

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DCPP UNITS 1 & 2 FSAR UPDATE pump seal failure resulting in a 'filtered" release is DCPP's licensing basis with respect to passive single failure.

-Section 3.1.1.1 (Single Failure Criteria / Definitions), Item 2; discusses passive failures - "The structural failure of a static component that limits the component's effectiveness in carrying out its design function. When applied to a fluid system, this means a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gpm for 30 minutes. Such leak rates are assumed for RHR pump seal failure."

-UFSAR Appendix 6.3A.3.2 (discusses passive failures), indicates that - the design of the auxiliary building and related equipment is based on handling of leaks up to a maximum of 50 gpm. Means are provided to detect and isolate such leaks in the emergency core cooling pathway within 30 mains. A review of the equipment in the RHR system loop and the 0S8 loop indicates that the largest leakage would result from 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, shows that 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 than an RHR pump seal failure leak.

-UFSAR Section 15.5.17.2.8, indicates that - failure of an RHR pump seal at 24 hrs is assumed as the single failure that can be tolerated without loss of the required functioning of the RHR system.

Therefore, the RHR Pump Seal Failure is retained as a release pathway for the AST dose consequence analysis.

5. Releases to the environment from the Miscellaneous Equipment Drain Tank (MEDT) which collects component leakage hard-piped to the MEDT. The collected-fluid includes both post-LOCA sump water and other non-radioactive fluid.

6. Releases to the environment via the refueling water storage tank (RWST) vent due to post-LOCA sump fluid back-leakage into the RWST via the mini-flow recirculation lines connecting the high head and low head safety injection pump discharge piping to the RWST.

The LOCA dose consequence analysis follows the requirements provided in the pertinent sections of Regulatory Guide 1.183 including Appendix A. Table 15.5-23A lists the key as~sumptions / parameters utilized to develop the radiological consequences following a LOCA at either unit.

Computer code RADTRAD 3.03, is used to calculate the control room and site boundary dose due to airborne radioactivity releases following a LOCA.

15.5-59 15.5-59Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE 15.5.17.2.2 Activity Release Transport Moe..pra .... som* ,,din Remo,.al Ra* es T**"he... conaincn ..pr.y syte (CSS) is, desied;.... in detail.. ong wi... "tha 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 the AILtho'-Ih a s'ubsen'-ent ssfe÷,- eX'-l!'Iat!Ol s"hr;-ed that the De~sign Case enefficient nf

.. . ... J - - - - - .... - - ... --.. .... .. - - - - . .--.. .. . .. .

,-pm., spray.header flow) should, be,-reduce to. appro....imately, 31 per"hour/(f'"or260 15.5.17.2.2.1 Containment Pressure /Vacuum Relief Line Release In accordance with Regulatory Guide 1.183, Appendix A, Section 3.8, for containments such as DCPP that are routinely purged during normal operations, the dose consequence analysis must assume that 100% of the radionuclide inventory in the

• primary coolant is released to the containment at the initiation of the.LOCA. The inventory of the release from containment should be based on Technical Specifications primary coolant equilibrium activity (refer to Table 15.5-78). Iodine spikes need not be considered.

Thus, in accordance with the above guidance, the 12 inch containment vacuum / over pressure relief valves are assumed to be open to the extent allowed by Technical Specifications (i.e., blocked to prevent opening beyond 50 degrees), at the initiation of the 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= 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. It is conservatively assumed that 40% of release flashes and is instantaneously and homogeneously mixed in the containment atmosphere and that the activity associated with the volatiles, i.e., 100% of the noble gases and 40% of the iodine in the reactor coolant is available for release to the environment via this pathway.

Containment pressurization (due to the RCS mass and energy release), combined with the relief line cross-sectional area, results in a 218 acts release of containment air to the 15.5-60 15.5-60Revision 19 May 2010

DCPP UNITS 1 & 2 FSAR UPDATE environment for a conservatively estimated period of 13 seconds. Credit is taken for pressure boundary integrity of the containment pressure / vacuum relief system ductwork 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 the gap phase release, releases associated with fuel damage are not postulated. The chemical form of the iodine released from the RCS to the environment is assumed to be 97% elemental and 3% organic.

15.5.17.2.2.2 Containment Leakage The inventory of fission products in the reactor core available for release into the containment following a LOCA is provided in Table 15.5-77 which represents a conservative equilibrium reactor core inventory of the dose significant isotopes, assuming maximum full power operation at 1 .05 times the current licensed thermal power, and taking into consideration fuel enrichment and burnup. The notes provided at the bottom of Table 15.5-77 provide information on isotopes used to estimate the inhalation and submersion doses following a LOCA, vs isotopes that are considered to estimate the post-LOCA direct shine dose.

Per Regulatory Guide 1 .183, the fission products released from the fuel are assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment 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 accident and continues for 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (i.e., t=1.8 hrs).

b. The core inventory release fractions, by radionuclide groups, for the gap and early in-vessel damage are as follows:

Early In-Vessel Group Gap Release Phase Release Phase Noble gas 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium Group -0.05 Ba, Sr -0.02 Noble Metals ____________0.0025 Cerium Group 0.0005 Lanthan ides 0.0002 15.5-61 15.5-61Revision 19 May 2010