ML15176A529

License Amendment Request 15-03 Application of Alternative Source Term - Updated Final Safety Analysis Report Markup. Part 3 of 8

ML15176A529
Person / Time
Site:	Diablo Canyon
Issue date:	06/17/2015
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation
References
DCL-15-069
Download: ML15176A529 (70)

Text

Enclosure Attachment 8 PG&E Letter DCL-1 5-069 ATTACHMENT 8 Diablo Canyon Power Plant Updated Final Safety Analysis Report Markup (For Information Only)

DCPP UNITS 1 & 2 FSAR 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 C.1.c of Regulatory Guide 1.111, Revision 1. This information is used as input to develop the design basis radiological analysis X/Q 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, C.2.1.2 and C.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 Revision 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 TMI Action Plan Requirements Item III.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 II1.A.2.2

-Meteorological Data: NUREG-0737, Supplement 1, January 1983 provides the requirements for II1.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 characteristics of Diablo Canyon Power Plant (DCPP) site consists of climatological summaries, technical studies, and reports by Dye (Reference 2), Edinger (Reference 3), Elford (Reference 4), 2.3-3 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE description of computer program EN-1 13) 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-1 13 calculates z/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 X/Q's are as follows: X/Qi = {(u)[(7t)(a'y)(@Tz)

+ (A12)]}1 (2.3-7)X/Q2 = [(u)(3n)(ay)(z)]-1 (2.3-8)X/Q3 = [(u)(/)(-)(cz)]

1 (2.3-9)where: x/Q = relative concentration (sec/i 3);ay, a, = horizontal and vertical dispersion coefficients, respectively, based on stability class and horizontal downwind distance (m);u = wind speed at the 10-meter elevation (m/sec);A = cross-sectional building area (M);_.y = (M)(ay) for distances of 800 meters or less; and y = [(M-11)(y800m)

+ cry] 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/Qi and X/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-1 13, 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 2.3-2, taking into consideration a 45-degree azimuth sector centered on each 22.5 0-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 (refer to Figure 2.3-5).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 X/Q calculations is identified in 2.3-29 Revision 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 X/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 ,and -no Rfadioactive decay or plume depletion due to deposition is not considered.

The highest EAB and LPZ X/Q values from among all 22.5 0-downwind sectors for each release/receptor combination and accident period are summarized in Table 2.3-145.EAB X/Q values are presented for releases from Unit 1 and Unit 2, while the LPZ X/Q values are applicable to both units. The 0.5% sector dependent X/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 X/Q values 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 Revision 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 (refer to Figure 2.3-5).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 x/Q calculations:

• The plume centerline from each release is transported directly over the control room or technical support center air intake/receptor (conservative);

• The 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 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE* All 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 Z/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 X/Q can be considered an average value for in-leakage locations around the control room envelope.

The X/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 X/Q from the accident release point to the center of the TSC at its roof level is utilized for TSC in-leakage since the above x/Q 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 X/Q 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 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 ARCON96 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 C.3.4 of Regulatory Guide 1.194, June 2003. Thus no X/Qs are developed for the above release point/receptor combinations.

Ener-getic Releases 2.3-32 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The vertical velocity of the MSSV and 10% ADVs releases is at least 95 times larger than the 95th percentile wind speed of 1 m/sec and approximately 5 times larger than the 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/10% 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 I m/sec due to the large vertical velocities of the releases.

Thus, for purposes of estimating dose consequences, it is appropriate to use the Z/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 x/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 m/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 X/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 x/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 X/Q, 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 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE degrees azimuth with the use of ARCON96, as described in Regulatory Position C.3.3.2 of Regulatory Guide 1.194, June 2003.Redundant Radiation Monitors Per Regulatory Guide 1.194, June 2003, Regulatory Position C.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 can be reduced by a factor of 4 and utilized to estimate the dose consequences.

PG&E Design Class II Lines Connecting to PG&E Design Class 1 Plant Vent The 16 inch PG&E Design Class II gland 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 /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 X/Q values 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 2.3-146 provides the release point / receptor combinations that were evaluated.

Tables 2.3-147 and 2.3-148 provide the control room X/Q values for the individual release point-receptor combinations for Unit 1 and Unit 2, respectively.

The X/Q 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 2.3-147 and 2.3-148.2.3-34 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Table 2.3-149 presents the X/Q values for the individual post-LOCA release point TSC receptor combinations for Unit 1 and Unit 2 applicable to the TSC normal intake and the center of the TSC boundary at roof level (considered an average value for potential TSC unfiltered in-leakage locations around the envelope).

The Unit 1 and Unit 2 control room pressurization air intakes also serve the TSC during the emergency mode. Thus, the X/Qs presented in Tables 2.3-147 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.2.3.6 LONG-TERM (ROUTINE)

DIFFUSION ESTIMATES HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.2.3.6.1 Objective Annual relative concentrations (X/Q) were estimated for distances out to 80 kilometers from onsite meteorological data for the period May 1973 through April 1975. These relative concentrations are presented in Table 2.3-2; they were estimated using the models described in Reference

18. The same program also produces cumulative frequency distributions for selected averaging periods using overlapping means having hourly updates. For critical offsite locations, measured lateral standard deviations of wind direction, o-A, and bulk Richardson number, Ri, were used as the stability parameters in the computations.

The meteorological input data were measured at the 10 meter level of the meteorological tower at DCPP site. Annual averaged relative concentrations calculated by the above methods are presented in Table 2.3-4.2.3.6.2 Calculations The meteorological instrumentation that was used to obtain the input data for the previously discussed relative concentration calculations at DCPP site is described in Section 2.3.4. Procedures for obtaining annual averaged relative concentrations are described in detail in Reference 15.2.3.6.3 Meteorological Parameters The following assumptions were used in developing the meteorological input parameters required in the dispersion model: (1) There is no wind direction change with height (2) Wind speed changes with height can be estimated by a power law function where the exponent, P, varies with stability class and is assigned the following values: Pasquill Stability Class Exponent (P)2.3-35 Revision 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 pFeepeFaiea-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 control room provide information for use in determining the magnitude of the release of 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 Releases from Light-Water-Cooled Reactors The pre-operational values of dilution factor and deposition factor used in the calculation of annual average offsite radiation dose are discussed in Section 11.3.7. The values of 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 calculate the intermediate averaging time X/Q values at the LPZ for the periods of 2-8 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/Q 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 Revision 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 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, 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 III.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 II1.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--l1 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 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 24. ANSI/ANS 2.5, American National Standard for Determining Meteorological 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.2.3-42 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-145 EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE ATMOSPHERIC DISPERSION FACTORS%IQ (sec/m 3)Receptor 0 -2 hours 2 -8 hours 8 -24 hours 1 -4 days 4 -30 days Unit 1 EAB (NW) 2.50E-04 -Unit 2 EAB (SSE) 2.30E-04 Unit 1/2 LPZ NW) 2.12E-05 9.26E-06 6.26E-06 2.67E-06 7.86E-07 Notes: 1: An EAB x/Q value of 2.5E-04 sec/m 3 is used for radiological dose calculations from all release points.2. The 0.5% sector dependent x/Q values are presented with the worst case downwind sector indicated in parentheses.

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-146 ON-SITE ATMOSPHERIC DISPERSION FACTOR EVALUATION POST-ACCIDENT RELEASE POINT / RECEPTOR COMBINATIONS 1.2.3.4.5.6.7.8.9.10.11.12.13.14.15.16.17.18.19.20.Release Points Unit I Containment Building Edge Unit 2 Containment Building Edge Unit I Plant Vent Unit 2 Plant Vent Unit 1 Refueling Water Storage Tank (RWST) Vent 1 Unit 2 RWST Vent 1 Unit 1 Containment Penetration (GE Area)Unit 2 Containment Penetration (GE Area)Unit I Containment Penetration (GW/FW Area)Unit 2 Containment Penetration (GW/FW Area)Unit I Fuel Handling Building Unit 2 Fuel Handling Building Unit 1 Equipment Hatch Unit 2 Equipment Hatch Unit 1 Main Steam Safety Valves (MSSVs)Unit 2 MSSVs Unit 1 10% Atmospheric Dump Valves Unit 1 10% Atmospheric Dump Valves Unit 1 Main Steam Line Break Location Unit 2 Main Steam Line Break Location 1.2.3.4.5.6.7.On-Site Receptors Unit 1 Control Room Normal Intake Unit 2 Control Room Normal Intake Unit 1 Control Room Emergency Intake Unit 2 Control Room Emergency Intake Control Room Center (i.e., In-leakage)

TSC Normal Intake 2 TSC Center2(i.e., In-leakage)

Notes: 1. X/Q values for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent.2. X/Q values developed only for the LOCA (i.e., release points 1 through 10).

I DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-146A ON-SITE ATMOSPHERIC DISPERSION FACTOR EVALUATION POST-ACCIDENT RELEASE POINT & RECEPTOR LOCATION ID 1 ReleaselRecepto Description (See Figure 2.3-5) r Note 2 Release Point Unit 1 Containment Building (CB) edge Note 2 Release Point Unit 2 Containment Building (CB) edge 1 Release Point U1 Plant Vent 2 Release Point U2 Plant Vent 3 Receptor U1 Control Room Normal Intake 4 Receptor U2 Control Room Normal Intake 5 Receptor U1 Control Room Emergency Intake 6 Receptor U2 Control Room Emergency Intake 7 Release Point U1 RWST Vent 8 Release Point U2 RWST Vent 9 Receptor Control Room Center (location assigned for unfiltered inleakage) 10 Release Point Unit 1 Containment Penetration Area, GE 11 Release Point Unit 2 Containment Penetration Area, GE 12 Release Point Unit 1 Containment Penetration Area, FW/GW 13 Release Point Unit 2 Containment Penetration Area, FW/GW 14 Release Point U1 Fuel Handling Building 15 Release Point U2 Fuel Handling Building 16 Release Point U1 Equipment Hatch 17 Release Point U2 Equipment Hatch 18 Release Point Ul MSSV 19 Release Point U2 MSSV 20 Release Point U1 10% ADVs 21 Release Point U2 10% ADVs 22 Release Point U1 MSL Break location 23 Release Point U2 MSL Break location 24 Receptor TSC Normal Intake 25 Receptor TSC Center (location assigned for unfiltered inleakage)

Note 1: Refer to Figure 2.3-5 for location of the above release points / receptors on the site layout and arrangement drawing Note 2: Though not depicted in Figure 2.3-5, atmospheric dispersion factors were also calculated from the closest edge of the containment building to the various receptors; this release point was treated as a diffuse source.

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-147 UNIT 1 CONTROL ROOM INTAKE AND CENTER ATMOSPHERIC DISPERSION FACTORS (SEC/M 3)Release Point and Receptor Unit 1 Containment Building Edge to Unit 1 Control Room (CR) Normal Intake Unit 1 Containment Building Edge to Unit 2 CR Normal Intake Unit 1 Containment Building Edge to Unit 1 CR Emergency Intake 4 Unit 1 Containment Building Edge to Unit 2 CR Emergency Intake 4 Unit 1 Containment Building Edge to CR Center Unit I Plant Vent to Unit I CR Normal Intake Unit 1 Plant Vent to Unit 2 CR Normal Intake Unit 1 Plant Vent to Unit 1 CR Emergency Intake 4 Unit 1 Plant Vent to Unit 2 CR Emergency Intake 4 Unit 1 Plant Vent to CR Center Unit 1 Containment Penetration (GE Area) to Unit 1 CR Normal Intake Unit 1 Containment Penetration (GE Area) to Unit 2 CR Normal Intake Unit 1 Containment Penetration (GE Area) to Unit 1 CR Emergency Intake'Unit 1 Containment Penetration (GE Area) to Unit 2 CR Emergency Intake4 Unit 1 Containment Penetration (GE Area) to CR Center Unit 1 Containment Penetration (GWiFW Area) to Unit 1 CR Normal Intake Unit 1 Containment Penetration (GW/FW Area) to Unit 2 CR Normal Intake Unit 1 Containment Penetration (GW/FW Area) to Unit 1 CR Emergency Intake 4 Unit 1 Containment Penetration (GW/FW Area) to Unit 2 CR Emergency Intake 4 Unit 1 Containment Penetration (GW/FW Area) to CR Center Unit 1 RWST Vent to Unit 1 CR Emergency Intake4*5 Unit 1 RWST Vent to Unit 2 CR Emergency Intake's Unit 1 RWST Vent to CR Center 5 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 1.28E-03 7.12E-04 6.52E-04 3.51E-04 4.11E-04 2.30E-04 1.67E-04 7.95E-05 8.85E-04 4,43E-04 11.67E-03 1,22E-03 9.10E-04 6.57E-04 5,59E-04 3.38E-04 2.26E-04 1.48E-04 1.26E-03 8.96E-04 6.84E-03 3.08E-03 2,24E-03 1.15E-03 3.75E-04 2.33E-04 2.55E-04 1,25E-04 3.22E-03 1.42E-03 4.90E-03 3.45E-03 1.38E-03 9.83E-04 8,20E-04 S40E-04 2.58E-04 1.54E-04 2.59E-03 1.81E-03 3.27E-04 1.90E-04 2.10E-04 9.83E-05 I 07E-03 4 86E-M4 2.87E-04 2.90E-04 1.51E-04 1.49E-04 9.62E-05 8.69E-05 2,63E-05 2.81E-05 1.75E-04 1.77E-04 4.90E-04 4.90E-04 2.68E-04 2.62E-04 1.32E-04 1.12E-04 5.40E-05 5.47E-05 3.44E-04 3.44E-04 1.21E-03 1.12E-03 3.98E-04 3.89E-04 9.12E-05 8.45E-05 4.42E-05 4.38E-05 5.54E-04 5.20E-04 1.37E-03 1.37E-03 3.92E-04 3.88E-04 2,15E-04 1.87E-04 4.95E-05 526E-05 7.29E-04 7.15E-04 7.13E-05 6,99E-05 3.73E-05 3.53E-05 1 99E-04 1 75;F-n4 4-30 Day 2.84E-04 1.371E-04 7.03E-05 2.34E-05 1.65E-04 4.44E-04 2.45E-04 8.38E-05 4.45E-05 2.99E-04 8.75E-04 3.20E-04 6.62E-05 3.55E-05 4.21 E-04 1.28E-03 3.65E-04 1.43E-04 4.48E-05 6.64E-04 5.76E-05 2.86E-05 1 43E-n,4 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-147 (Continued)

UNIT 1 CONTROL ROOM INTAKE AND CENTER ATMOSPHERIC DISPERSION FACTORS (SECIM 3)Release Point and Receptor 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Da Unit I MSSVs to Unit 1 CR Normal Intake 1.2 N/A N/A NIA N/A NIA Unit 1 MSSVs to Unit 2 CR Normal Intake 3 4.29E-03 2.76E-03 1.04E-03 1.06E-03 9.46E-04 Unit I MSSVs to Unit 1 CR Emergency Intake 3.4 4.66E-04 2.92E-04 1.16E-04 1.04E-04 8.08E-05 Unit 1 MSSVs to Unit 2 CR Emergency Intake 3 ,4 3.14E-04 1.53E-04 5.12E-05 5.29E-05 4.38E-05 Unit 1 MSSVs to CR Center 3 1,39E-02 7.40E-03 2.38E-03 2.56E-03 2.15E-03 Unit 1 10% ADVs to Unit 1 CR Normal Intake 1' 2 N/A N/A N/A N/A N/A Unit 1 10% ADVs to Unit 2 CR Normal Intake 3 4.30E-03 2.79E-03 1.05E-03 1.06E-03 9.49E-04 Unit 1 10% ADVs to Unit 1 CR Emergency Intake 3 4 4.66E-04 2.92E-04 1.16E-04 1.04E-04 8.07E-05 Unit 1 10% ADVs to Unit 2 CR Emergency Intake 3,4 3.13E-04 1.54E-04 5.13E-05 5.30E-05 4.39E-05 Unit 1 10% ADVs to CR Center 3 1.39E-02 7.45E-03 2.39E-03 2.59E-03 2.15E-03 Unit 1 MSL Break Location to Unit 1 CR Normal Intake' N/A N/A N/A N/A N/A Unit 1 MSL Break Location to Unit 2 CR Normal Intake 4.23E-03 2.90E-03 1.13E-03 1.11E-03 1.02E-03 Unit 1 MSL Break Location to Unit I CR Emergency Intake 4 4.35E-04 2.94E-04 1.15E-04 1.01E-04 7.76E-05 Unit 1 MSL Break Location to Unit 2 CR Emergency Intake 4 3.06E-04 1.54E-04 5.19E-05 5.32E-05 4.38E-05 Unit 1 MSL Break Location to CR Center 1.24E-02 7.10E-03 2.24E-03 2.43E-03 2.07E-03 Unit 1 FHB to Unit 1 CR Normal Intake 6,98E-03 Unit 1 FHB to Unit 2 CR Normal Intake 2.93E-03 Unit 1 FHB to Unit 1 CR Emergency Intake 4 3.31E-04 Unit I FHB to Unit 2 CR Emergency Intake 4 2.56E-04 Unit 1 FHB to CR Center 3.78E-03 Unit 1 Equipment Hatch to Unit 1 CR Normal Intake 2.61 E-02 Unit 1 Equipment Hatch to Unit 2 CR Normal Intake 2.88E-03 Unit 1 Equipment Hatch to Unit 1 CR Emergency Intake' 4.36E-04 Unit 1 Equipment Hatch to Unit 2 CR Emergency Intake' 2.64E-04 Unit 1 Equipment Hatch to CR Center 5.51 E-03 DCPP UNITS 1 & 2 FSAR UPDATE Notes (Refer to Section 2.3.5.2 for additional detail): 1. ARCON96 based X/Q s are not applicable for these cases given that the horizontal distance from the source to the receptor is 1.5 meters (which is much less than the 10 meters required by ARCON96 methodology).

2. Due to the proximity of the release from the MSSVs/i1 0% (ADVs, to the normal operation control room intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10%

ADVs, the resultant plume from the MSSVsi10%

ADVs will not contaminate the control room normal intake of the affected unit.3. For releases from the MSSVs and 10% ADVs (they are uncapped / vertically oriented, have a high vertical velocity discharge 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), a X/Q reduction factor of 5 is applicable to the values listed above. Since X/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 only up to 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 XIQ values applicable to averaging periods beyond the 2-8 hours averaging period. For assessment of an environmental release between T= 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 ./Q with the factor of 5 reduction is acceptable and slightly conservative.

4. The more favorable

./Q value presented above for the control room pressurization Intakes is further reduced by a factor of 4 to address the "dual intake" credit and the capability of initial selection of the cleaner intake and expectation that the operator will manually make the proper intake selection throughout the event.5. X1Q values for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent.

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-148 UNIT 2 CONTROL ROOM INTAKE AND CENTER ATMOSPHERIC DISPERSION FACTORS (SEC/M 3)Release Point and Receptor 0-2 Hour 2-8 Hour 8-24 Hour Unit 2 Containment Edge to Unit 2 CR Normal Intake Unit 2 Containment Edge to Unit I CR Normal Intake Unit 2 Containment Edge to Unit 1 CR Emergency Intake 4 Unit 2 Containment Edge to Unit 2 CR Emergency Intake 4 Unit 2 Containment Edge to CR Center Unit 2 Plant Vent to Unit 1 CR Normal Intake Unit 2 Plant Vent to Unit 1 CR Normal Intake Unit 2 Plant Vent to Unit 1 CR Emergency Intake 4 Unit 2 Plant Vent to Unit 2 CR Emergency Intakei Unit 2 Plant Vent to CR Center Unit 2 Containment Penetration (GE Area) to Unit 2 CR Normal Intake Unit 2 Containment Penetration (GE Area) to Unit 1 CR Normal Intake Unit 2 Containment Penetration (GE Area) to Unit I CR Emergency Intake4 Unit 2 Containment Penetration (GE Area) to Unit 2 CR Emergency Intake 4 Unit 2 Containment Penetration (GE Area) to CR Center Unit 2 Containment Penetration (GW/FW Area) to Unit 2 CR Normal Intake Unit 2 Containment Penetration (GW/FW Area) to Unit 1 CR Normal Intake Unit 2 Containment Penetration (GW/FW Area)to Unit CR Emergency Intake 4 Unit 2 Containment Penetration (GW/FW Area)to Unit 2 CR Emergency Intake 4 Unit 2 Containment Penetration (GW/FW Area)to CR Center Unit 2 RWST Vent to Unit 1 CR Emergency Intake 4 5 Unit 2 RWST Vent to Unit 2 CR Emergency Intake 4's Unit 2 RWST Vent to CR CenterW 1.96E-03 9.42E-04 6&93E-04 3.84E-04 1.70E-04 1.06E-04 3.85E-04 1.47E-04 1.08E-03 5.46E-04 1.51E-03 9.41E-04 7,88E-04 4.86E-04 2.03E-04 1.29E-04 5.71 E-04 296E-04 1.13E-03 7.08E-04 6.71 E-03 3.12E-03 2.14E-03 1.39E-03 2.28E-04 1.60E-04 3.97E-04 1.76E-04 3.16E-03 1.85E-03 3,55E-03 1.19E-03 1.22E-03 6.26E-04 2.28E-04 1.62E-04 8.64E-04 4.23E-04 2.21E-03 1.17E-03 1.91E-04 1121E-04 3.29E-04 1,61E-04 1.07E-03 5.80E-04 4.48E-04 1 .67E-04 4.23E-05 5.94E-05 2.47E-04 3,86E-04 2.01E-04 5.13E-05 1 .20E-04 2.85E-04 1.21 E-03 5.72E-G4 6.25E-05 6.93E-05 7,17E-04 4.82E-04 2.53E-04 6.58E-05 1.50E-04 4.70E-04 4.58E-05 6,10E-05 2.1 8E-04 1-4 Da 4-30 Day 3.98E-04 3.18E-04 1 A42E-04 1.08E-04 3181E-05 2.95E-05 5.84E-05 4.84E-05 2.12E-04 1.68E-04 3.23E-04 2.23E-04 1.69E-04 1.17E-04 4.32E-05 3.19E-05 1.04E-04 8.19E-05 2.39E-04 1.70E-04 1.22E-03 1.02E-03 4.83E-04 3.62E-04 5.52E-05 4.21E-05 6.44E-05 5.27E-05 6.84E-04 5.43E-04 4.56E&-04 3.03E-04 2,12E-04 1.41E-04 5.43E-05 3199E-05 1.48E-04 1.20E-04 3.90E-04 2.61E-04 4.39E-05 3.53E-05 5.53E-05 4.45E-05 2.19E-04 1.79E-04 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-148 (Continued)

UNIT 2 CONTROL ROOM INTAKE AND CENTER ATMOSPHERIC DISPERSION FACTORS (SEC/M 3)Release Point and Receotor 0-2 Hour 2-8 Hour 8-24 Hour 14 Day 4-30 Day Unit 2 MSSVs to Unit 1 CR Normal Intake 3 387E-03 2.42E-03 9.89E-04 8.17E-04 6.09E-04 Unit 2 MSSVs to Unit 2 CR Normal Intake 1 , 2 N/A N/A N/A N/A N/A Unit 2 MSSVs to Unit 1 CR Emergency Intake 3'4 2.89E-04 1.91E-04 7.45E-05 6.62E-05 5.08E-05 Unit 2 MSSVs to Unit 2 CR Emergency Intake 3'4 4.90E-04 2.29E-04 824E-05 8.07E-05 6.49E-05 Unit 2 MSSVs to CR Center 3 1.22E-02 8.10E-03 3.27E-03 2.76E-03 2.08E-03 Unit 2 10% ADVs to Unit 1 CR Normal Intake 3 3.88E-03 2.43E-03 9.94E-04 8.19E-04 6.10E-04 Unit 2 10% ADVs to Unit 2 CR Normal Intake 1'2 N/A N/A N/A N/A N/A Unit 2 10% ADVs to Unit 1 CR Emergency Intake 3.4 2.87E-04 1.92E-04 7.48E-05 6.61E-05 5.07E-05 Unit 2 10% ADVs to Unit 2 CR Emergency Intake 3'4 4.90E-04 2,29E-04 8.24E-05 8.08E-05 6.48E-05 Unit 2 10% ADVs to CR Center 3 1.22E-02 8.16E-03 3.28E-03 2.78E-03 2.09E-03 Unit 2 MSL Break Location to Unit 1 CR Normal Intake 3.81 E-03 2,40E-03 1.01 E-03 8.09E-04 5.88E-04 Unit 2 MSL Break Location to Unit 2 CR Normal Intake' NIA N/A N/A N/A N/A Unit 2 MSL Break Location to Unit 1 CR Emergency Intake 4 2.75E-04 1.91 E-04 7.45E-05 6.53E-05 4.86E-05 Unit 2 MSL Break Location to Unit 2 CR Emergency Intake4 4.76E-04 2.24E-04 8.14E-05 7.94E-05 6.40E-05 Unit 2 MSL Break Location to CR Center 1.09E-02 7T35E-03 3.011E-03 2.48E-03 1.86E-03 Unit 2 FHB to Unit 1 CR Normal Intake 2.72E-03 Unit 2 FHB to Unit 2 CR Normal Intake 6.98E-03 Unit 2 FHB to Unit I CR Emergency Intake 4 2.49E-04 Unit 2 FHB to Unit 2 CR Emergency Intake 4 3.50E-04 Unit 2 FHB to CR Center 3.71 E-03 Unit 2 Equipment Hatch to Unit 1 CR Normal Intake 2.49E-03 Unit 2 Equipment Hatch to Unit 2 CR Normal Intake 2.51 E-02 Unit 2 Equipment Hatch to Unit 1 CR Emergency Intake4 2.49E-04 Unit 2 Equipment Hatch to Unit 2 CR Emergency Intake 4 4.68E-04 Unit 2 Equipment Hatch to CR Center 5.19E-03 DCPP UNITS 1 & 2 FSAR UPDATE Notes (Refer to Section 2.3.5.2 for additional detail): 1 ARCON96 based x/Qs are not applicable for these cases given that the horizontal distance from the source to the receptor is 1.5 meters (which is much less than the 10 meters required by ARCON96 methodology).

2 Due to the proximity of the release from the MSSVs/1 0% ADVs, to the normal operation control room intake of the affected unit, and due to the high vertical velocity of the steam discharge from the MSSVs/10%

ADVs, the resultant plume from the MSSVs/10%

ADVs will not contaminate the control room normal intake of the affected unit.3 For releases from the MSSVs and 10% ADVs (they are uncapped I vertically oriented, have a high vertical velocity discharge 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), a x/Q reduction factor of 5 is applicable to the values listed above. Since X/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 only up to 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 X/Q values applicable to averaging periods beyond the 2-8 hours averaging period. For assessment of an environmental release between T= 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 X/Q with the factor of 5 reduction is acceptable and slightly conservative.

4. The more favorable X/Q value presented above for the control room pressurization Intakes is further reduced by a factor of 4 to address the "dual intake' credit and the capability of initial selection of the cleaner intake and expectation that the operator will manually make the proper intake selection throughout the event.5. X/Q values for RWST releases to the control room normal intakes are not needed for the dose calculations since the normal intakes are isolated prior to releases occurring from the RWST vent.

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 2.3-149 UNITS 1 AND 2 TECHNICAL SUPPORT CENTER INTAKE AND CENTER ATMOSPHERIC DISPERSION FACTORS (SEC/M 3)Release Point and Receptor UNIT 1 Unit 1 Containment Building Edge to TSC Normal Intake Unit 1 Containment Building Edge to TSC Center Unit 1 Plant Vent to TSC Normal Intake Unit 1 Plant Vent to TSC Center Unit 1 RWST Vent to TSC Normal Intake Unit 1 RWST Vent to TSC Center Unit 1 Containment Penetration (GE Area) to TSC Normal Intake Unit 1 Containment Penetration (GE Area) to TSC Center Unit 1 Containment Penetration (GW/FW Area) to TSC Normal Intake Unit 1 Containment Penetration (GW/FW Area) to TSC Center UNIT 2 Unit 2 Containment Building Edge to TSC Normal Intake Unit 2 Containment Building Edge to TSC Center Unit 2 Plant Vent to TSC Normal Intake Unit 2 Plant Vent to TSC Center Unit 2 RWST Vent to TSC Normal Intake Unit 2 RWST Vent to TSC Center Unit 2 Containment Penetration (GE Area) to TSC Normal Intake Unit 2 Containment Penetration (GE Area) to TSC Center Unit 2 Containment Penetration (GW/FW Area) to TSC Normal Intake Unit 2 Containment Penetration (GW/FW Area) to TSC Center 0-2 Hour 2-8 Hour 8-24 Hour 1-4 Day 4-30 Day 2.57E-04 2.90E-04 3.12E-04 1.54E-04 2.72E-04 2.94E-04 3.64E-04 4.27E-G4 4.80E-04 5.98E-04 5.48E-04 5.57E-04 5.52E-G4 5.43E-04 3.63E-04 3.72E-04 5,47E-04 5.72E-G4 1 .80E-03 1 .83E-03 1.18E-04 1.33E-04 1.77E-04 1.95E-04 1.27E-04 1.38E-04 1.74E-04 1.91 E-04 2.51 E-04 3.03E-04 2.OOE-04 2.01 E-04 2.35E-04 2.16E-04 1.68E-04 1.68E-04 2.41 E-04 2.43E-04 7.72E-04 7.49E-04 4.27E-05 4.98E-05 6.91E-05 7.71 E-05 4.80E-05 5.40E-05 6.55E-05 7.45E-05 8.31 E-05 1.04E-04 8.52E-05 8.81 E-05 1.06E-04 9.97E-05 6.47E-05 6.64E-05 9.36E-05 9.75E-05 3.07E-04 3.16E-04 4.24E-05 4.83E-05 6.29E-05 6.70E-05 4.49E-05 4.89E-05 6.14E-05 6.84E-05 8.64E-05 1 .03E-04 8.37E-05 8.89E-05 8.71 E-05 8.11 E-05 6.04E-05 6.17E-05 8.83E-05 9.12E-05 2.87E-04 2.92E-04 3.50E-05 4.02E-05 5.21E-05 5.67E-05 3.71 E-05 3.97E-05 5.00E-05 5.62E-05 6.95E-05 8.46E-05 6.84E-05 6.92E-05 6.95E-05 6.58E-05 4.91 E-05 5.10E-05 7.02E-05 7.52E-05 2.33E-04 2.41 E-04

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t SI-I I ' -I I-B i J~L~i I I i 1~I II II-I-II II I' -TThIKK r r -I I I-,-~- ---a I II-r.1.-I-E~~~LI D ii:1-II II r -~-I~ I --LI I I II I I II WEST 355330335 Ill B -I-I T.----j ----2k 6 0 0 0 0 0 0 0 0 0 0 0------------

--O W -I -1 7 --. -I ------------------'33 100 150 2800 I1 WI -! '1 I j 02BII I0 d~WI I I 21 1 4 W'I -j 21 -1 J I I t i I v 31 I B2 250 300 350 400 450 500 MNT 1 -. UNIT 2 55 ER 0 C ..B50 6BO 700 750 800 850 COOROINIATE-COLUAII LRE CROSS REFERENCE DCPP UNITS 1 & 2 FSAR UPDATE 3.1.4.1 Criterion 11, 1967 -Control Room (Category B)The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled.

Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10.CFR 20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause.Discussion The plant is provided with a centralized control room common to both units that contains the controls and instrumentation necessary for operation of both units under normal and accident conditions.

The control room is continuously occupied by the operating personnel under all operating and accident conditions.

Sufficient shielding, distance, and containment are provided to ensure that the control room personnel are not subject to radiation exposures in excess of 10 CFR 20 limits. 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 5 rem whele-bedytotal effective dose equivalent (TEDE), Or its equvalent to any part of the body, for the duration of the accident to meet the requirements of GDC 19, 4971-1999 (refer to Section 3.1.4.1.1).

Control room shielding is described in Section 12.1, and postaccident control room exposures are described in Section 15.5.The control room ventilation system is described in Section 9.4.1. It consists of a dual system providing a large percentage of recirculated air. In the event of fire in the control room, provisions are made for 100 percent outside air makeup operation.

In the event of airborne toxic gas outside the control room, provisions are made for operation with 100 percent recirculated air. In the event of airborne radioactivity outside the control room, provisions are made to isolate and pressurize the control room.The risk of fire is minimized by the use of noncombustible and fire retardant materials in the construction of the control room and its furnishings.

Fire fighting equipment is located in the control room, and the use and storage of combustible supplies are minimized.

Provisions are made to enable plant operators to readily shut down and maintain the plant at safe shutdown (Mode 3) by means of controls located outside the control room.3.1.4.1.1 Criterion 19, 1-1999

-Control Room A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection 3.1-1 Revision 22 May 2015 DCPP UNITS 1 & 2 FSAR UPDATE shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

Holders of operating licenses using an alternative source term under 10 CFR 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in 10 CFR 50.2 for the duration of the accident.A conrol1 room shl be provide ftmwh actions can be tae to ope~ae heLL.e power unit safely under normnal conditions and to maintain it in a- safe condition unde: accident conditions, including loss Of coolant accidents.

Adequate radiation pro9tection shall be proYided to permnit access and occu~pancy Of the contro!l room under accsidentt conditions without personne!

receVivig radiation exposrs inexes of 5 remwhlbo, Or 4ts equivalent to any' part of the bodly, for the duration of the accident.Discussion 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 5 rem whole body, Or its cquiv.alent to any part of the bodyTEDE for the duration of the accident to meet the requirements of GDC 19,14-971-1999.

Refer to Section 6.The DCPP design includes use of alternative source terms in accordance with 10 CFR 50.67 and Regulatory Guide 1. 183, July 2000. Refer to Section 6.4 and 15.5.3.1.8.22 Criterion 58, 1967 -Inspection of Containment Pressure-Reducing Systems (Category A)Design provisions shall be made to facilitate the periodic physical inspection of all important components of the containment pressure-reducing systems, such as, pumps, valves, spray nozzles, torus, and sumps.Discussion The containment heat removal systems (CHRS) is comprised of the containment fan cooler system (CFCS) and the containment spray system (CSS). Where practicable, all active components and passive components of the containment cooling systemCHRS are inspected periodically to demonstrate system readiness.

The pressure-containing systems are inspected for leaks from pump seals, valve packing, flanged joints, and safety valves. During operational testing of the containment spray pumps, the portions of the system subjected to pump pressure are inspected for leaks. The containment fan coolers are normally in use, which provides an additional check on the readiness of the system.Five fan coolers are provided.

Each is sized for one-quarter the capacity needed to maintain the containment temperature below 120'F during normal plant operation.

Following a LOCA, two of the five fan coolers provide sufficient capacity to maintain containment pressure below design value when used in conjunction with one containment 3.1-2 Revision 22 May 2015 DCPP UNITS 1 & 2 FSAR UPDATE spray pump during the injection phase. DUn,,, the opeAation is not Fequired.Additional details are found in Section 6.2.recrctJI1,1A:,n AnARP AnT~anment SpFay-3.1-3 Revision 22 May 2015 DCPP UNITS 1 & 2 FSAR UPDATE Note that there is no direct correlation between the design class of an SSC and the seismic design requirements.

Refer to Section 3.2.2.4 for the discussion of seismic qualification requirements.

3.2.2.1.1 PG&E Design Class I PG&E Design Class I is applicable to SSCs that are important to safety, including SSCs required to assure the following: (a) The integrity of the reactor coolant pressure boundary.(b) The capability to shut down the reactor and maintain it in a safe shutdown condition.(c) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100 or 10 CFR 50.67.(d) All plant features designated as PG&E Design Class I are designed to remain functional when subjected to the additional forces associated with the design basis earthquakes that they are required to withstand:

the Design earthquake (DE), the Double Design earthquake (DDE) and/or the Hosgri earthquake (HE). Refer to Section 3.2.2.4 for additional information regarding the earthquake requirements for the design of SSCs.The following SSCs, including their foundations and supports, are classified as PG&E Design Class I: (1) The reactor coolant pressure boundary (2) The reactor core and reactor vessel internals (3) Systems (Refer to Notes (i) and (iii) below) or portions of systems that are required for emergency core cooling, post-accident containment heat removal, or post-accident containment atmosphere cleanup (Refer to Note (iv) below)(4) Systems or portions of systems that are required for reactor shutdown and residual heat removal (5) Those portions of the main steam, feedwater, and steam generator blowdown systems extending from and including the secondary side of the steam generators up to and including the outermost containment isolation valves, and connected piping up to and including the first valve (including 3.2-3 Revision 15 September 2003 DCPP UNITS 1 & 2 FSAR UPDATE PG&E Design Class II or III may have some seismically qualified components (as defined in the Q-list) and have no PG&E Quality/Code Class designation.

However, some PG&E Design Class II components have been seismically designed; e.g., items in the Seismically Induced Systems Interaction Program, specific components required for post-HE shutdown, (e.g. CCW header C components), and items that were designed for the DE (for specific SSC classifications refer to the Q-list). For this reason, there is not a direct correlation between PG&E Design Class and the level of seismic qualification (except that PG&E Design Class I SSCs have seismically qualified components).

In addition, the classification of seismically qualified components does not indicate which of the three design basis earthquakes the SSC has been qualified for, nor whether that qualification is for passive or active function.

The design basis function of the equipment determines the type of seismic qualification required.

These classifications-and their relationships are summarized in Tables 3.2-1 and 3.2-2.The definitions of the three DCPP Earthquakes are included below: Design Earthguake The design earthquake (0.2g) is defined as the maximum size earthquake that can be expected to occur at DCPP during the life of the reactor. The design earthquake is the equivalent of the operating basis earthquake, as described in 10 CFR 100, Appendix A.Double Design Earthquake The double design earthquake (0.4g) is defined as the hypothetical earthquake that would produce accelerations twice those of the design earthquake.

The double design earthquake is the equivalent of the safe shutdown earthquake, as described in 10 CFR 100, Appendix A.Hosgri Earthquake The Hosgri earthquake (0.75g) is defined as the predicted ground motion at DCPP due to a Richter magnitude

7.5 earthquake

on the offshore Hosgri fault. The Hosgri earthquake does not correspond to an operating basis earthquake or safe shutdown earthquake.

3.2.2.5 PG&E Instrument System Classifications DCPP Instrumentation is classified as Instrument Class IA, IB, IC, ID, and II depending upon the function performed.

Instrument Class IA, IB, IC and ID devices have PG&E Design Class I functions or other special requirements.

Instrument Class II devices have PG&E Design Class II functions.

Instrument classes are defined as follows: (1) Instrument Class IA -Class IA instruments and controls are those that initiate and maintain safe shutdown of the reactor, mitigate the consequences of an accident, or prevent exceeding 10 CFR Part 100 or 3.2-13 Revision 15 September 2003 DCPP UNITS 1 & 2 FSAR UPDATE 10 CFR 50.67 off-site dose limits. Class IA instruments and controls enable the PG&E Design Class I systems to automatically accomplish their appropriate safety functions, or they enable operating personnel to manually accomplish appropriate safety actions when a monitored condition reaches a preset level.(2) Instrument Class 1B -Class IB instruments and controls are those that are required for post-accident monitoring (PAM) of Category 1 and 2 variables in accordance with Regulatory Guide 1.97, Revision 3 (refer to Table 7.5-6). Regulatory Guide 1.97, Revision 3, Category 3 instruments required for PAM are Instrument Class II. PAM instrumentation enables operating personnel to: (a) Determine when a condition monitored by PG&E Design Class I instruments and controls reaches a level requiring manual action (b) Assess the accomplishment of plant safety functions (c) Assess fission product barrier integrity (d) Provide information to indicate the operability of individual systems important to safety (e) Assess the magnitude of radioactive materials releases from the plant PAM instruments are further divided into five variable types (A through E)as detailed in Regulatory Guide 1.97, Revision 3. These are further detailed in Section 7.5.(3) Instrument Class IC -Instrument Class IC instruments and controls have the passive function of maintaining the pressure boundary integrity (PBI)of PG&E Design Class I piping systems. Passive valve operators that are within PG&E Design Class I piping systems are Instrument Class IC and are seismically analyzed to assure structural and pressure boundary integrity of the valve operator assembly.

This classification also includes instruments installed in PG&E Design Class I HVAC ducting that are required to maintain pressure boundary integrity.

In addition, this classification is used for instruments that are part of seismically qualified, PG&E Design Class II systems and denotes that the instruments are required to maintain their pressure boundary integrity.

(4) Instrument Class ID -Instrument Class ID instruments and controls are components that have certain PG&E Design Class I attributes, but do not require conformance with all Instrument Class IA, IB or IC requirements.

The Instrument Class ID designation signifies that only certain design 3.2-14 Revision 15 September 2003 DCPP UNITS 1 & 2 FSAR UPDATE Environmental Qualification Report (Revision

1) (Reference

14) to the NRC. In addition to the item-by-item NUREG-0588 comparison, this report consolidated and summarized extensive information concerning the development and implementation of the EQ program for DCPP. NRC review and acceptance of the DCPP EQ program was documented in Appendix B of SSER No. 15, dated September 1981 (Reference 15).In 1983, the NRC issued 10 CFR 50.49 (Reference

16) to codify requirements for the environmental qualification of electric equipment important to safety. This regulation now forms the basis of the EQ program at DCPP. The regulation applies to the following three categories of equipment:

Safety-related electric equipment

("Class 1 E"): This equipment is that relied upon to remain functional during and following design basis events to ensure (a) the integrity of the reactor coolant pressure boundary, (b) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (c) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the 10 CFR 50.67 400 guidelines.

It is noted that DCPP has incorporated a full implementation of Alternative Source Terms (AST) as defined in Regulatory Guide 1.183, July 2000, Section 1.2.1. However, the post-LOCA integrated doses utilized for radiological environmental qualification of PG&E Design Class I equipment continue to be based on TID-14844 assumptions.

This approach is acceptable based on Section 1.3.5 of Regulatory Guide 1.183 and NUREG-0933, Section 3.0, Item 187 which concluded that there is no clear basis for a requirement to modify the design basis for EQ to adopt AST since there would be no discernible risk reduction associated with adopting AST for EQ.Nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the safety functions specified above by the safety-related equipment.

Certain postaccident monitoring equipment as described in RG 1.97, Revision 3 (Reference 17).The regulation explicitly excludes equipment located in a mild environment.

A mild environment is defined as an environment that would at no time be significantly more severe than the environment that would occur during normal plant operation, including anticipated operational occurrences.

Also significant in 10 CFR 50.49 are the provisions in its paragraphs (k) and (I): "(k) Applicants for and holders of operating licenses are not required to requalify electric equipment important to safety in accordance with the provisions of this section [i.e., in accordance with 10 CFR 50.49] if the Commission has previously required qualification of the equipment in 3.11-3 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Steam generator tube inspection is performed in accordance with the Technical Specifications (Reference

6) and the DCPP surveillance test procedure.

Eddy current non-destructive testing is used to perform tube inspections.

The steam generator tube surveillance program ensures that the structural and leakage integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on NEI 97-06 (Reference 5). Inservice inspection of SG tubing is essential in order to maintain surveillance of the conditions of the tubes in the event there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of SG tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.Tube degradation will be detected during scheduled inservice SG tube examinations.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness.

Plugging is required for all tubes with imperfections exceeding the plugging limit defined in the Technical Specifications.

Degradation may be left in service if qualified non-destructive examination sizing techniques verify that the imperfection is less than the plugging limit (reference PG&E response to NRC Generic Letter 97-05).5.5.2.5.2 Primary-to-Secondary Leakage The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the SG tubes. The extent of cracking during plant operation is limited by the limitation on SG tube leakage between the Reactor Coolant System and the Secondary Coolant System (primary-to-secondary leakage = 150 gallons per day per SG). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

DCPP has demonstrated that primary-to-secondary leakage of 150 gallons per day per SG can readily be detected during power operation.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.The dose consequence analyses in Section 15.5.18.1 proVides the assessment for address accident-induced leakage up to 0.75 gpm (total for all 4 SGs)10.5 gpm at standard temperature and pressureroom tcmperature conditions in an', one SG following an SLB.5.5.3 REACTOR COOLANT PIPING Reactor coolant piping provides a flowpath connecting the major components of each RCS loop.5.5.3.1 Design Bases 5.5-16 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE from the containment atmosphere during the injection and recirculation phases. The CSS function of mixing the containment atmosphere for hydrogen control is discussed below.(4) The Containment Fan Cooler System (CFCS)The CFCS functions in conjunction with the CSS to limit the temperature and pressure in the containment structure in the event of a LOCA or MSLB (refer to Section 6.2.2). The CFCS also provides mixing of the sprayed and unsprayed regions of the containment atmosphere to improve airborne fission product removal (refer to Section 6.2.3). The CFCS function of mixing the containment atmosphere for hydrogen control is discussed below.(5) The Spray Additive System The SAS functions by adding sodium hydroxide, an effective iodine scrubbing solution, to the CSS water to reduce the content of iodine and other fission products in the containment atmosphere and prevent the re-evolution of the iodine in the recirculated containment spray and core cooling solution following a LOCA (refer to Section 6.2.3).(6) Containment Combustible Gas Control The long-term buildup of gaseous hydrogen in the containment following a LOCA is primarily controlled by ensuring a mixed containment atmosphere and providing equipment for monitoring hydrogen concentrations.

The CFCS and the CSS are is-the primary means credited for containment atmosphere mixing (refer to Section 6.2.5).(7) The Fuel Handling Building Ventilation System (FHB VS)The FHBVS provides the capability for a-significant reduction in the amounts of volatile radioactive materials that could be released to the atmosphere in the event of a major fuel handling accident (refer to Section 9.4.4).(8) The Auxiliary Building Ventilation System(ABVS)

The ABVS provides the capability for significant reductions in the amounts of volatile radioactive materials that could be released to the atmosphere in the event of leakage from the residual heat removal (RHR) system recirculation loop following a LOCA (refer to Section 9.4.2).6.1-3 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.1.6 Materials Containment structural heat sink materials used for containment integrity analyses following a LOCA or main steam line break are listed in Table 6.2D-19; corresponding material properties are listed in Table 6.2D-20 (refer to Section 6.2D.3.2.4).

A current record of paint used on containment heat sink structures and equipment is maintained in engineering files.6.2.2 CONTAINMENT HEAT REMOVAL SYSTEMS The containment heat removal systems (CHRS) are the containment fan cooler system (CFCS) and the containment spray system (CSS). The functional performance objectives of the CHRS are: (1) The CFCS limits the containment ambient temperature during normal plant operating conditions (refer to Section 9.4.5);(2) The CFCS and CSS reduce the containment ambient temperature and pressure following a loss-of-coolant accident (LOCA) or main steam line break (MSLB) inside containment.

While performing this cooling function, the CHRS also helps limit offsite radiation levels by reducing the pressure differential between containment and outside atmosphere, thus reducing the driving force for leakage of fission products from the containment atmosphere; (3) The CFCS provides mixing of the sprayed and unsprayed regions of the containment to improve airborne fission product removal (refer to Section 6.2.3);(4) The CSS removes airborne fission products from the containment atmosphere following a LOCA (refer to Section 6.2.3);(5) The CSS, in conjunction with the spray additive system (SAS), prevents the re-evolution of the iodine in the recirculated containment spray and core cooling solution (i.e. sump water) following a LOCA (refer to Section 6.2.3);(6) The CFCS and CSS provides a mixed atmosphere for hydrogen control (refer to Section 6.2.5).Used in conjunction with one another during the injection phase, one containment spray pump and two containment fan cooler units (CFCUs) will provide the heat removal capability to maintain the post-accident containment atmospheric pressure and temperature below the design values of 47 psig and 271'F, respectively.

The CFCS is credited for long-term containment pressure and temperature control throughout the injection and recirculation phases following a LOCA or MSLB. The operation of CSS in 6.2-12 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE the injection phase is credited only for operation during the spray ,nje.tion pha.efor heat removal following a LOCA or MSLB. To address the delayed radioactivity release associated with the core damage sequence of Alternate Source Terms (AST), credit is taken for CSS operation in both the injection and folloWing a LOCA or .SLB;, it is not required for operat on durig the recirculation phases to support fission product removal following a LOCA (refer to Section 15.5). for mitigating the of a LOGCA.. The physical SSC design bases, testing, and inspection requirements of the CFCS and CSS are discussed in this section.6.2.2.1 Design Bases 6.2.2.1.1 General Design Criterion 2,1967 -Performance Standards The CFCS and CSS are designed to withstand the effects of, or are protected against, natural phenomena, such as earthquakes, flooding, tornadoes, winds, and other local site effects.6.2.2.1.2 General Design Criterion 10, 1967 -Containment The CFCS and CSS are designed to aid other ESFs in retaining the functional capability of the containment to protect the public in the event of gross equipment failures, such as a large coolant boundary break.6.2.2.1.3 General Design Criterion 11, 1967 -Control Room The CFCS and CSS are designed to support actions to maintain and control the safe operational status of the plant from the control room.6.2.2.1.4 General Design Criterion 12, 1967 -Instrumentation and Control Systems The CFCS and CSS are provided with instrumentation and controls as required to monitor and maintain the CHRS variables within prescribed operating ranges.6.2.2.1.5 General Design Criterion 15, 1967 -Engineered Safety Features Protection Systems The CFCS and CSS are provided with instrumentation for sensing accident conditions.

6.2.2.1.6 General Design Criterion 19, 197-11999

-Control Room The CFCS and CSS, in conjunction with the SAS, are designed to limit radiation exposure to personnel to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE, or its equivalent to any part of the body,-for the duration of the accident.6.2-13 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.2.2 System Description The CHRS is designed to provide sufficient heat removal capability to maintain the post-accident containment atmospheric pressure and temperature below the design values of 47 psig and 271 'F, respectively (refer to Section 3.8.1.1).

The containment atmospheric temperature after a MSLB does briefly increase greater than the containment design temperature, but, as discussed in Section 6.2D.4.2.4, there is no explicit design temperature limit for a MSLB. Heat energy sources considered are described in Section 6.2D.2.1.12.

The CFCS also functions during normal operating conditions to limit the containment ambient temperature to 120'F and is described in Section 9.4.5.The CFCS, shown schematically in Figure 9.4-4, consists of five identical fan coolers, each including cooling coils, fan and drive motor, locked-open air flow dampers and pressure relief dampers, duct distribution system, instrumentation, and control. During operation of the units, air is drawn into the cooling coils, cooled, and discharged back through the ductwork to the containment atmosphere.

The design parameters for the CHRS components and materials are listed in Table 6.2-26. Codes and standards used as a basis for the design of the components are given in Table 6.2-25.Ductwork distributes the cooled air to the various containment compartments and areas.During normal and post-accident operations, the flow sequence through each air fan cooler is as follows: locked-open normal and accident air flow dampers, cooling coils, fan, and distribution ductwork.Airflow through the exhaust ducting, towards the fan, can occur when the fan is idle.Incorporated into the fan/motor coupling is an anti-reverse rotation device (AARD) that precludes the fan motor from rotating backwards.

This device replaces backdraft dampers previously installed in the fan discharge duct.The CSS, shown schematically in Figure 3.2-12, consists of two pumps, spray ring headers and nozzles, valves, and connecting piping. Following a LOCA, water from the refueling water storage tank (RWST) is initially used for containment spray. Later, water recirculated from the containment sump supplied by the residual heat removal (RHR) pumps for recirculation spray. if no component failures affec.t the RHR train capability, the emergency procedures direct the initiation of sprays.HeweyeF-sSingle failures that result in the loss of one RHR train cause the decision of hew-teare taken into consideration in system alignment to ensure appropriate division of divide-the recirculation flow between containment spray and core injection to ensure fission product removal from the containment atmosphere and core cooling.to-be-made-by the technical support center in charge of accident mitigation.

6.2-18 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE (2) The CCW exit temperatures are indicated locally outside the containment building.(3) Bearing temperatures are indicated and alarmed on the plant process computer.6.2.2.3.5 General Design Criterion 15, 1967 -Engineered Safety Features Protection Systems The CHRS is designed with instrumentation that measures containment pressure.

The CSS will be actuated through ESFAS by a "P" signal (high-high containment pressure actuation setpoint) on coincidence of two-out-of-four high-high containment pressure signals. Coincidence of "S" (safety injection) and "P" signals starts the containment spray pumps and opens the discharge valves to the spray headers. Refer to Section 7.3 for a description of the ESFAS.6.2.2.3.6 General Design Criterion 19, 49741999-Control Room The CSS is designed to deliver borated water from the RWST to the containment atmosphere during the injection phase following a LOCA. The CSS, in conjunction with the CFCS, reduces the containment ambient temperature and pressure following a LOCA or MSLB, thus reducing the driving force for leakage of fission products from the containment atmosphere.

The CFCS provides mixing of the sprayed and unsprayed regions of the containment to improve airborne fission product removal. The SAS, as discussed in Section 6.2.3, injects sodium hydroxide to the suction of the CSS for delivery to the containment atmosphere to prevent the re-evolution of the iodine in the recirculated core cooling solution.

The CSS removes radioactive iodines and particulates from the containment atmosphere during the injection and recirculation phases following a LOCA to ensure that radiological exposures for control room personnel are within 5 rem TEDE, for the duration of the accidentwhoe-,bdy.

Refer to Section 15.5 for radiological consequences of plant accidents.

Therefore, the CSS, in conjunction with the CFCS and SAS, limits the control room doses to 5 rem TEDEwhele body.6.2.2.3.7 General Design Criterion 21, 1967 -Single Failure Definition The CFCS and CSS are designed such that no single failure in either train will prevent the CHRS from performing its design function.

The CSS is comprised of two completely independent trains of containment spray. Each CSS train of pumps and valves are powered by separate Class 1 E 4.16-kV and 480-V power supplies, respectively.

The CFCS is comprised of five independent fan coolers; two are powered from Class 1 E 480-V Bus F, two are powered from Class 1 E 480-V Bus G, and one is powered from Class 1 E 480-V Bus H.6.2-24 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Any single failure will still leave sufficient CSS and CFCS capability to together mitigate DBAs. Used in conjunction with one another during the injection phase, one containment spray pump and two containment fan cooler units will provide the heat removal capability to maintain the post-accident containment pressure below the design value of 47 psig. CHRS design parameters are listed in Table 6.2-26.A single failure analysis on all active components of the CHRS was performed to show that the failure of any single component will not prevent performance of the design function.

This analysis is summarized in Table 6.2-27.6.2.2.3.8 General Design Criterion 37, 1967 -Engineered Safety Features Basis for Design Operation of tThe CSS i6 cred.ted only for opera#tio during the injection phase following a LOCA and MSLB is credited for heat removal. As stated in Section 6.2.2, credit is taken for CSS operation during the injection and recirculation phases to support fission product removal following a LOCA. The CFCS is credited for containment pressure and temperature control and mixing of the sprayed and unsprayed regions of the containment to improve airborne fission product removal throughout the injection and recirculation phases following a LOCA or MSLB.The CSS and CFCS limit the effects of post blowdown energy additions to the containment during the injection phase following a LOCA. For a detailed description of the analytical methods and models used to assess the performance capability of the CHRS, refer to the containment integrity analysis presented in Appendix 6.2D.The CHRS provides a backup to the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. As discussed in this section, the CHRS is designed to withstand any size reactor or secondary coolant pressure boundary break, including a LOCA or MSLB.6.2.2.3.8.1 Containment Spray System The CSS, in conjunction with the CFCS, reduces the containment ambient temperature and pressure following a LOCA or MSLB, thus reducing the driving force for leakage of fission products from the containment atmosphere.

The CSS removes radioactive

-airborne iodine and particulates from the containment atmosphere following a LOCA (refer to Section 6.2.3). The CSS, in conjunction with the spray additive system (SAS), prevents the re-evolution of the iodine in the recirculated spray and core cooling solution (also known as sump water) following a LOCA (refer to Section 6.2.3).6.2.2.3.8.1.1 Containment Spray Pumps The containment spray pumps are designed to perform at rated capacity against a total head composed of containment design pressure, nozzle elevation head, and the line and nozzle pressure losses. Adequate NPSH is available for operation of the 6.2-25 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.2.3.8.1.3.2 Recirculation Phase if-the-CSS is-used-operation in the spray recirculation -phasemode is manually initiated within 12 minutes of termination of injection spray. -,-Recirculation spray is provided by the RHR pumps, which draw suction from the containment sump. During the spray recirculation phase, a single RHR pump delivers a minimum flow rate of 1211 gpm to a single spray header. The minimum ECCS core flow available when containment spray is operating in the recirculation mode exceeds the required core flow acceptance criteria taking into consideration the worst case single active failure to minimize the number of pumps available for core cooling. Spray recirculation phase operation of the RHR pumps is discussed in Section 6.3.2.4.3.1 and Section 5.5.6.6.2.2.3.8.2 Containment Fan Cooler System 6.2.2.3.8.2.1 Containment Fan Coolers In addition to limiting maximum containment pressure and temperature, the containment fan coolers provide mixing of the containment atmosphere for radioactive iodine and particulate removal, and control of hydrogen buildup as discussed in Sections 6.2.3 and 6.2.5, respectively.

The heat removal capability of the containment fan cooler cooling coils is 81 x 106 Btu/hr per fan cooler unit at saturation conditions (271 OF, 47 psig), with 2000 gpm cooling water supply at 125 0 F (refer to Table 6.2-26).The design internal pressure of each coil is 200 psig and the coils can withstand an external pressure of 47 psig at a temperature of 271'F without damage.Each cooling coil assembly has a top and bottom horizontal coil casing made of galvanized steel. The safety function of the cooling coil casings is to fill the air gap between each stacked cooling coil assembly and direct airflow through the cooling fins to ensure that adequate heat transfer occurs within the containment fan cooler units.Each fan can provide a minimum flowrate of 47,000 cfm (in low speed) when operating against the system resistance of approximately 3-3/4 inches of water existing during the accident condition.

Minimum flowrate assumptions used to establish containment mixing for fission product removal is discussed in Section 6.2.3 and 15.5.As discussed in Section 6.2.2.2.1.2.3, each CFCS unit is comprised of four, pre-fabricated modular units. As a result of the modifications made to the fan cooler enclosure assemblies, only Module 3 serves to direct the airflow into the coils for normal and post-LOCA operations.

Modules 1 and 2 are no longer required for service.Module 1 contains the locked-open accident flow inlet dampers. Locking the accident inlet damper open prevents module overpressurization during a LOCA. Module 2 contains the locked-closed accident flow outlet dampers. Modules 1 and 2 are open to 6.2-27 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.3 CONTAINMENT AIR PURIFICATION AND CLEANUP SYSTEMS The containment air purification and cleanup systems are made up of the spray additive system (SAS), the containment spray system (CSS), and the containment fan cooler system (CFCS). The functional performance objectives of the containment air purification and cleanup systems are: (1) The SAS provides a chemical additive to the CSS to prevent the re-evolution of the iodine in the recirculated spray and core cooling solution (also known as sump water) following a loss-of-coolant-accident (LOCA);(2) The CSS removes airborne radioactive iodine and particulates from the containment atmosphere following a LOCA;(3) The CFCS provides mixing of the sprayed and unsprayed regions of the containment to improve airborne iodine removal. Mixing the containment atmosphere maximizes the gas volume treated by the containment spray;(4) The CFCS and CSS reduce the containment ambient temperature and pressure following a LOCA or a main steam line break (MSLB). While performing this cooling function, the containment heat removal system (CHRS) also helps limit offsite radiation levels by reducing the pressure differential between containment and outside atmospheres, thus reducing the driving force for leakage of fission products from the containment atmosphere (refer to Section 6.2.2);(5) The CFCS and CSS provides a mixed containment atmosphere for hydrogen control (refer to Section 6.2.5);(6) The CFCS limits the containment ambient temperature during normal plant operating conditions (refer to Section 9.4.5).The SAS boundary consists of all piping and valves between, and including, the isolation valve from the refueling water storage tank (RWST), the spray additive tank and the spray eductors.

The safety function of iodine removal occurs during operation of the containment spray system and while core coolant is retained in the containment sump.The physical SSC design bases, testing and inspection requirements of the CSS and CFCS are discussed in Section 6.2.2, therefore the following section addresses the SAS SSCs only.6.2-43 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The two small charcoal filter units in the containment air purification system are not classified as ESFs and are described in Section 9.4.5. These units are not necessary for cleanup during accident conditions.

6.2.3.1 Design Bases 6.2.3.1.1 General Design Criterion 2,1967 -Performance Standards The SAS is designed to withstand the effects of, or is protected against, natural phenomena, such as earthquakes, winds and tornadoes, floods and tsunamis, and other local site effects.6.2.3.1.2 General Design Criterion 3,1971 -Fire Protection The SAS is designed and located to minimize, consistent with other safety requirements, the probability and effects of fires and explosions.

6.2.3.1.3 General Design Criterion 11, 1967- Control Room The SAS is designed to support actions to maintain and control the safe operational status of the plant from the control room.6.2.3.1.4 General Design Criterion 12, 1967 -Instrumentation and Control Systems Instrumentation and controls are provided as required to monitor and maintain SAS variables within prescribed operating ranges.6.2.3.1.5 General Design Criterion 19, 197-11999

-Control Room The SAS, in conjunction with the iodine removal function of the CSS and CFCS, is designed to support radiation protection to permit access and occupancy of the control room under accident conditions without personnel receiving radiation in excess of 5 rem whole body, or its equivalent to any part of the body TEDE, for the duration of the accident.6.2.3.1.6 General Design Criterion 21, 1967 -Single Failure Definition The SAS is designed to tolerate a single failure during the period of recovery following an accident without loss of its protective function, including multiple failures resulting from a single event, which is treated as a single failure.6.2.3.1.7 General Design Criterion 37, 1967 -Engineered Safety Features Basis for Design 6.2-44 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The SAS is designed for in-situ periodic testing and surveillance.

6.2.3.1.15 General Design Criterion 65, 1967 -Testing of Operational Sequence of Air Cleanup Systems The SAS is designed with the capability to test, under conditions as close to design as practical, the full operational sequence that would bring the SAS into action.6.2.3.1.16 General Design Criterion 70,1967 -Control of Releases of Radioactivity to the Environment The SAS, in conjunction with the iodine removal function of the CSS and CFCS, is designed with provisions for maintaining control over the plant's radioactive gaseous effluents.

6.2.3.1.17 10 CFR 50.55a(f)

-Inservice Testing Requirements American Society of Mechanical Engineers (ASME) code components within the SAS are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.

6.2.3.1.18 10 CFR 50.55a(g)

-Inservice Inspection Requirements ASME code components within the SAS are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.

6.2.3.1.19 Regulatory Guide 1.183, July 2000, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors The steady state elemental iodine removal coefficients, and the time dependent particulate aerosol removal coefficients, are developed in accordance with Regulatory Guide 1.183, July 2000, Appendix A, Section 3.3.6.2.3.1 .19-20 Generic Letter 89-10, June 1989 -Safety-Related Motor-Operated Valve Testing and Surveillance The SAS PG&E Design Class I motor-operated valves (MOVs) meet the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.6.2.3.2 System Description 6.2.3.2.1 General Description The SAS is designed to add sodium hydroxide to the containment spray water to 6.2-46 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The SAS is designed to the fire protection guidelines of Branch Technical Position APCSB 9.5-1 (refer to Appendix 9.5B Table B-i).6.2.3.3.3 General Design Criterion 11, 1967- Control Room Each of the system's MOVs can be operated individually by switches in the control room. Valve position indication on the same control room panel is used to verify valve operability.

The spray additive tank level instruments provide two alarms to announce when the solution in the tank has dropped below a level approaching the Technical Specification minimum requirements.

6.2.3.3.4 General Design Criterion 12, 1967 -Instrumentation and Control Systems Analog and logic channels employed for initiation of SAS operation are discussed in Section 7.3. All alarms will be annunciated in the control room.The SAS will be actuated by a "P" signal initiated either manually from the control room or on coincidence of two sets of two-out-of-four high-high containment pressure signals.Coincidence of "S" and "P" signals will start the containment spray pumps and open the discharge valves to the spray headers. The "P" signal alone will open the valves associated with the spray additive tank.A locally mounted indicator on the nitrogen line monitors the spray additive tank pressure while adding nitrogen and during periodic inspections.

A flow element is located in the discharge line from the spray additive tank. Flow indication is provided in the control room.For the spray additive tank, two separate instruments are provided:

one to supply readout in the control room and the other to provide local indication.

6.2.3.3.5 General Design Criterion 19,--9741999

-Control Room The CSS provides iodine removal from the containment atmosphere.

The CFCS provides mixing of the containment atmosphere to maximize iodine removal. The SAS increases pH in the containment recirculation sump water. The SAS, CSS and CFCS work in conjunction to ensure that radiological exposures for control room personnel are within 5 rem TEDE w4GIe-for the duration of the accident.

bG* R-efer to Section 15.5 for radiological consequences of plant accidents.

6.2.3.3.6 General Design Criterion 21, 1967 -Single Failure Definition 6.2-49 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE A failure analysis was conducted on the active components, MOVs 8994A/B, of the SAS to show that failure of any single active component will not prevent the design function from being fulfilled.

This analysis is summarized in Table 6.2-38.MOV 8992 in the spray additive line, which is required only in the short term, is not included in the single failure analysis of Table 6.2-38 because it is not an active component.

It performs no active function, is normally open, receives a "P" signal to ensure positive opening, and is designed to fail as is, i.e., in the open position.In addition, during power operation, power is removed from the single inline SAS outlet MOV 8992 at the circuit breaker at the motor control center with the valve in the open position to prevent a single active failure.The single failure analysis for the CSS is given in Section 6.2.2.6.2.3.3.7 General Design Criterion 37, 1967- Engineered Safety Features Basis for Design The design basis of the system-SAS is two-fold: (1) Sufficient sodium hydroxide must be added to the containment spray water to ensure rapid absorption by the spray of elemental iodine present in the containment atmosphere following a LOCA.(2) During the injection phase operation of the spray pumps, a sufficient amount of sodium hydroxide must be carried to the containment sump water via the containment spray to ensure retention of the iodine in the sump solution.Performance of the (eS9 as an idine emoal Meshuf isservatively evaluated at the containment design temperatumohre, afdresue.

Sinec this peak pfeasore condition is expected to exist for a few minutes at most, and mass transfer parameters, and spoay feioWate impreve with deaeaing pressure, an appreoiable yarogin is added to the evaluation.

The design eahe remtoal onitant for the SS pr ) proVided inr Table 6.2-36 w aS calculated by applying the moedel derived in Sectioan 66.2.3.23.7.23

_At this back pressure coendlition to the sprayed portion of the coentainment volume.The Spray-systeMCSS, by virtue of the large contact surface area provided between the droplets and the containment atmosphere, affords an excellent means of absorbing radioactive iodine released as a consequence of a LOCA. Sodium hydroxide is added to the spray fluid to increase the absorption of iodine in the spray to the point where the rate of absorption is largely limited by the transfer rate through the gas film surrounding the drops. Reference 5 describes in detail the analytical and experimental basis for the above containment atmosphere iodine removal mechanism.

The approach used is summarizedl below-.6.2-50 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE However, the SAS does not have a significant effect on aerosol particle removal as this removal process is largely mechanical.

The SAS is dependent upon the CSS for operation.

The CSS can function with one spray train operating (abnormal operating mode) or with both spray trains operating (normal operating mode). In addition, the operation of one or both ECCS trains affects the rate of withdrawal of water from the RWST, the duration of the spray injection phase, and thus the amount of sodium hydroxide added to the containment.

The removal has been evaluated for the design case, a double-ended LOCA, assuming that: (1) Only one-out-of-two spray pumps operate (one spray train operating)

(2) The ECCS operates at its maximum capacity (two ECCS trains operating)

(3) Borated water is retained in the RWST for the exclusive use of the spray during the first part of the ECCS recirculation phase The second assumption maximizes ECCS flow from the RWST. Overall, the first two assumptions give the most conservative prediction of sodium hydroxide introduction into the containment (minimum containment recirculation sump pH). The third assumption ensures that sufficient sodium hydroxide solution is added to provide for a sump pH of 8.0 or greater when spray injection terminates.

Figure 6.2-15 presents various resulting sump pH versus time curves.The variation of sump pH with time after the accident is shown for various cases in Figure 6.2-15 (Reference 51). At the time spray injection terminates, the sump water has reached an equilibrium pH of at least 8.0. The sump pH Will remain at the same pH after the spa ;netian phase because no add itional water is added to the 6ump-.Any reevolution of dissolved iodine from the sump to the containment atmosphere depends upon the concentration gradient between the liquid and vapor phases. The equilibrium between these iodine concentrations is given by the partition coefficient, H, and is a function of concentration, pH, and temperature.

The partition coefficient at pH 8.0 exceeds the value of approximately 4 x 103 required to maintain a decontamination factor of 100 in the containment atmosphere for sump temperature above 1200 F, and thus a containment atmosphere decontamination factor of 100 or greater can be expected.

Figure 6.2-16 presents equilibrium elemental iodine partition coefficients in the containment at various temperatures for the minimum sump pH case. The equations given by Eggleton (Reference

8) were used to determine the partition coefficients.

Although the iodate reaction is expected to contribute significantly to the iodine partition at high sump pH values, it has been neglected in these calculations in the interest of conservatism.

6.2-51 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE SRP Section 6.1.1, Revision 2,Section II, SRP Acceptance Criteria 2A (Reference 72)and SRP Section 6.5.2, Revision 4, Acceptance Criteria 1G (Reference

61) require that the pH of the sump water be controlled to maintain a minimum value of 7.0 following a LOCA. This is required to prevent re-evolution of the iodine that has been removed from the containment atmosphere by the containment spray and washed into the sump water. A neutral pH also limits material degradation, in particular, stress corrosion cracking of austenitic stainless steel components in the post LOCA environment.

Long-term retention of iodine in the sump fluid is strongly dependent on the pH. In determining whether the plant chemical addition system is adequate for long-term pH control, long-term production of acids (hydrochloric acid (HCI) and nitric acid (HNO 3)) by irradiation was addressed.

A conservative analysis was performed to confirm that the sump water pH at thirty (30) days following a LOCA remains greater than 7.5. The analysis assumed the minimum volume / concentration values for NaOH, in combination with the maximum volume / boration values for the water sources contributing to the sump water volume.SRP Section 6.5.2, Revision 4, Acceptance Criteria I1.1.g, states that long term iodine retention may be assumed when the equilibrium sump water pH, after mixing and dilution with the primary coolant, is above 7.0.NUREG/CR-5732, April 1992 (Reference 73), states that iodine re-evolution is not a factor if the ultimate sump water pH of > 7 is achieved prior to the time when iodine re-evolution could potentially occur. Section 3.1 of NUREG /CR-5732, April 1992 notes that the following phenomena occur during the first time interval between t = 0 to t =1000 min, a) events "leading" to the formation of 12 by radiolysis, and b) all HI effects except for those related to pH. Re-evolution (i.e. vapor phase elemental iodine produced by radiolysis and partitioned between the aqueous and gas) can occur in the second time interval which is from t=1000 min to t-2 to 3 weeks. Thus for Light Water Reactors, a pH of 7 must be achieved within 1000 min of the initial post-LOCA release.At DCPP, RWST drain-down and chemical addition is essentially complete within an hour post-LOCA; thus it is expected that as a result of recirculation, the sump water will be well mixed by t=1000 minutes or -16 hours.Based on the above assessment it was concluded that the DCPP sump water pH will remain greater than 7.0, and the post-LOCA dose consequence analyses need not consider iodine re-evolution from the sump fluid.6.2.3.3.7.1 Mixing between Sprayed and Unsprayed Regions of Containment As stated in Section 6.2.3.1.7, the CFCS is designed to ensure mixing of the sprayed and unsprayed regions of the containment atmosphere, to improve airborne fission product removal following any size reactor coolant pressure boundary break.6.2-52 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The containment sprayed volume for the spray injection mode is calculated based on the assumption that the unoccupied containment volume above the operating floor is 100 percent covered by sprays. It includes the sprayed volume below the grating in the operating floor deck, and the refueling cavity volume. Based on the above, aApproximately 17.5 percent of the containment free air volume is not reached by the spray. The values listed below were used to estimate the total unsprayed volume in the containment.

(1) Containment radius 70 ft Height between operating deck and spring line 91 ft Approximate deck area covered only by grating 2,990 ft 2 Height between elevation 91 feet and deck 49 ft Average fall height below deck through grating 12 ft (2) These data result in the following volumes: Volume in dome 717,000 ft 3 Volume in cylinder above deck 1,400,000 ft 3 Occupied volume above deck -95,000 ft 3 Sprayed volume below deck 36,000 ft 3 Sprayed refueling cavity volume 45,000 ft 3 Total sprayed volume 2,103,000 ft 3 Total free volume 2,550,000 ft 3 Total unsprayed volume 447,000 ft 3 Percent unsprayed volume -17.5%To justify the assumption that the unoccupied containment volume above the operating floor is 100 percent covered by sprays, the analysis of the containment spray coverage during the spray injection mode with only one CSS train operating estimated the projected unsprayed area percentage of the containment deck area of 42 percent for Unit 1 and 44 percent for Unit 2, using actual spray flow patterns.

The spray reduction factor of 0.5 used to address the spray compression effect due to elevated containment pressure was conservatively based on the containment design pressure.

It was concluded that the whole volume above the operating floor can be considered well mixed and totally subjected to the sprays given the estimated spray coverage calculated above based on actual spray patterns, and the high levels of turbulence due to spray action, entrainment of the air from the volume outside the spray patterns, and induced upflow of air to balance the downflow experienced in the sprayed volume.The effect of the reduction in spray flow rate between the spray injection phase (i.e., 2456 gpm) and the spray recirculation phase (i.e., 1211 gpm) on the containment spray coverage was evaluated.

The minimum volumetric flow rate of water through the spray nozzles, the associated nozzle water pressure drop, and Figure 1 of NUREG/CR-5966, June 1993 (Reference 63), were used to establish the effect of the reduction in spray flow rate on the spray pattern. The analysis determined that with the continued use of a 6.2-53 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE spray reduction factor of 0.5 for spray compression, there was a 3 percent increase in the unsprayed area percentage of the containment deck for both units. It was concluded that the continued use of a spray pattern reduction factor of 0.5 is extremely conservative when applied to the recirculation mode (the containment pressure is substantially lower than the design pressure), and that there is no significant increase in the unsprayed area percentage of the containment deck for both Unit 1 and Unit 2.Thus, a sprayed volume of 82.5 percent of the containment free volume is also applicable for the containment spray recirculation mode.The CFCUs support post-LOCA mixing of the sprayed and unsprayed volume of the containment at a rate higher than that justified by natural convection.

The containment mixing rate between the sprayed and unsprayed regions following a LOCA is determined to be 9.13 turnovers of the unsprayed regions per hour. This mixing rate is based on the operation of two CFCUs, with an assumed total volumetric flow rate of 68,000 cfm between the unsprayed regions and sprayed regions.Review of the locations of the intake and exhaust registers of the operating fan coolers, the major openings in the containment structure, and various active and passive mixing mechanisms, results in the conclusion that following a LOCA, credit can be taken for a)the entire flowrate provided by each operating CFCU to support mixing between the sprayed and unsprayed regions, and b) homogeneous mixing within the sprayed and unsprayed regions, of the volume of air transferred from one region to the other due to CFCU operation.

The of thyroid epue in Section 15.5.17.2.4 were based on the assumptieR Of unifoFrm mixing 4n the full free volume of the containment.

As a result of+ka i~n gI.n af ;~ frm *Ha a nna A natn af thk ,n+ ; an +ra .,r d im +n 4ka IrJ -1"- **~~tbAlff SPrayed areas by the CF-CS, good mixing is provided.

As shown in Table 6.2 26, the fan cooler Unit capacity is 17,000 cfm. if a simplified two volume moidel were used, in combination w ith assumptions that some iodine was available for leakage from the lower containment sectiOn (unsprayed), some reduction in the effective calculated spray-removal coefficaient would result.HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED In rderr to evaluate these ad net fher porSibl combinatons of degraded prrFormance of-the spr-ay system, a senisitivty study was performed to determine the effcct of a redU~ed removal on thyr4o The rFeulfts of this study are presented in F.gures 15.5 6, 15.5 7, and 15.5, .As shown in these figures, both. the sh,.. t-rm and the long term thyroid e sres are ve Ay insensitUv4e to reduced spray remail coffcintdon o alue1&s alow. as 10 h.4 -6.2.3.3.7.2 Fission Product Removal Regulatory Guide 1.183, July 2000, Appendix A, Section 3.3 (Reference 62), invokes the guidance of SRP 6.5.2, Revision 4 (ReferenGee64-and NUREG/ CR-5966, June 1993 (Reference

63) as acceptable models for removal of iodines and particulates.

6.2-54 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE In accordance with the above guidance, the fission product removal coefficients developed for the LOCA are based on the following:

1) Elemental iodine removal coefficients are calculated using guidance provided in SRP 6.5.2 which is invoked by Regulatory Guide 1.183, July 2000, Appendix A, Section 3.3.2) Time dependent particulate aerosol removal coefficients are estimated using guidance provided in Regulatory Guide 1.183, July 2000, Appendix A, Section 3.3, for alternative source terms, and use of computer program SWNAUA. Refer to Section 15.5.8.12 for a description of computer program SWNAUA.The total elemental iodine and particulate removal coefficients in the sprayed and unsprayed region of the containment as a function of time are summarized in Table 6.2-32. The methodology utilized to develop these values is summarized below.6.2.3.3.7.2.1 Elemental Iodine Removal The methodology presented in Section III, 4.C.i, of SRP 6.5.2, Revision 4 is used to estimate the elemental iodine removal coefficients.

The removal of elemental iodine from the containment atmosphere can be attributed due to wall deposition (AE, Wall) and due to the action of containment spray (AE, Spray).Elemental Iodine Removal Coefficients Due to Wall Deposition (AEWall)The elemental iodine removal coefficients due to wall deposition can be estimated using the equation provided in SRP 6.5.2, Revision 4.AE, Wall = K,'A/V Where: Kw = mass transfer coefficient (ft/hr)A = wetted surface area (ft 2)V = volume of the containment (ft 3)Note: K, value of 4.9 m/hr (or 16.08 ft/hr.) conservatively envelopes all available experimental data (SRP 6.5.2, Revision 4 The total containment surface area (435,256 ft 2) is initially available for wall deposition due to condensation on heat sink surfaces prior to spray actuation after accident.Subsequently, due to heat-up, certain portions of the heat sink surfaces become non-condensing and can no longer be considered as "wetted" surfaces.

However, after spray actuation, since the heat sink surfaces in the sprayed region are continuously wetted by sprays, elemental iodine removal due to wall deposition in the sprayed region is valid over the entire period of containment spray operation.

The wetted surface area within the sprayed volume is conservatively assumed to be limited to the carbon steel lined containment shell surface area (90,560 ft 2) multiplied by the spray coverage 6.2-55 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE fraction of containment volume. Therefore, wall deposition elemental iodine removal coefficients are calculated during the period (1) prior to containment spray actuation, applicable to the entire containment, and (2) post containment spray actuation, applicable in the sprayed region for the duration of spray operation.

Elemental Iodine Removal Coefficients Due to Sprays in the Sprayed Region (AESpray)The elemental iodine removal coefficients due to spray actuation can be estimated using the equation provided in Section III, 4.C.i, of SRP 6.5.2.AE, Spray = (6 Kg't-F) /(Vs'd)Note: This equation is valid for 10 hr-1 5 AE, Spray < 20 hr- to prevent extrapolation beyond the existing data for boric acid solution with a pH of 5.Where: Kg = gas phase mass transfer coefficient (ft/hr)t = time of fall of the spray droplets (= h/UT) (hr)F = volumetric spray flow rate (ft 3/hr)V, = effectively sprayed containment free volume (ft 3)d = mass-mean diameter of the spray droplets (fM)h = mean fall height of the spray droplet (ft)UT = terminal velocity of the spray droplet (ft/hr)The gas phase mass transfer coefficient is determined by using the equation provided in Reference 64.Kg = (Dg/d) x 2.0 x (1 + 0.276.Re1/

2 Sc"/3)Where: Dg = diffusivity of iodine in the gas film surrounding the drop (ft 2/hr)Re = dimensionless Reynolds number Sc = dimensionless Schmidt number 6.2.3.3.7.2.2 Particulate Removal There are several aerosol mechanics phenomena that promote the depletion of aerosols from the containment atmosphere.

These include the natural phenomena of particle growth due to agglomeration, gravitational settling of particles (also called gravitational sedimentation), diffuional plate out, diffusiophoresis (particulate removal due to steam condensation), and removal by fluid mechanical interaction with the falling droplets that enter the containment atmosphere through the spray system nozzles (i.e., containment spray).All of the above phenomena are credited for DCPP. Agglomeration of the aerosol is considered in both sprayed and unsprayed regions. In the sprayed region, the particulate removal calculation takes credit for the removal effectiveness of 6.2-56 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE diffusiophoresis and sprays. Gravitational settling is considered only in the unsprayed region.Removal of Particulates by Sprays The particulate removal rate is calculated using computer code SWNAUA. The model correlations implemented into SWNAUA conservatively underestimates the spray removal coefficient.

The spray model incorporated in the SWNAUA code was originally described in Reference

66. When performing DBA calculations to determine particulate removal in the effectively sprayed region of the containment, only the conservatively developed spray removal models and conservative steam condensation rates for the diffusiophoresis calculation are utilized.

While agglomeration is considered in the calculation, its impact on the resulting particulate removal rates is negligible.

In summary, the aerosol removal rates calculated by SWNAUA are conservative lower bound estimates.

The spray model in SWNAUA evaluates the particulate removal efficiency for each particle size in the aerosol by the following mechanisms:

inertial impaction, interception, and Brownian diffusion.

The aerosol removal constant due to spray is presented in NUREG-0772, June 1981 (Reference

67) as: 3 Fm h x Vspray -Vsed 4 Rsp Pw V Vspray Where Aspray -Particulate removal constant for spray, Fm = Spray mass flow rate, h = Spray fall height, E = Collision efficiency, Rsp = Spray droplet radius, Pw = Density of the spray droplet, V = Effectively sprayed volume of containment, Vspray = Velocity of the spray droplets, and Vsed = Aerosol sedimentation velocity.The collision efficiency is divided into three contributing mechanisms as described in BMI-2104 (Reference 68): Where E = Efficiency due to inertial impaction, Er = Efficiency due interception and Ed = Efficiency due to Brownian diffusion.

6.2-57 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE For viscous flow around the spray droplet, the inertial impaction efficiency is given in Reference 67:]0.75 In (2 Stk )L+ Stk --1.214 The critical Stokes number, Stk, for viscous flow is 1.214; for Stk below this value, the model assumes the efficiency of inertial impaction is 0.0. The Stk is calculated from Reference 68: Stk 2 pp r 2 Cc (Vspray -Vsed)9 11 Rsp Where r = Aerosol particle radius, pp Aerosol density, C = Cunningham slip correction factor, p = Gas viscosity.

For droplet sizes typical of nuclear plant spray systems, the data of Walton and Woolcock (Reference

70) show that collision efficiency will be closer to that predicted for potential flow around the droplet. Calvert (Reference

71) fitted this data to the expression:

CStk )2= Stk +0.7 The collision efficiency predicted by this equation is always higher than that predicted by the viscous flow expression given above. The Calvert's fit is employed in this calculation.

As for the remaining constituents of the collision efficiency, the spray model employs an interception efficiency of the form: 2R3 Rp)which is a conservative approximation of the expression given by Reference

68. The efficiency due to Brownian motion is also taken from this report: 6.2-58 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 3 =3.P5p'2 3 Where Pe =DB =kBotz =T =Peclet number 2 VsprayRsp/DB, Aerosol diffusion coefficient kBOIzTB (Fuchs: Reference 69), Boltzmann constant 1.3804 X 1016 erg/K.Temperature, K Reference 69 gives the aerosol mobility, B: B= Cc In most cases, the overall collision efficiency is dominated by inertial impaction, but for small aerosols, Brownian diffusion may become dominant.

The collision efficiency due to inertial impaction increases as the aerosol size is increased, whereas that due to Brownian diffusion increases as the aerosol size decreases.

Removal of Particulates by Diffusiophoresis Particulate matter is entrained in the steam as it flows to the condensation surfaces.This phenomenon is called diffusiophoresis.

Steam is assumed to condense on spray droplets, on the containment fan cooler units, and on heat sinks.The key input parameters used for the particulate removal coefficients are presented in Tables 6.2-33, 6.2-34, 6.2-35, and 6.2-37.6.2.3.3.7-.1 Drop Size itibto The drop size ditribut,,in used in the analytical model is based on data obtained from-.easu.rements of the actual size distributin from the Sp.ao 17-1 3A nozzle for the range Of pressure drops encountcred during operatien of the spray system. A complete analysis of the expected drop s-ize dr iOtrihbutions, including a stati;tical analysis is contained on 5, 6, and 33. The parameterS used in applying these distributions to the calculation of the iodine Fernoeyal coefficient for the DCPP units are gi',ven ;n Tables 6.2 29, 6.2 36, and 6.2 37.6.2.3.3.7.2 Condensation As the spray solution enters the high temperature containment atmosphere, steam will condense on the spray drops. Thea aRmoGunt Of condensation is calculated by an enthalpy balance on the drop.6.2-59 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE mnh + m ,,,'hf (6-2-4-G)wheFe--n and mn' -mass of the drop befor and after condensatio

_____________

-ass Of condensate, lb h nta enthalpy of the drop, Btu~lb hg-, -4hf-= saturation enthalpy of water vapor and liquid, respectively, The in i each drop diamete r*A the distribution is, therefore, by d ' = v )( g )d Vf hfg '.where-Vf -Gpecii voluImeo liquid at saturation, fti4b V -specii volume of the drop before condensation, Wt/ltI 1 fg IlateIL HeatL a: evapo~atio, Ew;:uf_________

h5_ -enthalpy of steam at saturation, Btu/lb-drop diameter before GAconenaRtion, cM d'- drop diam~eter;after condensation, cm The increase in drop size due to condensation is expected to be complete in a few feet of fall forF the majority of drop si-zes in the distribution.

More detailed calculations by ParSly show. that eveni for the largest drops in the diStribution, thermnal equilibrium is reac;nhed in less than half the available drop fall height.6.2.3.3.7.3 Mass Transfcr Model The b deFiye FeFme asic equation for the d fromn a mnaterial ba ral by the spray systi lodine concentation i n the containment atmosphere is Ia R ee g i ' t i e e e i ..R. ...i u, e u , H , .. ..R.M e...t ..fI u i c e~m may be expressed by: dC dt-EF(l 'g -- -jý (. 12'I.....* : ....

• ar ...... h_______________%

I -euritairlueRi ii-ee voumuue, GG___________G9

-ioine concentration in the containn H -iodine partition coefficient, (gm/liter F -spray flowrate, cc/sec ient atmoasphere, gm/cc.~e-,.".,ieFf gas)6.2-60 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The resulting rchange i the drop .ize di-tribution is taken into transfer calcul~ations described below;.consideration in the mass The variable E, is tho efficiency, Which may also be decribed as th frational to saturationL FE -CL2 -CL1 CL. CL1\v.1r,

• 1 wheFe4 CL-- iodine coneRtrFation in the liquid enterin, the dispersed Phase, gm/CGe-iR the liquid leaving the dispersed phase,?lPI +iI II

• I #tI"- eqUllrllm

oin"concenlnteraOion n the iquG, g This aboPrption efficiency is calculated from the time dependent mass transfer model suggested by L. F. Parely (Reference 7).:The- abSOrption efficiency calculated is a function Of dro sie jn the Fremoval constant summation over all drOp size groups: '- EjFH (6.2 !4)..i=1 V c Asfuther diSGUsSiof drop sizi&ULII, trajectores, opalescence mass transfer modeling is presented in References 5, 65, and 33.A 2.3.3.7 A Exneromontal leip fication of Mofidels HISTORICAL INFORMATION

!N ITALICS BELOW NOT REQUIRED TO BE REVISED The ability of the moedel described to give consen'ative estimates of actual spray performan.e was demenstratd i'n test runs m.ade at Oak Ridge National Laborato.y-(ORWL and Battelle Pacific Noah west Laboratory4 The results of these tests-Ref ~e, ,..sh.wn in Figure 6.2 14 for Run A6, verifid that the spray re.oval model used is eonser-vative in a#l cases.6.2.3.3.8 General Design Criterion 38, 1967 -Reliability and Testability of Engineered Safety Features The SAS testing and inspections are enveloped by the inservice inspection (ISI) and inservice testing (IST) programs.

Pressure containing portions of the CSS and SAS are inspected in accordance with ASME BPVC, Section Xl, as required by the Technical 6.2-61 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.3.3.19 Regulatory Guide 1.183, July 2000, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors The steady state elemental iodine removal coefficients, and the time dependent particulate aerosol removal coefficients, are developed in accordance with Regulatory Guide 1.183, July 2000, Appendix A, Section 3.3, and discussed in detail in Section 6.2.3.3.7.2.

6.2.3.3.49-20 Generic Letter 89-10, June 1989 -Safety-Related Motor-Operated Valve Testing and Surveillance The SAS MOVs are subject to the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996, and meet the requirements of the DCPP MOV Program Plan.6.2.3.4 Tests and Inspections HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED The CSS was tested functionally in accordance with written procedures, as outlined in Chapter 14.Spray pump delivered flow and head data were recorded to verify that the containment spray pumps meet design criteria.Spray additive eductor performance data were provided by the manufacturer based on actual tests of a similar eductor. These tests were conducted using a 1.3 specific gravity solution to verify eductor design performance.

Additional manufacturer's tests were run using water so that comparative performance data were available for the two different additive solutions at eductor design conditions.

Eductor performance was checked subsequent to installation into the system. Spray additive flowrates were measured, with resulting rates in the range 31.5 to 38.5 gpm (35 gpm _+/- 10 percent)considered acceptable.

Each containment spray header was tested individually by connecting a source of air to the normally capped flange connection on the spray pump discharge header, shutting the manual spray header isolation valve and opening the air test line isolation valve and the motor-operated spray header isolation valve. Individual nozzles were checked for proper performance by streamers, which indicated unobstructed air flow.The containment sump recirculation mode was tested initially as part of a preoperational flow test under ambient conditions of the safety injection system (SIS). The purpose of the test was to demonstrate the capability of appropriate subsystems to deliver fluid from the containment sump into the reactor coolant system (RCS) in the required time.6,2-64 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Position (5) -The CIS is designed so that the containment setpoint pressure that initiates containment isolation for nonessential penetrations is set to the minimum compatible with normal operating conditions with additional margin to allow for a small pressure transient.

Position (6) -The vacuum/overpressure relief valvesDCPP pu.ge system valves satisfy Branch Technical Position CSB 6-4, September 1975.The containment purge system valves are sealed closed in accordance with SRP 6.2.4, Item 11.6 and 11.14.Position (7) -The containment vent and purge isolation valves close on a high radiation signal.6.2.4.1.16 Generic Letter 89-10, June 1989 -Safety-Related Motor-Operated Valve Testing and Surveillance In the CIS, PG&E Design Class I position-changeable motor-operated valves (MOVs)meet the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996.6.2.4.1.17 Generic Letter 96-06, September 1996 -Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions The CIS is designed to prevent thermally induced overpressurization of isolated water-filled piping sections in containment during design-basis accidents.

6.2.4.2 System Description The CIS includes the mechanical and instrumentation fluid penetrations and associated valves and isolation devices. These penetrations are identified in Figure 6.2-19 and Table 6.2-39. The CIS design uses the following premises: (1) An automatic containment isolation barrier is provided by a closed system, a trip valve, or a check valve.(2) A closed system meets the following requirements:

a) Inside the containment:

1. No mass transfer with either the RCS or the reactor containment interior 2. Has the same safety classification as ESFs (PG&E Design Class I, Quality/Code Class II)6.2-69 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Position (6) -The DCPP purge system .ale ad ,vacuum/overpressure relief valves satisfy the operability criteria set forth in BTP CSB 6-4, 1975. The opening of the 12 inch vacuum/overpressure relief valves is restricted to no more than 50 degrees.The containment purge system valves are sealed closed in accordance with SRP Section 6.2.4, Revision 3, Item 11.6 and 11.14 during Mode 1 through 4 in accordance with the Technical Specifications, therefore satisfying NUREG-0737, November 1980 Item II.E.4.2, Position (6).Position (7) -The containment purge and vent isolation valves are closed automatically by any one of the following:

(1) Phase A containment isolation signal (2) High gaseous or air particulate radioactivity in containment (3) High radiation at the plant vent The containment purge valves are isolated during Modes 1 through 4 in accordance with the Technical Specifications, and are closed automatically by High gaseous or air particulate radioactivity in containment.

The phase A containment isolation signal is no longer credited for automatic isolation of the containment purge valves.6.2.4.4.16 Generic Letter 89-10, June 1989- Safety-Related Motor-Operated Valve Testing and Surveillance The CIS MOVs are subject to the requirements of Generic Letter 89-10, June 1989, and associated Generic Letter 96-05, September 1996, and meet the requirements of the DCPP MOV program.6.2.4.4.17 Generic Letter 96-06, September 1996 -Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions Containment isolation valves LWS-FCV-253 and LWS-FCV-500 (Penetrations 49 and 50, respectively) have both been modified through the addition of a pressure relief hole to prevent overpressurization of the associated isolated section of piping during design-basis accidents to ensure containment integrity is maintained.

All other piping penetrations that are susceptible to overpressurization either have valves whose design prevents overpressurization (air-operated diaphragm valves, solenoid valves, or air-operated globe valves) or have been drained to prevent overpressurization and thereafter maintained drained as appropriate.

6.2.4.5 Tests and Inspections 6.2-82 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE collections that may provide a flammability risk and the need to provide equipment for monitoring hydrogen concentrations in the containment atmosphere.

To fulfill the requirements of 10 CFR 50.44, PG&E credits the use of the Gentaiement-fan cooler system (CFCS and CSS) as the means of containment atmosphere mixing.Refer to Section 6.2.2 for discussion of ether-the CFCS and CSS design bases.HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED The research and development work discussed in more detail in Section 6.2.5.3 substantially reduced the uncertainties in both the expected rates of hydrogen accumulation and the potential exposures that would result from hydrogen control by venting as follows: (1) Research on the corrosion of aluminum and associated hydrogen production rates by Westinghouse (References 12 -15 and 42) has reduced uncertainties on corrosion rates in the expected post-accident environment and allowed a reduction in the expected corrosion rate from the 42 mg/dm 2/hr used in the Unit 2 PSAR to the value shown in Figure 6.2-24.(2) The amounts of aluminum used in the as-built plant have been minimized through materials design specifications.

Zinc is another significant contributor.

The uncertainties in the amounts of hydrogen produced from both have been reduced by itemized accounting (refer to Table 6.2-42).(3) The amounts of hydrogen expected to be produced by the zirconium-water reaction have been reduced by the more stringent limits established on ECCS performance.

(4) Research by the Atomic Energy Commission (AEC) and its contractors (a partial compilation is included in Reference

16) in the context of emergency core cooling system (ECCS) studies has substantially reduced uncertainties in the extent of zirconium-water reactions following a LOCA.(5) Reevaluation of energy generation rates has allowed reduction of hydrogen generation rate from sump radiolysis.

(6) Research on hydrogen yield in the core and sumps by Westinghouse (References 13-15) has reduced uncertainties in these constants.

(7) Refined analysis of the distribution of fission product decay energy (Reference

17) has resulted in more precise values for the fractions of beta and gamma energies absorbed by water.6.2-84 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE CFCS and CSS isare designed to ensure containment atmosphere mixing as a result of any size reactor coolant pressure boundary break to prevent the coalescence of local hydrogen concentrations within containment.

6.2.5.1.7 General Design Criterion 41, 1967 -Engineered Safety Features Performance Capability The CFCS and CSS areis designed to provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill containment atmosphere mixing.6.2.5.1.8 General Design Criterion 42, 1967- Engineered Safety Features Components Capability The CFCS and CSS areis designed so that the capability of each component and system to perform its required function is not impaired by the effects of a LOCA.6.2.5.1.9 General Design Criterion 49,1967 -Containment Design Basis The containment combustible gas control systems are designed so that the containment structure can accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.6.2.5.1.10 General Design Criterion 54, 1971 -Piping Systems Penetrating Containment The piping that is part of the CHPS and hydrogen monitoring system that penetrate containment is provided with leak detection, isolation, redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating this system.The piping is designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.6.2.5.1.11 General Design Criterion 56, 1971 -Primary Containment Isolation Valves The CHPS and hydrogen monitoring system contain valves in piping that penetrate containment and connect directly to the containment atmosphere.

Remote manual isolation valves are provided outside containment and automatic (check) valves are provided inside containment to ensure containment integrity is maintained.

6.2.5.1.12 10 CFR 50.44 -Combustible Gas Control for Nuclear Power Reactors 6.2-86 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE The CFCS and CSS ensures a mixed atmosphere is maintained within containment to prevent high localized concentrations of hydrogen gas accumulation.

The hydrogen monitoring system is designed to be functional, reliable, and capable of continuously measuring the concentration of hydrogen in the containment atmosphere.

6.2.5.1.13 10 CFR 50.49 -Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants CHPS, EHRS and the hydrogen monitoring system components that require environmental qualification (EQ) are qualified to the requirements of 10 CFR 50.49.6.2.5.1.14 10 CFR 50.55a(f)

-Inservice Testing Requirements American Society of Mechanical Engineers (ASME) code components of the CHPS and hydrogen monitoring system are tested to the requirements of 10 CFR 50.55a(f)(4) and 10 CFR 50.55a(f)(5) to the extent practical.

6.2.5.1.15 10 CFR 50.55a(g)

-Inservice Inspection Requirements ASME code components of the CHPS and hydrogen monitoring system are inspected to the requirements of 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(g)(5) to the extent practical.

6.2.5.1.16 Regulatory Guide 1.7, Revision 2, November 1978 -Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident The CFCS and CSS areis designed to provide a mixed atmosphere in containment and thus control combustible gas concentrations without relying on purging of the containment atmosphere following a LOCA. The CFCS and CSS meets the design, quality assurance, redundancy, energy source, and instrumentation requirements for an engineered safety feature. The hydrogen monitoring system provides a means to measure the hydrogen concentration in the containment.

6.2.5.1.17 Regulatory Guide 1.97, Revision 3, May 1983 -Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident The hydrogen monitoring system is designed to provide continuous indication in the control room of hydrogen concentration in the containment atmosphere following a beyond design basis accident and meets the design provisions of Regulatory Guide 1.97, Revision 3, May 1983 including qualification, redundancy and testability.

6.2.5.1.18 NUREG-0737 (Items I1.E.4.1, II.F.1), November 1980- Clarification of TMI Action Plan Requirements 6.2-87 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 100 eV or less. Regulatory Guide 1.7, Revision 2, November, 1978 does not, however, allow credit for the reduced hydrogen yields and a yield value of 0.5 molecules per 100 eV is used in the analyses.All containment volumes are connected by large vent areas to promote good air circulation.

Hydrogen will diffuse very rapidly giving an even distribution under the conditions existing in the containment structure.

In addition, thermal mixing effects, heating of air above the hot sump water, and possible steam released from the RCS will move the hydrogen-laden air from the points of generation toward the cool external walls. Although hydrogen is lighter than air, it will not concentrate significantly in high areas because of the high diffusion rate, the open design of the containment, and the fan cooler air mixing.The ability of hydrogen to diffuse rapidly into all volumes is inferred from a CSE experiment (Reference 23). These tests showed very good mixing in the main chamber and a rapid interchange by diffusion and mixing with the atmosphere of other chambers that had limited communication.

The diffusivity of hydrogen is approximately 10 times that of iodine, so a more uniform mixture is expected for hydrogen.

Also, higher concentration provides greater concentration gradients for better diffusion than indicated by the CSE tests.Table 6.2-45 summarizes the calculated hydrogen production and accumulation data.6.2.5.3.6.1.5 Results of the Hydrogen Generation and Accumulation Analyses The results of the hydrogen generation and accumulation analyses are presented in Figures 6.2-26 through 6.2-29.6.2.5.3.7 General Design Criterion 41, 1967 -Engineered Safety Features Performance Capability The CFCS, including required auxiliary systems, is designed to tolerate a single active failure following a LOCA without loss of protective function.

Refer to Section 6.2.2.3.2 for further discussion.

6.2.5.3.8 General Design Criterion 42, 1967 -Engineered Safety Features Components Capability The component design pressure and temperature conditions in Table 6.2-26 are specified as the most severe conditions to which each CFCS and CSS component is exposed during either normal or post-LOCA operation allowing the system to provide a mixed atmosphere in post-LOCA conditions.

Refer to Section 6.2.2.3.1 for further discussion.

6.2.5.3.9 General Design Criterion 49,1967 -Containment Design Basis 6.2-98 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 6.2.5.3.14 10 CFR 50.55a(f)

-Inservice Testing Requirements Periodic inservice testing (IST) of all containment isolation valves, in the system is performed.

The IST requirements are contained within the IST Program Plan and comply with the ASME code for Operation and Maintenance of Nuclear Power Plants.Refer to Section 6.2.2.4 for a discussion of testing of the CFCS.6.2.5.3.15 10 CFR 50.55a(g)-

Inservice Inspection Requirements The inservice inspection (ISI) requirements for the CHPS and containment hydrogen monitoring penetrations are contained within the ISI Program Plan and comply with the ASME BPVC,Section XI. Refer to Section 6.2.2.4 for a discussion of inspection of the CFCS and CSS.6.2.5.3.16 Regulatory Guide 1.7, Revision 2, November 1978 -Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident The CFCS and CSS, which servesewiig as the credited means for containment atmosphere mixing in accordance with 10 CFR 50.44, is-are designed and constructed to PG&E Design Class I standards.

The hydrogen monitoring system includes two hydrogen monitors to measure the hydrogen concentration in the containment.

The hydrogen monitoring system was originally designed and constructed as Category 1, as defined by Regulatory Guide 1.97, Revision 2, December 1980 however has been reclassified as Category 3, as defined by Regulatory Guide 1.97, Revision 3, May 1983 as a result of the rulemaking revision to 10 CFR 50.44 (refer to Section 6.2.5.3.17).

The EHRS is available to control containment hydrogen concentration following a LOCA to at or below 4.0 percent by volume without relying on the CHPS. Each of the two redundant recombiners is capable of providing the required removal capacity.

The CHPS is also available.

To ensure that the lower flammability limit (4 percent) will not be exceeded, the internal electric hydrogen recombiners will be started at or below 3.5 percent by volume.The EHRS and CHPS systems meet PG&E Design Class I design and construction standards.

The EHRS provides 100 percent redundancy since each recombiner and its associated power supply and control panel are capable of providing the required hydrogen removal capacity.

The second unit, including its associated power supply and control panels, is normally on standby following a postulated LOCA.6.2-100 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE Refer to Sections 6.2.5.3.3, 6.2.5.3.16, and 6.2.5.3.17 for additional discussion on the hydrogen monitors.6.2.5.5.2 Containment Hydrogen Purge System Instrumentation is provided to monitor the flowrate and the amount of radioactivity released by the purging operation.

The containment radiation monitoring system and the plant vent radiation monitors are used to monitor the radioactivity in containment and the hydrogen purge line. Refer to Sections 9.4.2 and 11.4 for information regarding the plant vent.A manual sample point is provided on each exhaust line to obtain a grab sample for laboratory analysis.Flow indicators are provided for each CHPS exhaust line. The indicators are PG&E Design Class I. The range is 500 to 4000 feet per minute (corresponding to flowrates of approximately 45 to 350 cfm).Containment isolation valves status is shown on the main control board as indicated in Table 6.2-39. Annunciation is provided to alarm on high radioactivity, high flowrate, and fan failure.Refer to Sections 6.2.5.3.3 and 6.2.5.3.4 for additional discussion on the CHPS instrumentation.

6.2.5.6 Materials Materials of construction of components are indicated in Section 6.2.5.2.6.

2.6 REFERENCES

1. Deleted in Revision 18.2. Deleted in Revision 18.3. Deleted in Revision 18.4. M. A. Styrikovich, et al, Atomnoya Energya, Vol. 17, No. 1, July 1964, pp. 45-49, (Translation in UD 621.039.562.5).

5. R. M. Kemper, Iodine Removal by Spray in the Diablo Canyon Station Containment, WCAP-7977, September 1973.6. DeletedW.

F. Pasedag and j. L. Gallaqher, "Drop Size Distribution and Spray E# v. n .N ,, i , 4', 1 '71, ,- 412 6.2-103 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 7. DeletedL.

F. -,,-,h, n ., f ;,pn4 .; e ..Dp,,,-,,,n of, .... ,na m....ORNII TM 2412, Part VII, 1970.8. A.E.J. Eggleton, A Theor-etica!

Examination of Iodine W~ater Partition Gpopeflcpint, AERE, (R) 4887, !967Deleted.

9. Preliminary Safety Analysis Report, Nuclear Unit Number 2, Diablo Canyon Site, Pacific Gas and Electric Company. Supplements Number 3 and 5.10. Safety Evaluation by the Division of Reactor Licensing (USAEC) in the Matter of Pacific Gas and Electric Company Diablo Canyon Nuclear Power Plant Unit 2 Docket No. 50-323, November 18, 1969.11. Report on Pacific Gas and Electric Company Nuclear Unit 2 -Diablo Canyon Site, Advisory Committee on Reactor Safeguard, USAEC, October 16, 1969.12. M. J. Bell, et a!, inv.estigation of Chemic~al Additives for ReactOr Containmcnt-SrsWestinghouse Eleti Corporation, WCAP 7826, Decemnber-4971-Deleted.

13. VV. K. Brunot, et a!, Control of the H4Ydrogen ConGentration Felfowinwa-

.. -7372, November ..96Deleted.

14. W. D. Fletcher, et al, "Post-LOCA Hydrogen Generation in PWR Containments," Journal of the American Nuclear Society, June 1970.15. W. D. Fletcher, et al, "Post-LOCA Hydrogen Generation in PWR Containments", Nuclear Technology 10, 1971, pp. 420-427, 1971.16. Draft Environmental Statement Concerning Proposed Rule-Makinq Action-Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactor, USAEC Regulatory Staff, December 1972.W.estinghuse E.lectr-ic Corpation, \A" CAP 7816, December 1971Deleted.

18. M. L. Mooney and H. E. Cramer, Study of the Doablo C;an.o NucleaFr PDe. Plant Sit e, Pacific Gas and lect Company, May l970Deleted.

19. M. L. Mooney, Supplement No. 1 to MeteGFro~ow Repot for Diablo Canynmie..

May 1974Deleted.

20. M. L. Mooney, Supplement No. 2 to-MeteoologY Report for Doable Canyon Site juine 1972Deleted.

6.2-104 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 21. V. K. EMERALD A D-.,,,. f the,,,, ,-,o A,-;,,;f,, Releases and Potential Doese from a Prgsgurized Water Rcac-tor PlantB, Pacifi Gar, and Company, October 1971 Deleted.22. Regulatory Guide 1.7, Rev. 2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, USNRC, November 1978.23. H. E. Zittel, Radiation and Thermal Stability of Spray Solutions, ORNL-NSRD Program Bi-Monthly Report for May-June, 1969. ORNL-TM-2663, September, 1969.24. j. J. DiNunno, et al, C,,,Ulation of Distance FPactor for Power and Sites, AEC Repot Number TOD 11811, March 23, 1962.Deleted

25. W. i3 P,., ll, tQ NLI r,..,r U,-Jfp, DS......y

.-e.,n l....Ip- n-- Dwe nt .. BL Monthly Reort forJuly August 1@68I ORNL TM 2368, November l96 Deleted.26. W.^1. I G ORNLM NUkl, Sa.'feoty

..'eeaF-h .-revele-ome.

t f a Dr'rm Monthly RePo- for FSeptember October, 1968, rRNL TM 2425, y 1969, p=.-53Deleted.

27. Deleted in Revision 18.28. H. E. Zittel T. H. ROw, and Thermal Stabilit' of Spra' SQolutiQons" Nuclear Techn..ogy, 10, 1971, pp. 136 43 Deleted.29. A. 0. Allen, The Radiation Chemistr, of Water and Aguepus Solutions,N. j., Van No-strand, 196 Deleted.30. Deleted in Revision 18.31. Deleted in Revision 18.32. Deleted in Revision 18.33. DeletedR.

M .K,,,p,,, d ...... SOW in the S.,,;,-containment, WCAP 7952, August192 34. Deleted in Revision 18.35. Deleted in Revision 18.36. G. A. Israelson, J. R. van Searen, W. C. Boettinger, Reactor Containment Fan Cooler Coolinq Test Coil, WCAP-7336-L, 1969.37. Deleted in Revision 18.6.2-105 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 38. Diablo Canyon Power Plant -Inservice Inspection Program Plan -The Third 10-Year Inspection Interval, Pacific Gas and Electric Company.39. EQ File IH-05, Containment Fan Cooler Motor, Pacific Gas and Electric Company.40. Standard Review Plan 6.2.5, Combustible Gas Control.41. Branch Technical Position CSB 6-2, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident.42. D. D Whyte., R. C. Burchell, Corrosion Study for Determining Hydrogen Generating From Aluminum and Zinc During Postaccident Conditions, WCAP-8776, April 1976.43. Deleted in Revision 18.44. Deleted in Revision 18.45. F. M. Bordelon, Analysis of the Transient Flow Distribution During Blowdown (TMD Code) WCAP-7548, 1970.46. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.47. IEEE-Std-334, Guide for Type Tests of Class I Motors Installed Inside the Containment of Nuclear Power Generating Stations, 1971.48. AEC Safety Guide 11, Instrument Lines Penetrating Primary Reactor Containment, March 10, 1971.49. Deleted in Revision 18.50. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment Spray Flow Reduction, February 7, 1991.51. Westinghouse Letter PGE-89-673, RWST Setpoint Evaluation, July 24, 1989.52. Deleted in Revision 18.53. Deleted in Revision 18.54. Deleted in Revision 18.55. J. C. A. L. Bacarella, Design C.nside.ations of ReactOr rnta, mcR. t Snrt, Q*,ptnrc, D ,nr4e III Tha C'nrrne;n nf I r;,Ip ;.n Qnrin Qr*0 I ,4nlmt ORNL T -. ,- ...l, Ocember l69Deleted.

56. Deleted in Revision 18.6.2-106 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 57. Deleted in Revision 18.58. Deleted in Revision 18.59. Westinghouse Mass and Energy Release Data for Containment Design, WCAP-8264-P-A, Rev. 1, (Proprietary), WCAP-8312-A, August 1975.60. Ice Condenser Containment Pressure Transient Analysis Methods, WCAP-8077 (Proprietary), WCAP-8078, March 1973.61. NUREG-0800, Standard Review Plan, 6.5.2, Revision 4 Containment Spray as a Fission Product Cleanup System, Revision 4.62. Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision OJuly 2000.63. NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays, June 1993.64. W. M. Rohsenow and Harry Choi, Heat, Mass and Momentum Transfer, Prentice-Hall 1961 65. Bunz, H., Kayro, M., Schock, W., NAUA/MOD4

-A Code for Calculating Aerosol Behaviour in LWR Core Melt Accidents, Code Description and User Manual, KFK, 1982.66. Elia, Frank A. Jr. and Lischer, D. Jeffrey, Advanced Method for Calculating the Removal of Airborne Particles with Sprays, ASME Paper No. 93-WA/SERA-5, 1993.67. NUREG-0772, Technical Bases for Estimating Fission Product Behavior During LWR Accidents, 1981.68. Battelle Columbus Laboratories, BMI 2104, Vol. III, draft report, Radionuclide Release Under Specific LWR Accident Conditions, 1984.69. Fuchs, N.A., The Mechanics of Aerosols, revised and enlarged edition, Dover Publications, Inc.,1964.

70. Walton, W. H., and Woolcock, A., The Suppression of Airborne Dust by Water Spray International Journal of Air Pollution 3, 129-153, 1960.71. Calvert, S., Venturi and Other Atomizing Scrubbers Efficiency and Pressure Drop, AlChE Journal 16, 392-396, 1970.6.2-107 Revision 21 September 2013 DCPP UNITS 1 & 2 FSAR UPDATE 72. NUREG-0800, Standard Review Plan, Section 6.1.1, "Engineered Safety Features Materials", Revision 2 60.73.NUREG/CR-5732, "Iodine Chemical Forms in LWR Severe Accidents

-Final Report," April 1992.6.2.7 REFERENCE DRAWINGS Figures representing controlled engineering drawings are incorporated by reference and are identified in Table 1.6-1. The contents of the drawings are controlled by DCPP procedures.

6.2-108 Revision 21 September 2013 DCPP Units 1 & 2 FSAR UPDATE TABLE 6.2-39 CONTAINMENT PIPING PENETRATIONS AND VALVING Sheet 14 of 24 Appli-Pentr System cable No. (Safety GDC Priority)(Note 33)0DC Confor- Figure Viv Valve ID Number mance 6.2-19 Lb (Note 51)Sheet No.Used Long Term Valve Operator Cntmt Coopol Normal Po-er Trip After Post-Type Type Locat. PG&E Room Position Fail. On LOCA LOCA Fluid Temp (Note 30) (Note 31) (Note 32) Piping Indication (Note 35) Position (Note (Note 36) Position (Note (Note Notes Group 25) (Note 37) 23) 24)59HI Conairce P51 Lresurme Tranomitc PT-933 &PT-935 (SA)56 Yes tIII A Scaled losnuceont B Sealed Bellows Sin Sbl W Cold 16 W Cold 16 60 Mcu-Equi N., a piping penetrati-o HatcN (NE)61 Cortaacr Sup~ply 62 Cntainoc ent 63 Ccccainm ent a-d ONE)64 FueI Tule (NE)65 Persocnl-Hatch (NT)66 Emer-gen Yes 18(11)Yes 18 (1w0 Yes 18(1)D VAC-FCV-660 E VAC-FCV-661 F VAC-RCV-11 G VAC-RCV-12 A VAC-FCV-662 B VAC-FCV-663 C VAC-FCV-664 J Speaccclc Flange But Air I AE Yes C Closed T N C G Cold 4856 But Air 0 AE Yes C Closed T N C G Cold 4-56 But Air I AE Yes C Closed T N C G Cold t856 But Air 0 AE Yes C Closed T N C G Cold t456 56 But But But Spf Air Air Air I 0 0 0 A A A A Yes Yes Yes C C C NIA Closed T Closed T Closed T N N N N/A C C NIA G Cold 18 G Cold 18 G Cold 18 G Cold 26-I -26 3 (11)Not a piping pemnetrtioc Not a piping prenvation NHr a piping penetation