ML16169A118

RWA-1313-015, Rev. 1, AST Radiological Analysis Technical Report.

ML16169A118
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Issue date:	03/31/2016
From:	Michael Pope Red Wolf Associates, PLLC
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AEP-NRC-2016-23 RWA-1313-015, Rev 1
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Enclosure 5 to AEP-NRC-2016-23 Red Wolf Associates (RWA) Technical Report RWA-1313-015, Rev. 1, "D. C. Cook AST Radiological Analyses Technical Report"

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D. C. Cook AST Radiological Analyses Technical Report Page 1 of 75 Report Number: RWA-1313-015 Revision: 1 I

D. C. Cook AST Radiological Analyses Technical Report Preparer: Mark Pope Date:

Digitally signed by Mark Pope Date:2016.03.31 15:51:44 EDT Reviewer: Katie Beasley Date:

Digitally signed by Katie Beasley Date: 2016.03.31 16:21:03 EDT Approver: Joe Sinodis Date:

Digitally signed by Joe Sinodis Date:2016.03.31 16:26:33 EDT

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 2 of75 Table of Contents 1 Introduction ........................................................................................................................................... 5 1.1 Overview ....................................................................................................................................... 5 1.2 Proposed Changes to the Licensing Basis ..................................................................................... 6 1.3 Computer Codes ............................................................................................................................ 6 2 Common Input Parameters, Assumptions, and Methods ...................................................................... 7 2.1 Control Room Dose Calculation Model... .............................................................................. ,...... 7 2.2 Source Terms ................................................................................................................................. 8 2.2. l Core Source Term .................................................................................. ,.............................. 8 2.2.2 Fuel Handling Accident Source Term ................................................................................... 9 2.2.3 RCS Source Term ................................................................................................................. 9 2.2.4 Steam Generator Secondary Source Term .......................................................................... 10 2.2.5 Fuel Rod Gap Fractions and High Burnup Rods ................................................................ 15 2.3 Atmospheric Dispersion Factors ................................................................................................. 16 2.3.1 Onsite.:\/QDetermination................................................................................................... 16 2.3 .2 Offsite .:\/Q Determination .................................................................................................. 17 2.3.3 Meteorological Data ............................................................................................................ 18 2.4 Direct Shine Dose***********************************************.************************************************************************ 37 3 Event Analyses .................................................................................................................................... 38 3.1 Loss of Coolant Accident. ........................................................................................................... 38 3.1.1 Containment Purge .............................................................................................................. 38 3.1.2 Containment Leakage ......................................................................................................... 39 3.1.3 ESF Leakage into the Auxiliary Building ........................................................................... 40 3.1.4 ESF Leakage into the RWST .............................................................................................. 40 3.2 Fuel Handling Accident .............................................................................................................. 51 .

3.3 Main Steam Line Break .............................................................................................................. 53 3.4 Steam Generator Tube Rupture ................................................................................................... 57

• 3.5 * .Locked Rotor .............................................................................................................................. 61 3.6 . Control Rod Ejection .................................................................................................................. 63

• 3.7 Waste Gas Decay Tank Rupture ....... ,......................................................................................... 67 3.8 Volume Control Tank Rupture ................................................................................................... 70 3.9 Results Summary ........................................................................................................................ 73 4 References ........................................................................................................................................... 74 List of Figures Figure 2.3-1: EAB Downwind Sectors ....................................................................................................... 19 Figure 2.3-2: LPZ Downwind Sectors ........................................................................................................ 20 Figure 2.3-3: Site Orientation ..................................................................................................................... 22 Figure 2.3-4: Onsite Release-Receptor Locations ...................................................................................... 23 Figure 2.3-5: Offsite Release-Receptor Locations ...................................................................................... 24

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~~~oci:~~ D. C. Cook AST Radiological Analyses Technical Report Page 3 of 75 List of Tables Table 2.1-1: Control Room Parameters ........................................................................................................ 8 Table 2.2-1: Source Term Inputs and Assumptions .................................................................................... 10 Table 2.2-2: Core Source Term .................................................................................................................. 11 Table 2.2-3: Fuel Handling Accident Source Term .................................................................................... 12 Table 2.2-4: RCS Source Term ............................................................................*....................................... 14 Table 2.2-5: SG Secondary Source Term********:****************************************************************************************** 15

• Table 2.2-6: Non-LOCA Fuel Rod Gap Inventory Fraction ....................................................................... 16 Table 2.3-1: Meteorological Data Recovery Rate ...................................................................................... 21 Table 2.3-2: Onsite Release-Receptor Combination Parameters ................................................................ 25 Table 2.3-3: Onsite Atmospheric Dispersion Factors ................................................................................. 26 Table 2.3-4: Offsite Release-Receptor Distances ....................................................................................... 27 Table 2.3-5: Downwind Direction- Met Data Correlation .......................................................................... 34 Table 2.3-6: Offsite Atmospheric Dispersion Factors ................................................................................ 34 Table 2.3-7: Release~Receptor Pairs Application to the Event Analyses ................................................... 36 Table 2.3-8: Offsite.Breathing Rates .......................................................................................................... 37 Table 2.4-1: LOCA Direct Shine Dose ....................................................................................................... 3 8 Table 3.1-1: Core Inventory Fraction Release into Containment ............................................................... 39 Table 3.1-2: LOCA Release Phase Timing ................................................................................................. 39 Table 3. l -3: LOCA Inputs and Assumptions .............................................................................................. 41 Table 3. l-4: RWST Liquid Iodine Concentration ....................................................................................... 45 Table 3.1-5: RWST pH ...................................................................................... :........................................ 46 Table 3.1-6: Elemental Iodine Release Fraction ......................................................................................... 47 Table 3.1-7: RWST Temperature Profile .................................................................................................... 48 Table 3.1-8: RWST 12 Partition Coefficient.. .............................................................................................. 49 Table 3.1-9: Adjusted RWST Iodine Release Rate ..................................................................................... 50 Table 3.1-10: LOCA TEDE Dose Results .................................................................................................. 50 Table 3 .2-1: Fuel Handling Inputs and Assumptions .................................................................................. 51 Table 3 .2-2: Fuel Handling Accident - Containment Release TEDE Dose Results ................................... 52 Table 3 .2-3: Fuel Handling Accident - Auxiliary Building Release TEDE Dose Results ......................... 53 Table 3 .3-1: Main Steam Line Break Iodine Appearance Rate Inputs and Assumptions ........................... 54 Table 3.3-2: Main Steam Line Break 500x Iodine Appearance .................................................................. 54 Table 3.3-3: Main Steam Line Break Inputs and Assumptions .................................................................. 55 Table 3.3-4: Main Steam Line Break Pre-Accident Spike TEDE Dose Results ......................................... 56 Table 3.3-5: Main Steam Line Break Concurrent Spike TEDE Dose Results ............................................ 56 Table 3.4-1: SGTR Iodine Appearance Rate Inputs and Assumptions ....................................................... 58 Table 3.4-2: SGTR 335x Iodine Appearance .............................................................................................. 58 Table 3.4-3: Steam Generator Tube Rupture Inputs and Assumptions ....................................................... 59 Table 3.4-4: Steam Generator Tube Rupture Pre-Accident Spike TEDE Dose Results ............................. 60 Table 3.4-5: Steam Generator Tube Rupture Concurrent Spike TEDE Dose Results ................................ 61 Table 3.5-1: Locked Rotor Inputs and Assumptions .................................................................................. 62

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page4 of 75 Table 3.5-2: Locked Rotor TEDE Dose Results ......................................................................................... 63 Table 3.6-1: Control Rod Ejection Inputs and Assumptions ...................................................................... 65 Table.3.6-2: Control Rod Ejection Secondary Release TEDE Dose Results .............................................. 67 Table 3.6-3: Control Rod Ejection Containment Release TEDE Dose Results .......................................... 67 Table 3.7-1: WGDT Source Term .............................................................................................................. 68 Table 3.7-2: WGDT Rupture Inputs and Assumptions ................................................................................ 69 Table 3.7-3: WGDT Rupture Dose Results ................................................................................................ 69 Table 3.8-1: VCT Source Term .................................................................................................................. 71 Table 3.8-2: VCT Rupture Inputs and Assumptions ................................................................................... 71 Table 3.8-3: VCT Rupture Dose Results .................................................................................................... 72 Table 3.9-1: Dose Results Summary ............................................................................................................ 73

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 5 of 75 1 Introduction The current D. C. Cook licensing basis for the radiological consequences analyses of accidents discussed in Chapter 14 of the Updated Final Safety Analysis Report (UJ?SAR) is based on methodologies prescribed in 1 Reg. Guide 1.195 for the offsite doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ), and are based upon the Alternative Source Term (AST) methodology from Regulatory Guide 1.183 for the control room doses. These analyses are being updated to fully implement the AST methodology for both the onsite and offsite dose locations.

1.1 Overview The revised design basis dose analyses are performed using the direction of Reg. Guide 1.183 (Reference

[4.1]) with additional guidance provided in Regulatory Issues Summary (RIS) 2006-04 (Reference [4.2]). This single set of analyses is. applicable to both Units 1 and 2, and as such, a limiting set of inputs are applied that is bounding for both units. In addition, releases from either unit must consider the dose impact on all receptor locations applicable to both units.

The following UFSAR Chapter 14 accidents are evaluated:

Loss-of-Coolant Accident (LOCA)

• Fuel Handling Accident (FHA)

Main Steam Line Break (MSLB)

• Steam Generator Tube Rupture (SGTR)

Locked Rotor

• Control Rod Ejection (CRE)

• Waste Gas Decay Tank (WGDT) Rupture

• Volume Control Tank (VCT) Rupture It is important to note that Reg. Guide 1.183 does not include guidance for the analysis of the \VGDT and VCT rupture accidents. These events are evaluated for completeness to apply a consistent source term to all of the radiological consequences presented in the D. C. Cook UFSAR and to assess the control room TEDE doses for these events against the acceptance criteria provided in 10CFR50.67. However, the offsite dose consequences resulting from the WGDT and VCT rupture events will continue to satisfy the acceptance criteria of 10CFR20

• as discussed in Item 11 of Reference [4.2].

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D. C. Cook AST Radiological Analyses Technical Report Page 6 of 75 1.2 Proposed Changes to the Licensing Basis As part of the full implementation of AST and the update of the D. C. Cook radiological dose analysis, the following changes to the Technical Specifications will be proposed.

The definition of Dose Equivalent I-131 in Section 1.1 is revised to reference Federal Guidance Report No.

11 as the source of effective dose conversion factors.

• Section 1.1 is revised to replace the definition ofE Average Disintegration Energy with the definition of Dose Equivalent Xe-133 using dose conversion factors from the effective column of Table III.1 of Federal Guidance Report No. 12.

The Limiting Condition for Operation related to RCS Activity in Section 3.4.16 is modified to replace the 100/E gross specific activity criterion with the Dose Equivalent Xe-133 limit.

• The accident induced leakage performance criterion established by the Steam Generator Program in Section 5.5.7.b.2 is revised to be 1 gpm for all steam generators and 0.25 gpm to any one steam generator.

The maximum allowable leakage rate, La at Pa specified by the Containment Leakage Rate Program in Section 5.5.14.c is reduced to 0.18%/day.

The maximum allowable inethyl iodide penetration for the Control Room Emergency Ventilation charcoal adsorber is increased to 2.5% in Section 5.5.9.c.

1.3 Computer Codes The following computer codes are used in performing the Cook radiological dose analyses:

Computer Code Version Reference RADTRAD 3.10 [4.3] - [4.6]

ARCON96 1997 [4.7]

PAVAN 2.0 [4.8]

JFREQ (METD) 1982 [4.29]

MicroShield 8.03 [4.10]

ORIGEN-ARP 6.1.3 [4.9]

GOTHIC 7.2a [4.11] - [4.13]

RADTRAD is used to determine the control room and offsite doses for each analyzed event using the source term andX/Q inputs. The code considers the release timing, filtration, hold-up, and chemical form of the nuclides released into the environment.

ARCON96 is used to determine the atmospheric dispersion factors (X/Qs) at the control room intakes for selected release locations from plant meteorological data.

PAVAN provides atmospheric dispersion factors (X!Qs) for various time periods at the EAB and LPZ boundaries using plant meteorological data.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 7 of 75 JFREQ is a program in the METD suite of programs that is used to compute thejoint frequency distribution of wind speed; wind direction, and atmospheric stability class for use as input to the PAVAN program.

MicroShield is used to determine the direct shine dose to the operators in the control room from the activity on the control room ventilation system filters.

ORIGEN-ARP calculates the fission product isotopic activity of the reactor core used in the development of the core and RCS source terms.

The GOTHIC code is used to simulate the RCS purification system to determine the relative concentrations of nuclides in the reactor coolant, and is also used to calculate the time-dependent RWST temperature due to backleakage from the containment sump.

• 2 Common Input Parameters, Assumptions, and Methods 2.1 Control Room Dose Calculation Model During normal operation, 880 cfm of unfiltered air enters the control room through the normal outside air intake. Following a safety injection signal, the control room ventilation system is automatically placed into recirculation after applicable delays for signal processing, emergency power restoration, and damper repositioning. In this configuration, the control room pressurization/cleanup fans circulate 5400 cfm of air through the control room filters, with 880 cfm ofthis flow supplied by fresh air from the emergency outdoor air intake and the remaining 4520 cfm taken from the control room envelope.

Unfiltered inleakage is assumed to enter the control room at a constant rate of 40 cfm during all modes of system operation. When the control room ventilation system is aligned in the pressurization/cle~up mode, the control room envelope is at a positive pressure with respect to the surrounding areas and leakage is predominantly out of the control room. However, this flow configuration creates a negative pressure in the system ducting downstream of the isolated normal intake dampers. Therefore, the control room unfiltered inleakage is assumed to enter the control room at the location of the normal intakes. Control room occupancy and breatlµng rates are taken from Position 4.2.6 of Reference [4.1]. The control room model parameters used in the analyses are listed in Table 2.1-1.

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~~i;O'-,i_atP.J;i D. C. Cook AST Radiological Analyses Technical Report Page 8 of 75 Table 2.1-1: Control Room Parameters Parameter Value Control Room Volume 50,616 ft3 Normal Operation Filtered Make-up Flow Rate 0 cfm Filtered Recirculation Flow Rate Ocfm Unfiltered Make-up Flow Rate 880 cfm Unfiltered Inleakage 40cfm Emergency Operation Recirculation Mode:

Filtered Make-up Flow Rate 880 cfm Filtered Recirculation Flow Rate 4520 cfm Unfiltered Make-up Flow Rate Ocfm Unfiltered Inleakage 40 cfm Filter Efficiencies Elemental 94.05%

Organic 94.05%

Particulate 98.01%

Occupancy 0-24 hrs 1.0 1-4 days 0.6 4-30 days 0.4 Breathing Rate 3.5 x 10-4 m3/sec 2.2 Source Terms 2.2.1 Core Source Term Consistent with the guidance of Position 3.1 of Reference [4.1], the inventory of fission products in the reactor core available for release is based upon the product of the maximum full power operation of the core at the licensed rated thermal power and the ECCS evaluation uncertainty. An ORIGEN-ARP model is created using a single fuel assembly as the mass basis with an assembly burnup selected which conservatively exceeds the expected end-of-cycle core average burnup. Separate ORIGEN-ARP cases are run covering the full range of licensed fuel enrichments, and the maximum activity for each dose-significant isotope is selected from among

RWA-1313-015, Rev~ 1 D. C. Cook AST Radiological Analyses Technical Report Page 9 of 75 these cases. In this manner, the fission product activity of a single assembly is derived which bounds the anticipated core design values for enrichment and bumup. The total core average source term is then obtained by multiplying the single-assembly isotopic activities by the number of fuel assemblies in the core. Key source term inputs are presented in Table 2.2-1.

The list of dose significant isotopes is generally based upon Table 5 of Reference [4.1 ], which identifies the radionuclide elements that should be considered in the design basis AST dose analysis. This element list is the same as that specified for the revised source term in Table 3.8 ofNUREG-1465 (Reference [4.14]). Section 1.4.3 .2 of Reference [4.4] documents the creation of a 60-isotope, 9-element list of radionuclides which meets the requirements ofNUREG-1465. However, some early AST submittals included a more comprehensive list of isotopes. Specifically, Table 3.1-4 of Reference [4.15] presents a core inventory which contains the standard 60 isotopes from Table 1.4.3.2-2 of Reference [4.4] plus an additional 61 nuclides. This extended list serves as the refined basis for the dose-significant isotopes selected for this analysis; however, the number of isotopes is reduced to a maximum of 100 to accommodate the limitations of RADTRAD 3 .10 identified in Section 1.1 of Reference [4.3]. The reduction from 121to100 nuclides is based upon the availability of valid dose conversion factors from References [4.16] and [4.17], nuclide half life, and significance of daughter products. The final list of the 100 dose significant isotopes and corresponding core source term activities is presented i!} Table 2.2-2. This source term is used in combination with dose conversion factors from Federal Guidance Report No. 11 (FGR 11) (Reference [4.16]) and Federal Guidance Report No. 12 (FGR 12)

(Reference [4.17]), which is consistent with Position 4.1 of Reference [4.1].

2.2.2 Fuel Handling Accident Source Term The fuel handling accident source term is developed from the core source term and follows the guidance of Position 3 .1 of Reference [4.1 ], which states that the fission product inventory of each damaged fuel rod for DBA events that do not involve the entire core is determined by dividing the total core inventory by the number of rods in the core. To account for differences in power level across the core, the radial peaking factor is applied to the inventory of the damaged rods. Since the fuel handling accident event involves the failure of all of the rods in a single fuel assembly, the FHA source term is derived by dividing each nuclide activity in the core source term by the number of fuel assemblies and multiplying by the radial peaking factor.

Two additional adjustments are needed to finalize the FHA source term. The gap inventory fractions for 1-131 and Kr-85 specified in Table 3 of Reference [4.1] are greater than the respective halogen and noble gas gap fractions. To accommodate these differences, the activities of these two isotopes are increased so that the proper quantities are released when the group gap fractions are applied. Specifically, the activity of 1-131 is increased by a factor of 0.08/0.05 = 1.6, and the activity ofKr-85 is increased by 0.1/0.05 = 2.0.* Thus, using the core average activities from Table 2.2-2, the radial peaking factor of 1.65 from Table 2.2-1, and 193 fuel assemblies in the core, the fuel handling source term shown in Table 2.2-3 is obtained.

2.2.3 RCS Source Term The equilibrium nuclide concentration in the RCS is calculated based on the core inventory described in Section 2.2.1. The rate of nuclide release from the core to the reactor coolant for applicable isotopes is calculated from fission product escape rate coefficients and assumes that 1% of fuel rods have defects. With this isotopic production rate in the coolant established, RCS concentrations are calculated with a hydraulic

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D. C. Cook AST Radiological Analyses Technical Report Page 10 of 75 model of the RCS purification system using GOTHIC. This model accounts for radioactive decay and daughter production, removal of nuclides by the demineralizers, degassing in the volume control tank (VCT),

and dilution of the nuclide concentration by normal makeup for RCS boron control. The GOTHIC model is .

run until the radionuclide concentrations reach equilibrium values. The GOTHIC output provides the relative .

distribution of isotopic concentrations in the RCS. These values are then manually scaled such that the iodine activities match the Dose Equivalent I-131 limit of 1.0 µCi/gm specified in the Technical Specifications, and the non-iodine isotopes are adjusted separately to meet the gross specific activity limits of 100/E. 'The resulting RCS equilibrium source term is shown in Table 2.2-4. _The noble gas concentrations from Table 2.2-4 correspond to a Dose Equivalent Xe-133 of215.1 µCi/gm based upon the definition of Dose Equivalent Xe-133 provided in TSTF-490 (Reference [4.30]).

2.2.4 Steam Generator Secondary Source Term

'The specific iodine activity on the secondary side of the steam generators is *limited to 0.1 µCi/gm Dose Equivalent 1-131, as shown in Table 2.2-1, which is one-tenth of the RCS iodine activity limit. Therefore, the secondary activities are developed by taking the iodine activities from Tal:ile 2.2-4 and reducing them by a factor often. The resulting secondary source term is presented in Table 2.2-5.

Table 2.2-1: Source Term Inputs and Assumptions Input/Assumption Value Core Power Level 3480 MWt (3468 MWt plus 0.34% uncertainty)

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Core Average Fuel Bumup 43,000 MWD/MTU Fuel Enrichment 1.5-5.0 w/o Uranium Mass per Fuel Assembly 498kg Number of Assemblies in the Core 193 Assembly Radial Peaking Factor 1.65 RCS Iodine Specific Activity Limit 1.0 µCi/gm Dose Equivalent 1-131 SG Secondary Iodine Specific Activity Limit 0.1 µCi/gm Dose Equivalent 1-1.31 Non-Iodine Gross Specific Activity Limit 100/E

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 11of75 Table 2.2-2: Core Source Term Activity Activity Nuclide Nuclide (Curies) (Curies)

Co-58 8.884E+05 Pr-143 l.398E+08 Co-60 6.796E+05 Nd-147 6.178E+07 Kr-85 l.280E+06 Np-239 2.609E+09 Kr-85m 2.364E+07 Pu-238 4.130E+05 Kr-87 4.661E+07 Pu-239 3.727E+04 Kr-88 6.222E+07 Pu-240 6.637E+04 Rb~86 2.272E+05 Pu-241 l.603E+07 Sr-89 8.677E+07 Am-241 l.707E+04 Sr-90 i.002E+07 Cm-242 7.417E+06 Sr-91 l.100E+08 Cm-244 l.838E+06 Sr-92 l.184E+08 Kr-83m l.119E+07 Y-90 l.038E+07 Br-82 3.972E+05 Y-91 l.142E+08 Br-83 l.106E+07 Y-92 l.197E+08 Br-84 2.009E+07 Y-93 l.358E+08 Rb-89 8.303E+07 Zr-95 l.566E+08 Y-91m 6.384E+07 Zr-97 l.586E+08 Y-95 l.496E+08 Nb-95 l.578E+08 Nb-95m l.795E+06 Mo-99 l.742E+08 Nb-97 l.596E+08 Tc-99m l.546E+08 Rh~103m l.849E+08 Ru-103 l.850E+08 Pd-109 5.749E+07 Ru-105 l.491E+08 Sb-124 l.434E+05 Ru-106 9.480E+07 Sb-125 l.231E+06 Rh-105 l.309E+08 Sb-126 5.873E+04 Sb-127 l.067E+07 Te-125m 2.725E+05

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Sb-129 3.215E+07 Te-131 8.174E+07 Te-127 l.054E+07 Te-133 l.024E+08 Te-127m l.841E+06 Te-133m 8.990E+07 Te-129 3.017E+07 Te-134 l.700E+08 Te-129m 5.821E+06 1-130 3.945E+06 Te-13lm 2.119E+07 Xe-131m l.385E+06

. Te-132 l.374E+08 Xe-133m 6.099E+06 1-131 9.814E+07 Xe-135m 4.335E+07 1-132 l.420E+08 Xe-138 l.627E+08

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 12 of 75 Activity Activity Nuclide Nuclide (Curies) (Curies)

I-133 l.916E+08 Cs-134m 5.865E+06 I-134 2.148E+08 Cs-138 l.776E+08 I-135 l.832E+08 Ba-141 l.522E+08 Xe-133 l.919E+08 La-143 l.419E+08 Xe-135 5.900E+07 Pm-147 l.944E+07 Cs-134 2.523E+07 Pm-148 l.841E+07 Cs-136 6.388E+06 Pm-148m 4.711E+06 Cs-137 1.325E+07 Pm-149 6.245E+07 Ba-139 l.693E+08 Pm-151 2.177E+07 Ba-140 l.639E+08 -Sm-153 6.797E+07 La-140 l.700E+08 Eu-154 9.557E+05 La-141 l.533E+08 Eu-155 4.427E+05 La-142 l.475E+08 Eu-156 4.798E+07 Ce-141 l.548E+08 Np-238 5.165E+07 Ce-143 l.430E+08 Pu-243 l.153E+08 Ce-144 l.296E+08 Am-242 l.148E+07 Table 2.2-3: Fuel Handling Accident Source Term Activity Activity Nuclide Nuclide (Ci) (Ci)

Co-58 7.595E+03 Pr-143 l.195E+06 Co-60 5.810E+03 Nd-147 5.282E+05 Kr-85 2.189E+04 Np-239 . 2.230E+07 Kr-85m 2.021E+05 Pu-238 3.531E+03 Kr-87 3.985E+05 Pu-239 3.186E+02 Kr-88 5.319E+05 Pu-240 5.674E+02 Rb-86 l.942E+03 Pu~241 l.370E+05 Sr-89 7.418E+05 Am-241 l.459E+02 Sr-90 8.566E+04 Cm-242 6.341E+04 Sr-91 9.404E+05 Cm-244 l.571E+04 Sr-92 l.012E+06 Kr-83m 9.567E+04 Y-90 8.874E+04 Br-82 3.396E+03 Y-91 9.763E+05 Br-83 9.455E+04 Y-92 l.023E+06 Br-84 l.718E+05

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 13 of 75 Activity Activity Nuclide Nuclide (Ci) (Ci)

Y-93 l.161E+06 Rb-89 7.098E+05 Zr-95 l.339E+06 Y-91m 5.458E+05 Zr-97 l.356E+06 Y-95 l.279E+06 Nb-95 l.349E+06 Nb-95m 1.535E+04 Mo-99 l.489E+06 Nb-97 l.364E+06 Tc-99m 1.322E+06 Rh-103m l.581E+06 Ru-103 1.582E+06 Pd-109 4.915E+05 Ru-105 l.275E+06 Sb-124 l.226E+03 Ru-106 8.105E+05 Sb-125 1.052E+04 Rh-105 l.119E+06 Sb-126 5.021E+02 Sb-127 9.122E+04 Te-125m 2.330E+03 Sb-129 2.749E+05 Te-131 6.988E+05 Te-127 9.011E+04 Te-133 8.754E+05 Te-127m l.574E+04 Te-133m 7.686E+05 Te-129 2.579E+05 Te-134 l.453E+06 Te-129m 4.977E+04 I-130 3.373E+04 Te-131m l.812E+05 Xe-131m l.184E+04 Te-132 l.175E+06 Xe-133m 5.214E+04 I-131 1.342E+06 Xe-135m 3.706E+05 I-132 l.214E+06 Xe-138 l.391E+06 1-133 l.638E+06 Cs-134m 5.014E+04 I-134 l.836E+06 Cs-138 l.518E+06 I-135 l.566E+06 Ba-141 l.301E+06

. Xe-133 l.641E+06 La-143 l.213E+06 Xe-135 5.044E+05 Pm-147 l.662E+05 Cs-134 2.157E+05 Pm-148 l.574E+05 Cs-136 5.461E+04 Pm-148m 4.028E+04 Cs-137 l.133E+05 Pm-149 5.339E+05 Ba-139 l.447E+06 Pm-151 l.86IE+05 Ba-140 l.401E+06 Sm-153 5.811E+05 La-140 1.453E+06 Eu-154 8.170E+03 La-141 l.311E+06 Eu-155 3.785E+03 La-142 l.261E+06 Eu-156 4.102E+05 Ce-141 l.323E+06 Np-238 4.416E+05 Ce-143 l.223E+06 Pu-243 9.857E+05 Ce-144 l.108E+06 Am-242 9.815E+04

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 14 of 75 Table 2.2-4: RCS Source Term Activity Activity Nuclide Nuclide

(µCi/g) (µCi/g)

Co-58 O.OOOE+OO Pr-143 6.713E-03 Co-60 O.OOOE+OO Nd-147 O.OOOE+OO Kr-85 2.385E+Ol Np-239 O.OOOE+OO Kr-85m 5.204E-01 Pu-238 O.OOOE+OO Kr-87 3.299E-01 Pu-239 O.OOOE+OO Kr-88 9.148E-01 Pu-240 O.OOOE+OO Rb-86 8.797E-02 Pu-241 O.OOOE+OO Sr-89 l.335E-03 Am-241 O.OOOE+OO Sr-90 l.237E-04 Cm-242 O.OOOE+OO Sr-91 5.681E-04 Cm-244 O.OOOE+OO Sr-92 2.488E-04 Kr-83m l.350E-01 Y-90 2.152E-04 Br-82 4.641E-03 Y-91 l.692E-02 Br-83 2.720E-02 Y-92 3.067E-04 Br-84 l.244E-02 Y-93 2.0lOE-04 Rb-89 2.530E-02 Zr-95 2.409E-02 Y-91m 3.314E-04 Zr-97 3.920E-04 Y-95 O.OOOE+OO Nb-95 3.478E-02 Nb-95m l.867E-04 Mo-99 2.070E+OO Nb-97 4.900E-05 Tc-99m l.980E+OO Rh-103m l.988E-02 Ru-103 l.991E-02 Pd-109 O.OOOE+OO Ru-105 9.723E-05 Sb-124 O.OOOE+OO Ru-106 3.340E-02 Sb-125 O.OOOE+OO Rh-105 7.689E-04 Sb-126 O.OOOE+OO Sb-127 O.OOOE+OO Te-125m 2.449E-02 Sb-129 O.OOOE+OO Te-131 l.599E-02 Te-127 2.489E-01 Te~133 O.OOOE+OO Te-127m 2.465E-01 Te-133m 7.643E-03 Te-129 2.281E-01 Te-134 l.092E-02 Te-129m 3.463E-01 1-130 O.OOOE+OO Te-131m 5.787E-02 Xe-131m l.600E+OO Te-132 9.639E-01 Xe-133m l.423E+OO I-131 8.087E-01 Xe-135m 2.138E-01 1-132 6.411E-01 Xe-138 2.292E-01

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~Jlis;uclat~ D. C. Cook AST Radiological Analyses Technical Report Page 15 of 75 Activity Activity Nuclide Nuclide

(µCi/g) (µCi/g)

I-133 l.0304E+OO Cs-134m 2.031E-02 I-134 l.23 lE-01 Cs-138 3.420E-Ol I-135 5.365E-01 Ba-141 4.233E-05 Xe-133 l.037E+02 La-143 O.OOOE+OO Xe-135 3.361E+OO Pm-147 O.OOOE+OO Cs-134 3.327E+Ol Pm-148 O.OOOE+OO Cs-136 2.188E+OO Pm-148m O.OOOE+OO Cs-137 l.852E+Ol Pm-149 O.OOOE+OO Ba-139 l.975E-04 Pm-151 O.OOOE+OO Ba-140 l.940E-03 Sm-153 O.OOOE+OO La-140 2.878E-03 Eu-154 O.OOOE+OO La-141 l.301E-04 Eu-155 O.OOOE+OO La-142 3.346E-05 Eu-156 O.OOOE+OO Ce-141 1.445E-02 Np-238 O.OOOE+OO Ce-143 6.91 lE-04 Pu-243 O.OOOE+OO Ce-144 4.229E-02 Am-242 O.OOOE+OO Table 2.2-5: SG Secondary Source Term Activity Nuclide

(µCi/g)

I-131 8.087E-02 I-132 6.41 lE-02 I-133 l.0304E-01 I-134 l.231E-02 I-135 5.365E-02 2.2.5 Fuel Rod Gap Fractions and High Burnup Rods The fraction of the core fission product inventory located within the fuel rod gap for non-LOCA events is specified in Table 3 of Reference [4.1] and shown below. These values apply to fuel rods that are damaged as a result of the event, and the entire contents of the fuel rod gap are instantaneously released from the fuel.

Note that separate gap inventory fractions of 10% for both noble gases and iodines are applied in the Control Rod Ejection event as required by Posit.ion 1 of Appendix H to Reference [4.1].

'~\

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RWA-1313-015, Rev. 1

'<'~~ocl_ate..-. D. C. Cook AST Radiological Analyses Technical Report Page 16 of 75 Table 2.2-6: Non-LOCA Fuel Rod Gap Inventory Fraction Group Fraction I-131 0.08 Kr-85 0.10 Other Noble Gases 0.05 Other Halogens . 0.05 Alkali Metals 0.12 Footnote 11 of Reference [4.1] states that these fuel rod gap inventories are acceptable for fuel rods with a peak burnup of up to 62,000 MWD/MTU provided that the linear heat generation rate does not exceed 6.3 kw/ft for rods with bumups greater than 54 GWD/MTU. For this analysis, a total of 150 rods in two fuel assemblies are assumed to exceed the burnup limits of Footnote 11 to allow for future core design margin.

Similar high burnup concerns were raised in alternative source term submittals by Fort Calhoun, Byron/Braidwood, and St. Lucie stations. For these plants, the issue was addressed by doubling the gap fractions for 100% of the rods in the affected assemblies and applying the maximum radial peaking factor.

This approach is discussed in Section 2.2.4 of References [4.18], Section 3.3 of Reference [4.19], and Section 2.9.2.2.2.1 of References [4.20] and [4.21]. This same methodology for addressing high bumup fuel is applied to this analysis.

2.3 Atmospheric Dispersion Factors The dose analysis addresses releases from either unit and must consider the dose Impact on all receptor locations applicable to both units. As such, atmospheric dispersion factors are developed for all possible release-receptor pairs, and the values applied in the analysis reflect the most limiting combination without regard to the unit in which the event occurs.

2.3.1 Onsite XIQ Determination New .x!Q factors for onsite release-receptor combinations are develope_d using the ARCON96 computer code (NUREG/CR-6331, Reference [4.7]). Reg. Guide 1.194 (Reference [4.26]) contains new guidance that supersedes the NUREG/CR-6331 recommendations for using certain default parameters as input. Therefore, the following changes from the default values are made:

• For surface roughness length, m, a value of 0.2 is used in lieu of the default value of 0.1, and

• For averaging sector width constant, a value of 4.3 is used in lieu of the default value of 4.0.

A number of various release-receptor combinations are considered for the onsite control room atmospheric dispersion factors. These different cases are considered to determine the limiting release-receptor combination for the events.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 17 of75 Figure 2.3-4 provides a sketch of the general layout of the Cook plant that has been annotated to highlight the onsite release and receptor point locations. All releases are taken as ground level releases per the guidance of Position 3.2.1 of Reg. Guide 1.194.

Table 2.3-2 provides information related to the relative elevations of the release-receptor combinations, the straight-line horizontal distance between the release point and the receptor location, and the direction (azimuth) from the receptor location to the release point. Angles are calculated based on trigonometric layout of release and receptor points in relation to the North-South and East-West axes. Plant North is offset 16.39 degrees clockwise from True North (17° 55' 40.5" - 1° 32' 32" = 16.39°) as shown in Figure 2.3-3. This offset is taken into account in development of the release-receptor pair angles.

A building wake term is only applied to releases close to or on the containment surface, where it is clear that the effect of the containment building wake will influence the result. Such releases include the Containment Vent, Containment Surface, Western SG PORVS/MSSVs, West Main Steam Enclosure, and Containment Diffuse Area locations for Units 1 and 2. The building area used for this wake term is 1,690 m2

• For scenarios with diffuse area releases, the building surface is modeled as a vertical planar area (or diffuse area) source in ARCON96 via two initial diffusion coefficients. The coefficients are calculated based on ~he guidance found in Position 3.2.4.4 of Reg. Guide 1.194. The area source width is the containment diameter of approximately 122 feet (ft), resulting in an initial diffusion coefficient ay0 of 6.2 m. The area source height, based conservatively on the elevation difference between the top of the containment building and the top of the tallest building connected to the containment surface, is approximately 89 .5 ft, resulting in an initial diffusion coefficient a 20 of 4.55 m.

Table 2.3-3 provides the Control Room.XIQ factors for the release-receptor combinations. These factors are not corrected for occupancy. This table summarizes the .XIQ factors for the control room intakes used in the various accident scenarios for onsite control room dose consequence analyses. Values are presented for the normal and emergency intakes for each unit.

Table 2.3-7 identifies the Release-Receptor pair and associated Control Room .XIQ factors from Table 2.3-3 that are used in the event analyses during each of the modes of control room ventilation.

2.3.2 Offsite X/Q Determination For offsite receptor locations, the new atmospheric dispersion (.XIQ) factors are developed using the PAVAN computer code (NUREG/CR-2858, Reference [4.8]). Table 2.3-4 provides the minimum distance to the EAB and LPZ in each direction for each release location. The offsite maximum .XIQ factors for the EAB and LPZ are presented in Table 2.3-6. According to Regulatory Position 4 of Reg. Guide 1.145 (Reference [4.27], the dose calculations should use the larger value between the maximum downwind sector and the 5% overall site

.XIQ values for each boundary and time period. For each individual PAVAN case evaluated, the 5% overall site

.XIQ values are bounded by the maximum sector .XIQ. Therefore, the .XIQ values for use in the EAB and LPZ dose evaluations for all releases are based on the maximum sector values provided in Table 2.3-6.

Offsite release-receptor pair locations are illustrated in Figure 2.3-5. The application of the X/Q values from Table 2.3-6 in the respective event analyses is identified in Table 2.3-7. All of the releases are considered to

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 18 of 75 be ground level releases based upon Position 1.3.2 of Reg. Guide 1.145. As such, the release height in PAVAN is set equal to 10.0 meters as required by Table 3.1 ofNUREG/CR-2858. The building area used for the building wake term is the same as for some of the ARCON96 onsite A'lQ cases, which is 1,690 m2

• The building height entered into PAVAN is the top elevation of the containment dome (49.5 m).

2.3.3 Meteorological Data Meteorological data from a five-year data set (2002, 2004, 2005, 2007, 2010) is used in the development of the new onsite and offsite A'lQ factors used in the analysis. These years were selected since they were the five most recent years with full periods of high quality data available (i.e., the data set from each year has greater than 90% of the possible hours recovered and used as input to the analyses). The meteorological data is converted from the raw format into the proper formatting required to create the meteorological data files for use with ARCON96 and PAVAN. The Cook Nuclear Plant records meteorological data from a primary, backup, and shoreline tower.

The data from the shoreline tower is considered to most accurately represent the meteorological conditions on-site based on its vicinity to the plant. However, the shoreline tower only records data at a height of 10 meters.

As a result, for on-site analyses, a hybrid meteorological data set is created using the 10 meter shoreline data for the lower level measurements and data from the 60 meter level on the primary tower for the upper level measurements. The stability classes are calculated based on the temperature difference between the 10 m and 60 m levels on the primary tower. The use of stability classes based on the primary tower data is considered to be conservative because the primary tower likely experiences more stable conditions due to its location further inland.

For off-site analyses, the use of primary tower data versus the shoreline tower data varies based on downwind sectors. A ridgeline running north-south is located approximately 0.5 miles east of the plant site. For locations west of the ridgeline, the shoreline tower data is more representative. For locations east of the ridgeline, the primary tower data is more representative. As a result, both the primary tower data and shoreline tower data is formatted into the joint frequency distribution for use in off-site analyses. Similar to the on-site analysis, the shoreline tower distribution is based on a hybrid meteorological data set using the 10 meter shoreline tower data for the lower level measurements, the 60 meter primary tower data for the upper level measurements, and

• stability classes calculated based on the temperature readings from the primary tower.

Due to the multiple joint frequency distributions, two PAVAN cases are developed for each release location, with one case incorporating the primary tower distribution and the other incorporating the shoreline tower distribution. The maximum A'lQ for each downwind direction at the EAB and LPZ is taken from the appropriate PAVAN output (shoreline or primary) based on the sector's location relative to the ridge line described above. The correlation between downwind direction and representative meteorological data set is provided in Table 2.3-5. Figure 2.3-1 and Figure 2.3-2 also present this information graphically.

RWA-1313-015 , Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 19 of 75 Figure 2.3-1: EAB Downwind Sectors

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RWA-1313-015, Rev. 1 D. C. Cook AST Rad iolog ical Analyses Technical Report Page 20 of 75 Figure 2.3-2: LPZ Downwind Sectors The years of data provided are based on the 5 most recent years that have valid data for both the primary and shoreline towers. Five years worth of meteorological data is used which meets the guidance set forth in Regulatory Position 3 .1 of Reg. Guide 1.194. The raw data for the years listed above was manipulated within a spreadsheet for appropriate formatting for use with ARCON96 and PAVAN.

The raw data was examined to identify and flag bad or missing data to ensure that the meteorological data used in the atmospheric dispersion factor determination were of high quality. All bad or missing data was removed from the data set prior to use in ARCON96 and PAV AN. No regulatory guidance is provided in Reg. Guide 1.194 and NUREG/CR-63 31 on the valid meteorological data recovery rate required for use in determining onsite >> Q values. However, Regulatory Position 5 of Reg. Guide 1.23 (Reference [4.28]) specifies a 90%

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 21of75 data recovery threshold for measuring and capturing meteorological data. The 90% data recovery rate applies to the composite of all variables needed to model atmospheric dispersion for each potential release pathway.

For the 2002, 2004, 2005, 2007, and 2010 data base, the meteorological data recovery rates are listed in Table 2.3-1. .

Table 2.3-1: Meteorological Data Recovery Rate Parameter 2002 2004 2005 2007 2010 Wind Speed I Om Primary 99.9 99.4 99.8 99.6 99.4 Wind Speed 60m Primary 99.9 99.4 99.8 94.3 89.5 Wind Speed I Om Shore 100 96 99.8 99.9 99.2 Wind Direction lOm Primary 99.9 99.4 99.8 99.6 99.5 Wind Direction 60m Primary 99.9 99.4 99.8 99.6 91.8 Wind Direction I Om Shore 100 96 99.8 100 98.6 Delta Temperature 60-IOm 99.8 99.3 99.6 99.4 98.1 While the 60 m wind speed for 20 I 0 has a recovery of 89 .5%, it is considered acceptable as the cumulative recovery rate for all years of 60 m wind speeds and all parameters for 2010 are well above 90%. With a total of five years worth of data, the contents of the meteorological data file are representative of the Jong-term meteorological trends at the Cook site.

The ~aw meteorological data was also processed into annual joint frequency distribution format for 2002, 2004, 2005, 2007, and 2010 for the offsite atmospheric dispersion factor analysis. The joint frequency distribution file requires the annual meteorological data to be sorted into three classifications that include wind direction, wind speed, and atmospheric stability class. The format for the file conforms to the format provided in Table 3 of Reg. Guide 1.23. The data for all years was sorted into wind speed bins using the guidance provided in RIS 2006-04.

The PAVAN code requires that the maximum speed for each wind speed category be input. The guidance provided in RIS 2006-04 gives a maximum wind speed category of 10 mis. However, in order to be consistent with Cook meteorological data evaluations, an upper limit of 14 m.Js is chosen in order to capture the average of all winds greater than 10 m.Js (approximately 12 m.Js) from the shoreline tower. The primary tower data set only has 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of data with wind speeds higher than 10 m.Js, which are included in the highest wind speed bin of 14 m.Js.

Red rf \ Wair RWA-1313-015, Rev. 1 A.._;.nr,l_~P.~.-..

D. C. Cook AST Radiological Analyses Technical Report Page 22 of75 Figure 2.3-3: Site Orientation GRID NORTH< .* PLANl NORTH

RWA-1313-015, Rev. 1 .

D. C. Cook AST Radiological Analyses Technical Report Page 23 of 75 Figure 2.3-4: Onsite Release-Receptor Locations 13

.17

• Unit 1 Contuinmcnt 0 5 9

11 19 21

---

1--CUE--+-------<EP--------------i 12 20 22 Unit 2 Cont al nnwnt \

10 4 2

• 6 18 14

• Figure is not to scale 1 - Unit 1 Containment Vent 12 - Unit 2 East Turbine Building 2 - Unit 2 Containment Vent 13 - Unit 1 RWST 3 - Closest Point on Unit 1 Containment 14 - Unit 2 RWST 4 - Closest Point on Unit 2 Containment 15 - Unit 1 Containment Diffuse Area 5 - Unit 1 Steam Jet Air Injectors 16 - Unit 2 Containment Diffuse Area 6 - Unit 2 Steam Jet Air Injections 17 - North Auxiliary Building Supply 7 - Unit 1 Western SG PORVs/MSSVs 18 - South Auxiliary Building Supply 8 - Unit 2 Western SG PORVs/MSSVs 19 - Unit 1 Normal Control Room Intake 9 - Unit 1 West Main Steam Enclosure 20 - Unit 2 Normal Control Room Intake 10 - Unit 2 West Main Steam Enclosure 21 - Unit 1 Emergency Control Room Intake 11 - Unit lEast Turbine Building 22 - Unit 2 Emergency Control Room Intake

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 24 of75 Figure 2.3-5: Offsite Release-Receptor Locations r  !'**North 7

• 3 9 r

• Unit I Containment 5

Unit 2 Omlainment 6

IO

• 4 8

• Figure is not to scale 1 - Unit 1 Containment Vent 8 - Unit 2 RWST 2 - Unit 2 Containment Vent 9 - North Auxiliary Building Supply 3 - Unit 1Turbine Building - NW Corner 10 - South Auxiliary Building Supply 4 - Unit 2 Turbine Building - SW Corner 11 - Unit 1 Containment Surface 5 - Unit 1 East Main Steam Enclosure 12 - Unit 2 Containment Surface 6 - Unit 2 East Main Steam Enclosure 13 - Unit 1 West Main Steam Enclosure

. 7 - Unit 1 RWST 14 - Unit 2 West Main Steam Enclosure

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D. C. Cook AST Radiological Analyses Technical Report Page 25 of 75 Table 2.3-2: Onsite Release-Receptor Combination Parameters Release- Receptor Release Receptor Building Distance Direction Receptor Release Location Location Height Height Area (m) (deg)

Pair (CR Intake) (m) (m) (m2)

A Ul Plant Vent Ul Normal 49.5 15.3 44.4 79 1690 B Ul Plant Vent Ul Emergency 49.5 14.7 42.0 77 1690 c U2 Plant Vent U2 Normal 49.5 14.4 44.4 133 1690 D U2 Plant Vent U2 Emergency 49.5 14.7 42.0 135 1690 E U2 Containment U2 Normal 14.4 14.4 25.8 133 1690 Closest Pt.

F U2 Containment U2 Emergency 14.7 14.7 23.4 136 1690 Closest Pt.

G U2 PORV/MSSV U2 Normal . 24.8 14.4 23.2 150 1690 H U2 PORV/MSSV U2 Emergency 24.8 14.7 21.4 155 1690 I U2 Turbine Bldg U2 Normal 14.4 14.4 9.8 286 0.01 J U2 Turbine Bldg U2 Emergency 14.7 14.7 12.5 286 0.01 K U2RWST U2 Normal 12.9 14.4 68.6 161 0.01 L U2RWST U2 Emergency 12.9 14.7 67.1 163 0.01 M U2 Containment U2 Normal 14.4 14.4 25.8 133 1690 Surface (Diffuse)

U2 Containment U2 Emergency 14.7 14.7 23.4 136 1690 N

Surface (Diffuse) 0 Ul SJAE Ul Normal 1.5 15.3 87.4 336 0.01 p South Auxiliary U2 Normal 10.9 14.4 27.2 188 0.01 Building Intake

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 26 of75 Table 2.3-3: Onsite Atmospheric Dispersion Factors Release-Receptor 0-2 hour 2-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Point XIQ XIQ XIQ XIQ XIQ Pair ID A Ul Plant Vent Ul Nopnal 2.03E l.37E-03 4.89E-04 3.84E-04 2.62E-04 Ul B Ul Plant Vent 2.17E-03 l.42E-03 5.18E-04 3.99E-04 2.72E-04 Emergency c U2 Plant Vent U2 Normal 2.lOE-03 1.48E-03 5.52E-04 3.98E-04 3.17E-04 U2 D U2 Plant Vent 2.28E-03 l.59E-03 5.95E-04 4.46E-04 3.49E-04 Emergency U2 Containment E U2 Normal 1.02E-02 8.41E-03 2.74E-03 2.66E-03 2.34E-03 Closest Pt.

U2 Containment U2 F l.24E-02 l.02E-02 3.32E-03 3.24E-03 2.84E-03 Closest Pt. Emergency G U2*PORV/MSSV U2 Normal l.09E-02 8.61E-03 2.87E-03 2.78E-03 2.SOE-03 U2 H U2 PORV/MSSV l.26E-02 9.72E-03 3.26E-03 3.l 7E-03 2.SOE-03 Emergency I U2 Turbine Bldg U2 Normal 4.57E-02 3.14E-02 l.27E-02 8.30E-03 6.73E-03 U2 J U2 Turbine Bldg 2.91E-02 2.02E-02 8.14E-03 5.34E-03 4.32E~03 Emergency K U2RWST U2 Normal l.74E-03 l.44E-03 5.24E-04 4.42E-04 3.86E-04 U2 L U2RWST l.81E-03 1.45E-03 5.43E-04 4.43E-04 3.86E-04 Emergency U2 Containment M U2 Normal 2.SlE-03 l.94E-03 6.66E-04 6.44E-04 5.61E-04 Surface (Diffuse)

U2 Containment U2 N 2.73E-03* 2.0SE-03 7.04E-04 6.89E-04 5.95E-04 Surface (Diffuse) Emergency 0 Ul SJAE Ul Normal 8.SOE-04 p South Auxiliary U2 Normal 7.91E-03 5.93E-03 2.12E-03 l.SOE-03 l.OlE-03 Building Intake

• 0-2 hour value is from Unit 1 containment to Unit 1 emergency intake

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 27 of 75 Table 2.3-4: Offsite Release-Receptor Distances Minimum EAB MinimumLPZ Release Direction Distance (m) Distance (m)

N 542 3,149 NNE 604 3,149 NE 702 3,149 ENE 823 3,158 E 1,525 3,176 ESE 1,292 3,201 SE 982 3,219 SSE 695 3,219 Unit 1 Containment Vent s 683 3,219 SSW 610 3,219 SW 610 3,219 WSW 610 3,219 w 599 3,211 WNW 575 3,185 NW 554 3,166 NNW 542 3,152 N 609 3,218 NNE 609 3,218 NE 769 3,218 ENE 946 3,218 E 1,573 3,211 ESE 1,267 3,185 SE 790 3,166 SSE 632 3,152 Unit 2 Containment Vent s 613 3,149 SSW 542 3,149 SW 542 3,149 WSW 548 3,158 w 565 3,176 WNW 589 3,201 NW 609 3,219 NNW 609 3,218

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 28 of 75 MinimumEAB MinimumLPZ Release Direction Distance (m) Distance (m)

N 434 3,043 NNE 442 3,054 NE 580 3,086 ENE 770 3,141 E 1,563 3,207 ESE 1,413 3,275 SE 965 3,329 Unit 1 Turbine Building-NW SSE 78,6 3,317 Comer s 666 3,281 SSW 615 3,235 SW 566 3,186 WSW 527 3,142 w 483 3,103 WNW 449 3,061 NW' 434 3,044 NNW 434 3,043 N 591 3,211 NNE 644 3,259 NE 793 3,301 ENE 990 3,333 E 1,695 3,305 ESE 1,048 3,241 SE 674 3,172 Unit 2 Turbine Building~sw SSE 563 3,110 Comer s 519 3,065 SSW 435 3,045 SW 434 3,043 WSW 434 3,043 w 439 3,051 WNW 465 3,080 NW 506 3,128 NNW 548 3,164

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 29 of 75 Minimum EAB MinimumLPZ Release Direction Distance (m) Distance (m)

N 551 3,146 NNE 638 3,143 NE 695 3,143 ENE 810 3,144 E 1,523 3,154 ESE 1,267 3,174 SE 965 3,192 Unit 1 East Main Steam SSE 685 3,195 Enclosure s 680 3,202 SSW 617 3,212 SW 617 3,222 WSW 623 3,232 w 626 3,238 WNW 597 3,209 NW 569 3,181 NNW 551 3,161 N 617 3,216 NNE 684 3,206 NE 764 3,198 ENE 933 3,193 E 1,548 3,184 ESE 1,242 3,163 SE 798 3,148 Unit 2 East Main Steam SSE 620 3,144 Enclosure s 610 3,144 SSW 550 3,144 SW 550 3,153 WSW 560 3,172 w 583 3,198 WNW 612 3,227 NW 627 3,236 NNW 617 3,227

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D. C. Cook AST Radiological Analyses Technical Report RWA-1313-015, Rev. 1 Page 30 of 75 Minimum EAB MinimumLPZ Release Direction Distance (m) Distance (m)

N 508 3,112 NNE 603 3,112 NE 667 3,113 ENE 783 3,126 E 1,503 3,154 ESE 1,305 3,187 SE 1,010 3,223 SSE 728 3,237 Unit 1 RWST s 719 3,247 SSW 641 3,251 SW 633 3,242 WSW "620 3,230 w 592 3,207 WNW 552 3,168 NW 524 3,136 NNW 508 3,118 N 637 3,247 NNE 644 3,251 NE 805 3,242 ENE 978 3,230 E 1,587 3,207 ESE 1,256 3,168 SE 763 3,136 SSE 600 3,118 Unit 2 RWST s 577 3,113 SSW 508 3,113 SW 508 3,113 WSW 515 3,127 w 539 3,155 WNW 572 3,188 .

NW 613 3,223 NNW 627 3,237

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 31 of75 MinimumEAB MinimumLPZ Relea.se Direction Distance (m) Distance (m)

N 525 3,134 NNE 526 3,135 NE 705 3,147 ENE 816 3,170 E 1,549 3,200 ESE 1,327 3,234 SE 898 3,247 North A~xiliary Building SSE 716 3,237 Supply s 694 3,223 SSW* 599 3,209 SW 587 3,196 WSW 577 3,187 w 563 3,175 WNW 540 3,152 NW 527 3,137 NNW 525 3,134 N 592 3,202 NNE 606 3,216 NE 786 3,230 ENE 908 3,242 E 1,609 3,246 ESE 1,299 3,213 SE 808 3,181 South Auxiliary Building SSE 642 3,155 Supply s* 611 3,139 SSW 525 3,134 SW 525 3,134 WSW 525 3,134 w 534 3,144 WNW 553 3,165 NW 575 3,184 NNW 581 3,191

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RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 32 of 75 MinimumEAB MinimumLPZ Release Direction Distance (m) Distance (m)

N 523 3,130 NNE 585 3,130 NE 683 3,130 ENE 804 3,139 E 1,506 3,157 ESE 1,273 3,182 SE 963 3,200 SSE 676 3,200 Unit 1 Containment Surface s 664 3,200 SSW 591 3,200 SW ' 591 3,200 WSW 591 3,200 w 580 3,192 WNW 556 3,166 NW 5'.?5 3,147 NNW 523 3,133 N 590 3,199 NNE 590 3,199 NE 750 3,199

. ENE 927 3,1~9 E 1,554 3,192 ESE 1,248 3,166 SE 771 3,147 SSE 613 3,133 Unit 2 Containment Surface s 594 3,130 SSW 523 3,130 SW 523 3,130 WSW 529 3,139 w 546 3,157 WNW 570 3,182 NW 590 3,200 NNW 590 3,199

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 33 of 75 MinimumEAB Minimum LPZ Release Direction Distance (m) Distance (m)

N 541 3,150 NNE 541 3,150 NE 715 3,158 ENE 833 3,176 E 1,549 3,200 ESE 1,313 3,227 SE 881 3,231 Unit 1 West Main Steam SSE 700 3,221 Enclosure s 680 3,211 SSW 592 3,202 SW 586 3,195 WSW 583 3,192 w 577 3,186 WNW 555 3,167 NW 544 3,153 NNW 541 3,150 N 589 3,198 NNE 596 3,206 NE 770 3,215 ENE 891 3,226 E 1,595 3,235 ESE 1,294 3,210 SE 812 3,185 Unit 2 West Main Steam SSE 649 3,166 Enclosure s 623 3,153 SSW 541 3,150 SW 541 3,150 WSW 542 3,151 w 550 3,160 WNW 566 3,178 NW 583 3,192 NNW 584 3,193

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 34 of 75 Table 2.3-5: Downwind Direction- Met Data Correlation MetData Set Direction EAB LPZ N Shoreline Shoreline NNE Shoreline Primary NE Primary Primary ENE Primary Primary E Primary Primary ESE Primary Primary SE Primary Primary SSE Shoreline Primary s Shoreline Primary SSW Shoreline Shoreline SW Shoreline Shoreline WSW Shoreline Shoreline w Shoreline Shoreline WNW Shoreline Shoreline NW Shoreline Shoreline NNW Shoreline Shoreline Table 2.3-6: Offsite Atmospheric Dispersion Factors Release- '

Receptor 0-2 hour 0-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Point XIQ XIQ XIQ XIQ XIQ Pair ID Q Ul Plant Vent BAB 5.69E-04 R Ul Plant Vent LPZ l.13E-04 5.28E-05 3.62E-05 1.64E-05 6.30E-06 Ul Containment s Surface BAB 6.04E-04 Ul Containment T LPZ l.13E-04 5.32E-05 3.64E-05 l.66E-05 6.37E-06 Surface Ul Turbine u Bldg BAB 8.62E-04 Ul Turbine v Bldg LPZ l.16E-04 5.45E-05 3.74E-05 l.74E-05 6.74E-06 Ul West Main w Steam Enclosure BAB 5.87E-04

. RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 35 of 75 Release-Receptor 0-2 hour 0-8 hour 8-24 hour 1-4 days 4-30 days Receptor Release Point Point X/Q X/Q X/Q X/Q X/Q Pair ID Ul Main Steam x Enclosures LPZ l.13E-04 5.29E-05 3.63E-05 l.65E-05 6.36E-06 y Ul RWST EAB 6.25E-04 z Ul RWST LPZ l.14E-04 5.35E-05 3.67E-05 l.65E-05 6.36E-06 North Auxiliary AA EAB 6.19E-04 Building Suppiy North Auxiliary BB Building Supply LPZ l.13E-04 5.3 IE-05 3.64E-05 1.67E-05 6.42E-06

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 36 of75 Table 2.3-7: Release-Receptor Pairs Application to the Event Analyses Normal Emergency Event EAB LPZ Intake (I) Intake (I)

LOCA:

- Containment Purge c D Q R

- Containment Leakage M N s T

- ESP Leakage c D Q R

- RWST Backleakage K L y z FHA

- Containment Release E F s T

- Auxiliary Bldg. Release A B Q R MSLB:

- Break Release I J u v

- Intact SG Release G H u v SGTR 0 0<2) H u w<3) v x<3l Locked Rotor G H w x Control Rod Ejection:

- Containment Leakage M N s T

- Secondary Side Release G H w x WGDT Rupture p n/a AA BB VCTRupture p n/a AA BB (1) Control room makeup flow enters through the normal intake prior to realignment of the control room ventilation system and enters through the emergency intake after control room isolation. Unfiltered inleakage enters the control room envelope through the normal intake for the duration of the event.

(2) Prior to reactor trip, the release receptor pair is from the SJAE to the normal intake. The release point changes to the PORV/MSSV immediately after the trip, and the receptor point shifts to the emergency intake following control room isolation.

(3) Prior to reactor trip, the release is from the turbine building. The release point changes to the west niain steam enclosure following the trip.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 37 of75 The breathing rates at EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Table 2.3-8: Offsite Breathing Rates Time EAB/LPZ (hours) (m 3/sec) 0.0 3.5 x 10-4 8.0 1.8 x 10-4 24.0 2.3 x 10-4 720.0 2.3 x 10-4 2.4 Direct Shine Dose In addition to the dose from contamination of the control room atmosphere by intake or infiltration, the total control room dose also requires consideration of direct shine dose contributions from control room filters, from the external radiation plume, and from radioactive material in the containment building. The filter shine dose is calculated by first determining the maximum activity loading on the control room ventilation system filters during the LOCA event. This is done by considering the control room ventilation maximum fan capacity flow rate along with filter efficiencies of 100%. The activities from the recirculation filter edit of the RADTRAD output files are then input into a MicroShield 8.03 model that reflects the geometry of the control room filter housing and the recirculation air handler unit position with respect to the control room. Credit is taken for shielding by structural materials and attenuation in air. An integrated 30-day dose is calculated for control room personnel.

The control room dose due to direct radiation streaming through the equipment hatch following a LOCA conservatively assumes that the control room receptor is positioned directly in front of the equipment hatch such that the exposure from containment is due to a direct line of sight through the entire area of the hatch.

The analysis also under-estimates the total thickness of structural walls in the Auxiliary Building. The result is a conservative 30-day dose to an individual in the control room due to direct shine through the containment equipment hatch.

The dose contribution from the external cloud is assessed qualitatively using the guidelines of Reference [4.22]

which states that 18 inches of concrete is generally adequate to attenuate the external DBA radiation to negligible levels. A review of site drawings revealed that while the minimum thickness of the control room walls is 18 inches, the control room is surrounded by 3 ft thick Auxiliary Building and 3 '-6" Turbine Building walls. Similarly, the thickness of the concrete ceiling immediately above the control room is 18"; however, the thickness of the concrete roof on the elevation above the control room is 2' -6". As such, the shielding against external radiation sources well exceeds the amount identified as adequate in the guidance. Therefore, the additional shine dose contribution to control room personnel from the radioactive plume would be negligible.

The total LOCA shine dose is presented in Table 2.4-1. The filter shine dose following the LOCA event is conservatively applied to all other events.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technipal Report Page 38 of 75 Table 2.4-1: LOCA Direct Shine Dose Dose Source (rem)

Containment 0.246 Control Room Filters 0.139 External Cloud Negligible Total 0.385 3 Event Analyses

3. I Loss of Coolant Accident Control room and offsite doses are calculated for the LOCA event using the methodology outlined in Appendix A of Reference [4.1]. The dose contribution from the following four different radionuclide release pathways are determined separately and then combined to obtain the total dose for the event:

• Containment Purge

• Containment Leakage

• ESF Leakage to the Allxiliary Building

• ESF Leakage to the Refueling Water Storage Tank (RWST)

For all four cases, the control room ventilation system is automatically placed into the pressurization mode upon receipt of a safety injection signal. Parameters used in modeling the control room ventilation system are shown in Table 2.1-1.

3.1.1 Containment Purge This containment purge release pathway represents releases through the Containment Purge Supply and Exhaust System* prior to containment isolation. Since the purge system is isolated within 15 seconds following the initiation of the event, the release is secured well before the onset of the gap release at 30 seconds as defined in Table 4 of Reference [4.1]. Therefore, only those isotopes initially contained in the RCS fluid (Table 2.2-4) are available for release from containment, which are assumed to be instantaneously and homogeneously mixed throughout the containment atmosphere at the initiation of the event. The containment is modeled as a single compartment without credit for isotope removal by sprays or deposition. Radionuclides are released from containment directly to the environment without mitigation until the containment purge system is isolated. In addition, there is no radionuclide reduction by the containment purge ventilation system, which exhausts to the plant vent.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 39 of 75 3.1.2 Containment Leakage For the containment leakage case, 100% of the core becomes damaged and a phased release of the core fission product inventory (Table 2.2-2) occurs. The fraction of the nuclides in the core that become deposited into containment and the timing ofthis release are shown in Table 3.1-1 and Table 3.1-2.

Table 3.1-1: Core Inventory Fraction Release into Containment Gap Early Group Group Elements Release In-Vessel Number Name Phase Phase 1 Noble Gases Xe,Kr 0.05 0.95 2 Halogens I, Br 0.05 0.35 3 Cesium Cs, Rb 0.05 0.25 4 Tellurium Te, Sb, Se 0.0 0.05 5 Strontium Sr 0.0 0.02 6 Barium Ba 0.0 0.02 7 Ruthenium Ru,Rh,Pd,:rvio, Tc, Co 0.0 0.0025 8 Cerium Ce, Pu, Np 0.0 0.0005 9 Lanthanum La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am 0.0 0.0002 Table 3.1-2: LOCA Release Phase Timing Phase Onset Duration Gap Release 30 sec 0.5 hr Early In-Vessel 0.5 hr 1.3 hr The released nuclides are assumed to be instantaneously and homogeneously mixed throughout the containment. The D. C. Cook containment building is modeled as seven separate regions. The entire containment building is served by the safety related Containment Ventilation (CEQ) System. The CEQ System does not have any filtration capability; however, it does provide some additional level of mixing between the regions. Three of the regions are capable of being sprayed by the Containment Spray (CTS)

System. Spray induced mixing is also credited between adjacent sprayed and unsprayed regions based on two turnovers of the unsprayed volume per hour. The CTS System is assumed to be secured after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Iodine is released into the containment with a chemical composition of 95% particulate, 4.85% elemental, and 0.15% organic. The containment sump pH is maintained greater than 7.0 following the onset of containment sprays; therefore, re-evolution of particulate iodine into elemental iodine is not considered. Spray removal of elemental and aerosol iodine is credited using the guidance of Reference [4.23]. Removal of elemental iodine in each of the sprayed regions is terminated when the elemental decontamination factor in the region reaches a value of 200. Similarly, the removal rate of the aerosol iodine is reduced by a factor of 10 when the aerosol decontamination factor reaches 50. Natural deposition of only the aerosol iodine is considered in the analysis, and deposition is credited only in unsprayed regions and in sprayed regions after the CTS system has been secured.

,(' \

Red.Walt: RWA-1313-015, Rev. 1 A.;f.nclar~.

D. C. Cook AST Radiological Analyses Technical Report Page 40 of 75 Unfiltered leakage from the containment to the environment is assumed to occur uniformly from all seven regions at an initial rate of0.18%/day. This leakage rate is reduced by 50% to 0.09%/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The release from the containment is based upon an atmospheric dispersion factor assuming a diffuse source from the containment surface.

3.1.3 ESF Leakage into the Auxiliary Building ESF leakage outside of containinent results from operation of ECCS systems which take suction from the containment sump and allow system fluid to be released into the Auxiliary Building through pump seals, valve packing glands, and flanged connections. For this case, a portion of the core source term from Table 2.2-2 is deposited into liquid in the containment sump according to the release fraction and timing shown in Table 3.1-1 and Table 3.1-2. With the exception of noble gases, the fission products released from the fuel are assumed to instantaneously and homogeneously mix in the containment sump water. Leakage from the ECCS systems begins at the onset of switchover to recirculation, and the leak rate into the building is taken as two times the allowable limit established by the leak rate monitoring program. Once the sump fluid exits the system, particulate nuclides are assumed to be retained in the liquid phase, which limits the release to iodine isotopes only. Ten percent of the iodines in the sump fluid are then assumed to become airborne and are released directly to the environment. No credit is taken for holdup or dilution in the Auxiliary Building, or for filtration removal by the ESF Ventilation system. All releases from the Auxiliary Building occur from the plant vent.

3.1.4 ESF Leakage into the RWST The evaluation ofESF leakage through valves that isolate interfacing systems from the RWST is performed in a separate case. The sump activity for this case is identical to that applied in the case of ESF leakage to the Auxiliary Building described in Section 3.1.3. For this case, flow from the sump into the RWST is assumed to begin immediately upon switchover to recirculation at a rate of 1 gpm (two times 0.5 gpm). The modeling of the release of volatile iodine from the tank to the atmosphere is based upon the guidance ofNUREG/CR-5950 (Reference [4.25]). Based upon sump pH controls, the iodine in the sump is considered to be nonvolatile.

However, when introduced into the acidic solution of the RWST, a portion of the particulate iodine is converted to elemental iodine. The methodology of NUREG/CR-5950 accounts for two iodine transport/release mechanisms. The first is the fraction of the total iodine in the tank that is released in the form elemental iodine, and the second is the partitioning of elemental iodine between the liquid and vapor phases of the tank.

The fraction of the total iodine that becomes elemental is a function of both the RWST pH and the total iodine concentration in the tank. The analysis determines the time-dependent RWST pH profile as the sump fluid, with a constant pH of 7 .0, mixes with the remaining inventory in the _tank following switchover to recirculation. Similarly, the total RWST iodine concentration is calculated as sump iodine is transported into the tank from the sump. Calculation of the time-dependent iodine concentration in the RWST liquid for purposes of determining the elemental iodine release fraction conservatively neglects the reduction in concentration due to the release of iodine into the vapor phase. These two parameters combine to produce the elemental iodine fraction, which increases from a value of 0.0 at the beginning of the event to a maximum of 0.1914.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 41 of 75 The ratio of the elemental iodine concentrations between the liquid and vapor phases of the tank is determined by a partition coefficient that is a function of the RWST liquid temperature. The analysis calculates a conservatively high RWST temperature profile using GOTHIC by introducing hot sump fluid into the tank without credit for heat removal in the piping between the sump and the tank or heat losses through the tank walls. This results in a time-dependent partition coefficient that decreases from 45.41 at the beginning of the event to 31.92 after 30 days. The elemental iodine fraction and partition coefficient are applied to the leakage flow rate from the sump to the RWST to obtain an adjusted elemental iodine release rate from the tank. A similar approach is taken with the organic iodine, using a release fraction of 0.0015 taken from Position 2 of Appendix A to Reference [4.1] and assuming a conservative partition coefficient of 1.0. The release location for this event is from the RWST vent.

Values of key inputs and assumptions important to the LOCA analysis are provided in Table 3.1-3. The dose consequences for this event are presented in Table 3.1-10.

Table 3.1-3: LOCA Inputs and Assumptions Input/Assumption Value Containment Purge Source Term Initial RCS Activity (Table 2.2-4)

Iodine Chemical Form 95% aerosol, 4.85% elemental, 0.15% organic Containment Volume 1,066,352 ft 3 (minimum)

Containment Purge Flow Rate 36,300 cfm Containment Purge Isolation time 15 seconds Containment Purge Filtration 0%

Removal by Wall Deposition None Removal by Sprays None Containment Leakage Source Term Core Inventory (Table 2.2-2)

Iodine Chemical Form 95% aerosol, 4.85% elemental, 0.15% organic Containment Sump pH >7.0 Compartment Volumes (max}

Upper Containment (Sprayed) 621,968 ft 3 Lower Containment (Sprayed) 103,770 ft 3 Fan Rooms (Sprayed) 48,913 ft 3 Upper Containment (Unsprayed) 122,600 ft 3 Ice Condenser (Unsprayed) 105,577 ft 3 Lower Containment (Unsprayed) 66,188 ft 3 Dead-End (Unsprayed) 18,663 ft 3

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 42 of75 Input/Assumption Value Containment Ventilation Start Time 300 seconds 1

Containment Ventilation Flow Rate Fan Rooms to Lower Containment (Unsprayed) 14,580.5 cfm Fan Rooms to Lower Containment (Sprayed) 22,859.5 cfm Lower Containment (Unsprayed) to Dead -End 90cfm Dead-End to Fan Rooms 90cfm Lower Containment (Unsprayed) to Fan Rooms l,350 cfm Lower Containment (Unsprayed) to Ice Condenser 13,140.5 cfm Lower Containment (Sprayed) to Ice Condenser 22,859.5 cfm Ice Condenser to Upper Containment (Sprayed) 30,072.3 cfm Ice Condenser to Upper Containment (Unsprayed) 5,927.7 cfm Upper Containment (Sprayed) to Fan Rooms 30,072.3 cfm Upper Containment (Unsprayed) to Fan Rooms 5,927.7 cfm Lower Containment - Sprayed to/from Unsprayed 2206.3 cfm (spray induced circulation)

Uppe~ Containment - Sprayed to/from Unsprayed 4086.7 cfm (spray incluced circulation)

Sprayed/Unsprayed Volume Induced Mixing Flow 2 Turnovers of Unsprayed Compartment/hour Rate Containment Spray Start Time 300 seconds Containment Spray Stop Time 0 .319-0 .426 hours0.00493 days <br />0.118 hours <br />7.043651e-4 weeks <br />1.62093e-4 months <br /> and after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Containment Spray Flow Rate Upper Containment 1466 gpm Lower Containment 660 gpm Fan Rooms 201 gpm Containment Spray Drop Fall Height Upper Containment 58.6 ft Lower Containmep.t 28,5 ft Fan Rooms 20.1 ft Containment Spray Mean Drop Diameter Upper Containment 609 microns Lower Containment 671 microns Fan Rooms 671 microns Elemental Iodine Spray Removal Coefficient 20 hr"1, with a total decontamination factor of200 Time that Total Elemental DF reaches 200 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 43 of 75 Input/Assumption Value Aerosol Spray Removal Coefficient Upper Containment 5.06 hr-I Lower Containment 6.65 hr-I Fan Rooms 3.03 hr- 1 Time that Total Aerosol DF reaches 50 2.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> Organic Iodine Spray Removal None Elemental, Organic Iodine - None Natural Deposition Aerosols - 0.1 hr- 1in unsprayed regions only Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.18 %/day 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 30 days 0.09 %/day Containment Leakage Filtration 0%

ESF Leakage to the Auxiliary Building Source Term Core Inventory (Table 2.2-2)

Iodine Chemical Form 0% aerosol, 97% elemental, 3% organic

~ontainment Sump Volume 50,955 ft3 ECCS Recirculation Start Time 1388.4 seconds ESF Leakage Flow Rate 0.2 gpm (two times the allowable value)

ESF Leakage Flashing Fraction 10%

Auxiliary Building Ventilation Filtration 0%

ESF Leakage to the RWST Source Term Core Inventory (Table 2.2-2)

Containment Sump Volume 50,955 ft3 ECCS Recirculation Start Time 1388.4 seconds ESF Leakage Flow Rate 1.0 gpm (two times the allowable value)

Total iodine mass released into the sump 12,035.5 grams Sump iodine concentration 6.573xlo-05 g-atom/liter Sump pH 7.0 Total RWST Volume 420,000 gallons 53,637.5 gallons (minimum at time of Initial RWST Liquid Volume switchover)

RWST Liquid Iodine Concentration 0.0 - 2.93lxl0-05 g-atom/liter (Table 3.1-4)

(if\,

\ Red Woll' RWA-1313-015, Rev. 1

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D. C. Cook AST Radiological Analyses Technical Report Page 44 of 75 Input/Assumption Value RWSTpH 4.479 - 4.734 (Table 3.1-5)

Elemental Iodine Release Fraction 0.0 - 0.1914 (Table 3.1-6)

Organic Iodine Fraction 0.0015 RWST Liquid Temperature 100-118.5 °F (Table 3.1-7)

RWST Liquid/Vapor Iodine Partition Coefficient Elemental 31.92 - 45.41 (Table 3.1-8)

Organic 1.0 Adjusted RWST Iodine Release Rate Table 3.1-9 Dose Conversion Inputs:

Atmospheric Dispersion Factors Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11&FGR12 Breathing Rates EAB/LPZ The breathing rates at the BAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 10-4 m3/sec Control Room Model Inputs Table 2.1-1 Control Room Isolation Time 70 seconds (Safety Injection)

Control Room Occupancy Factor Table 2.1-1

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 45 of 75 Table 3.1-4: RWST Liquid Iodine Concentration Iodine Time Concentration (hr}

(g-atom/liter) 0.0 O.OOOE+OO 0.386 O.OOOE+OO 0.50 8.381E-09 1.0 4.512E-08 5.0 3.375E-07 10.0 6.994E-07 15.0 l.057E-06 25.0 l.761E-06 50.0 3.456E-06 75.0 5.064E-06 100.0 6.590E-06 125.0 . 8.042E-06 150.0 9.424E-06

. 200.0 l.200E-05 250.0 l.435E-05 300.0 l.650E-05 350.0 l.848E-05 400.0 2.031E-05 450.0 2.200E-05 500.0 2.357E-05 550.0 2.503E-05 600.0 2.639E-05 650.0 2.766E-05 700.0 2.886E-05 720.0 2.931E-05

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 46 of75 Table 3.1-5: RWST pH Time pH (hr) 0.0 4.479 0.386 4.479 0.50 4.479" 1.0 4.479 5.0 4.481 10.0 4.484 15.0 4.486 25.0 4.491 50.0 4.502 75.0 4.514 100.0 4.525 125.0 4.535 150.0 4.546 200.0 4.566 250.0 4.586 300.0 4.604 350.0 4.622 400.0 4.639 450.0 4.655 500.0 4.671 550.0 4.686 600.0 4.701 650.0 4.715 700.0 4.729 720.0 4.734

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 47 of 75 Table 3.1-6: Elemental Iodine Release Fraction Time 12 Fraction (hr) 0.0 0.0000 0.386 0.0000 0.50 0.0002 1.0 0.0011 5.0 0.0080 10.0 0.0161 15.0 . 0.0238 25.0 0.0379 50.0 0.0673 75.0 0.0902 100.0 0.1085 125.0 0.1234 150.0 0.1356 200.0 0.1541 250.0 0.1670 300.0 0.1761 350.0 0.1824 400.0 0.1866 450.0 0.1893 500.0 0.1908 550.0 0.1914 600.0 0.1914 650.0 0.1907 700.0 0.1896 720.0 0.1890

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 48 of 75 Table 3.1-7: RWST Temperature Profile Time Temperature (hr) OF 0.0 100.0 0.386 100.0 0.50 . 100.0 1.0 100.l 5.0 100.3 10.0 100.5 15.0 100.8 25.0 101.3 50.0 102.4 75.0 103.5 100.0 104.5 125.0 105.5 150.0 106.4 200.0 108.1 250.0 109.7 300.0 111.1 350.0 112.2 400.0 113.3 450.0 114.3 500.0 115.2 550.0 116.0 600.0 116.8 650.0 117.5 700.0 118.2 720.0 118.5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 49 of 75 Table 3.1-8: RWST 12 Partition Coefficient Time Partition (hr) Coefficient 0.0 45.41 0.386 45.41 0.50 45.41 1.0 45.33 5.0 45.15 10.0 44.98 15.0 44.73 25.0 44.30 50.0 43.38 75.0 42.48 100.0 41.68 125.0 40.89 150.0 40.20 200.0 38.92 250.0 37.75 300.0 36.75 350.0 35.99 400.0 35.24 450.0 34.58 500.0 33.99 550.0 33.48 600.0 32.97 650.0 32.53 700.0 32.10 720.0 31.92

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 50 of75 Table 3.1~9: Adjusted RWST Iodine Release Rate Time Release Rate (hr) (cfm) 0.0 0.0 0.386 6.790E-07 10.0 2.969E-06 25.0 1.300E-05 75.0 3.318E-05 125.0 6.398E-05 200.0 l.118E-04 300.0 l.SOOE-04 450.0 2.560E-04 600.0 3.167E-04 720.0 3.167E-04 Table 3.1-10: LOCA TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Containment Purge l.0601E+OO 2.1053E-01 7.2674E-01 Containment Leakage l.7972E+Ol 4.5696E+OO *1.7181E+OO ESF Leakage 2.4192E+OO 3.4498E+OO l.6848E+OO RWST Backleakage 2.1091E-02 6.2971E-02 3.9434E-02 Control Room Shine 0.385 Total 21.48 8.30 4.56 Acceptance Limit 25 25 5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 51 of 75 3 .2 Fuel Handling Accident The Fuel Handling Accident is evaluated as a-drop of a single fuel assembly in which 100% of the rods in the dropped assembly are assumed to fail. The analysis considers both a drop in the containment building without established containment integrity, and a drop in the Auxiliary Building with the Fuel Handling Area Exhaust Ventilation (FHAEV) in service. The source term for this event is described in Section 2.2.2; however, the nuclides available for release are limited to those isotopes present in the fuel rod gap listed in Table 2.2-6. As discussed in Section 2.2.5, the gap inventories shown in Table 2.2-6 are doubled to account for the presence of high bumup fuel rocis in the dropped assembly.

The analysis of the Fuel Handling Accident in both locations is modeled as a fuel assembly drop which occurs 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after reactor shutdown and results in the activity which escapes from the pools being released to the environment over a 2-hour period. The water in the pool is credited with a decontamination factor of285 for elemental iodine and 1.0 for organic iodine as discussed in Item 8 of Reference [4.2]. The pool water is assumed to retain 100% of the alkali metals. It is also assumed that the pool water will have no impact on the noble gases. No credit is taken for mixing or holdup in either the Auxiliary Building or the containment.

. However, iodine removal by the FHAEV system filters is modeled. The release location for the FHA in .

containment for the control room dose is assumed to be a point on the external containment surface closest to the control room intakes. For the drop in the Auxiliary Building, the FHAEV system discharges to the plant vent. The control room is assumed to be manually placed into the pressurization mode 20 minutes after tb.e start of the event by the plant operators. Control room ventilation parameters are listed in Table 2.1-1.

• Major inputs and assumptions important to this event ai;e provided in Table 3.2-1. The dose consequences are presented in Table 3.2-2 and Table 3.2-3. Note that the direct shine dose contribution from the control room

• filters conservatively reflects values from the LOCA event.

Table 3.2-1: Fuel Handling Inputs and Assumptions I

Input/Assumption Value Source Term FHA Inventory (Table 2.2-3)

Iodine Chemical Form 0% aerosol, 99.85% elemental, 0.15% organic

_Number of Fuel Assemblies Damaged 1 Percentage of Fuel Rods Failed 100%

No. of rods exceeding 6.3 kw/ft _above 54 GWD/MTU 150 High bumup multiplier applied to gap fractions 2.0 Water Level Above Damaged Fuel 23 feet Elemental - 285 Pool Decontamination Factors Organic - 1.0 Delay Before Fuel Movement 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> Containment Release Filtration 0%

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 52 of75 Input/Assumption Value Aerosol- 98.01%

Fuel Handling Area EXhaust Ventilation Filtration Elemental- 89.1%

Organic - 89.1%

Dose C(mversion Inputs:

.Atmospheric Dispersion Factors Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11&FGR12 Breathing Rates EAB/LPZ The breathing rates at the EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 104 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time 20 minutes (Manual)

Control Room Occupancy Factor Table 2.1-1 Table 3.2-2: Fuel Handling Accident - Containment Release TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Containment Release 3.5676E+OO 6.6745E-01 4.3492E+OO Control Room Shine 0.139 Total 3.57 0.67 4.49 Acceptance Limit 6.3 6.3 5

~I Red Wall' RWA-1313-015, Rev. 1 A~s;~l,_~tet.

D. C. Cook AST Radiological Analyses Technical Report Page 53 of 75 Table 3.2-3: Fuel Handling Accident - Auxiliary Building Release TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Auxiliary Building Release 6.7137E-Ol 1.3343E-01 l.3039E-Ol Control Room Shine 0.139 Total 0.68 0.14 0.27 Acceptance Limit 6.3 6.3 5 3 .3 Main Steam Lin_e Break This event consists of a break in one main steam line outside of containment in which the faulted steam generator (SG) completely depressurizes and instantly releases the initial contents of the steam generator secondary side to the environment. The plant cooldown continues by dumping steam with the intact steam generators. In addition to the release of nuclides that are initially present in the steam generator secondary side, leakage of primary coolant into the steam generator secondary side occurs at a rate equal to the proposed Tech. Spec. program limit of 0.25 gpm/SG.

This event does not result in fuel damage. Consequently, two iodine spike cases are considered. In the first (pre-accident spike) case, a reactor transient is assumed to occur prior to the MSLB in which the primary coolant iodine concentration has increased to the maximum Tech. Spec. value of 60 µCi/gm. In the second (concurrent spike) case, the iodine release rate into the primary coolant increases to a value that is 500 times greater than the 'normal' release rate that corresponds to the Tech. Spec. specific iodine limit of 1 µCi/gm.

This concurrent iodine spike is assumed to have a duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Inputs used in the development of the concurrent iodine spike appearance rate are shown in Table 3.3-1 and the 8-hour total iodine activity released into the reactor coolant is provided in Table 3.3-2. In both cases, the remaining non-iodine isotopes in the RCS listed in Table 2.2-4 are also available for release.

Leakage from the RCS into all of the steam generators, and steam release from the intact steam generators, continues until the RCS is cooled to 212 °F after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During this period, all of the noble gases and all of the nuclides which leak into the faulted steam generator are released directly to the environment without mitigation. Leakage into the intact steam generators mixes with the bulk fluid where a portion of the activity is released based upon the steaming rate and a partition coefficient. A partition coefficient of 100 is applied to the iodine nuclides, and the particulate release is limited by the moisture carryover. It is recognized that early in the transient, the water level in the intact steam generator secondary may be below the top of the tube bundle and the bulk water partitioning may not apply. In this case, a flashing fraction is calculated based upon the thermodynamic conditions in the reactor and secondary coolant. The portion of the primary-to-secondary leakage which flashes to vapor is assumed to be released directly to the environment without mixing. The iodine and particulate partition coefficients are applied to the unflashed portion. The tube bundles in the intact steam generators are assumed to be fully covered after 40 minutes.

The release locations from the faulted steam generator are selected to maximize the control room and offsite doses without regard to the location of the break with respect to the main steam isolation valves (MSIV) and

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 54 of75 without credit for MSIV closure. Releases from the intact steam generators occur from the PORVs/MSSVs.

The control room is automatically realigned into the pressurization mode (Table 2.1-1) upon receipt of a safety injection signal.

Parameters important to the analysis of the MSLB event are shown in Table 3.3-3, and the analysis results are provided in Table 3.3-4 and Table 3.3-5.

Table 3.3-1: Main Steam Line Break Iodine Appearance Rate Inputs and Assumptions Input/Assumption Value Letdown Flow Rate 132 gpm Identified RCS Leakage IOgpm Unidentified RCS Leakage 1 gpm RCS Mass 607,290.6 lbm (maximum)

I-131 Decay Constant 0.000060 min- 1 I-132 Decay Constant 0.005023 min- 1 I-133 Decay Constant 0.000555 min- 1 I-134 Decay Constant 0.013176 min- 1 1-135 Decay Constant 0.001748 min- 1 Table 3.3-2: Main Steam Line Break 500x Iodine Appearance 8-Hour Appearance Rate Isotope Production (Ci/min)

(Ci)

I-131 223.45 107,256 I-132 615.35 295,368 1-133 354.95 170,376 1-134 256.40 123,096 I-135 272.95 131,016

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D. C. Cook AST Radiological Analyses Technical Report Page 55 of 75 Table 3.3-3: Main Steam Line Break Inputs and Assumptions Input/Assumption Value Source Term Initial RCS Activity (Table 2.2-4)

Maximum Pre-Accident Iodine Spike Concentration 60 µCi/gm Dose Equivalent I-131 Concurrent Iodine Spike Appearance Rate 500x Equilibrium (Table 3.3-2)

Initial Steam Generator Iodine Source Term 0.1 µCi/gm Dose Equivalent I-131 Iodine Chemical Form 0% aerosol, 97% elemental, 3% organic Percentage of Fuel Rods Failed 0%

RCS Mass 466,141.5 lbm (minimum)

Steam Generator Secondary Liquid Mass 97,515.7 lbm/SG (minimum) 161,000 lbm/SG (maximum)

Intact Steam Generator Steam Release 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 4~6,000 lbm 2 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s: 1,186,000 lbm 8 - 24 _hours: 1,347,000 lbm Primary-Secondary Leak Rate 0.25 gpm to each steam generator Density Used for Leakage Volume-to-Mass Conversion 62.3 lbm/ft3 Duration of Intact SG Tube Uncovery After Reactor Trip 40 minutes Tube Leakage Flashing Fraction During Uncovery 0- 400 seconds: 8%

400-900 seconds: 6%

900-1700 seconds: 5.5%

1700 seconds-40 min: 4%

Time to Cool RCS to 212 °F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Intact Steam Generator Iodine Partition Coefficient Unflashed Leakage - 100 Flashed Leakage - 0 Intact Steam Generator Moisture Carryover Fraction 0.2% (Particulate Partition Coefficient= 500)

Dose Conversion Inputs:

Atmospheric Dispersion Factors Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11 & FGR 12 Breathing Rates EAB/LPZ The breathing rates at the EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

(.~ RWA-1313-015, Rev. 1

\ REid Walt

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D. C. Cook AST Radiological Analyses Technical Report Page 56 of 75 Input/Assumption Value Control Room 3.5 x 10-4 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time 70 seconds (Safety Injection)

Control Room Occupancy Factor Table 2.1-1 Table 3.3-4: Main Steam Line Break Pre-Accident Spike TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas 4.2923£-04 2.0284£-04 l.7912E-03 Pre-Accident Iodine Spike 1.6032£-01 7.8786£-02 4.7971£-01 Initial SG Secondary Iodine 8.0183E-02 l.1235E-02 7.5755E-01 Control Room Shine 0.139 Total 0.25 0.10 1.38 Acceptance Limit 25 25 5 Table 3.3-5: Main Steam Line Break Concurrent Spike TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas 4.2923£-04 2.0284£-04 l.7912E-03 Iodine Release 6.7612£-01 2.3694£-01 l.8792E+OO RCS Activity Release 8.1207£-02 3.9657E-02 2.0951E-Ol Initial SG Secondary Iodine 8.0183£-02 l.1235E-02 7.5755£-01 Control Room Shine 0.139 Total 0.84 0.29 2.99 Acceptance Limit 2.5 2.5 5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 57 of 75 3 .4 Steam Generator Tube Rupture The SGTR event represents an instantaneous rupture of a steam generator tube that releases primary coolant into the lower pressure secondary system. In addition to the break flow rate, primary-to-secondary leakage occurs at a rate equal to the proposed Tech. Spec. program limit of 0.25 gpm/SG to the ruptured and to each of the intact steam generators. All leakage flow into the ruptured steam generator is secured after 30 minutes.

Leakage into the intact steam generators continues until the RCS is cooled to 212 °F after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A portion of the break and leakage flow to the ruptured steam generator flashes to vapor based upon the thermodynamic conditions in the reactor and secondary coolant. The portion of the primary coolant that does flash in the steam generator secondary is released directly to the environment without mitigation. Although non-iodine particulates are not volatile and unlikely to be transported directly to the environment, the RCS fluid flashing fractions are conservatively applied to the particulate nuclides, which contributes significantly to the total dose.

The unflashed break and leakage flow mixes with the bulk water in the steam generator where the activity is released based upon the steaming rate and a partition coefficient. A steam generator partition coefficient of 100 is applied to the iodine nuclides, and the particulate release is limited by the moisture carryover. Prior to the reactor trip, an additional condenser partition coefficient of 100 is applied to the released activity. This same approach is applied to the primary-to-secondary leakage into the intact steam generators at the start of the event when the SG tube bundles are assumed to become uncovered following the reactor trip. After 40 minutes, flashing of the leakage flow is no longer applicable because the tube bundles have become fully submerged.

This event results in no fuel damage, and as such, two iodine spike cases are considered. In the pre-accident iodine spike case, a re~ctor transient is assumed to occur prior to the event in which the primary coolant iodine concentration has increased to the maximum Tech. Spec. value of 60 µCi/gm. In the concurrent iodine spike case, the iodine release rate into the primary coolant increases to a value that is 335 times greater than the

'normal' release rate that corresponds to the Tech. Spec. specific iodine limit of 1 µCi/gm. This concurrent iodine spike is assumed to have a duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Inputs used in the development of the concurrent iodine spike appearance rate are shown in Table 3.4-1, and the 8-hour total iodine activity released into the reactor coolant is provided in Table 3.4-2. In both cases, the remaining non-iodine isotopes in the RCS listed in Table 2.2-4 are also available for release. In addition to the activity transported into the steam generators from the primary coolant, the analysis considers the release of the iodine activity initially present in the steam generator secondary inventory.

Prior to the reactor trip, the activity is assumed to be released from the Steam Jet Air Ejector in the Turbine Building. Following the trip, the release location shifts to the PORVs/MSSVs. The control room is automatically realigned into the pressurization mode (Table 2.1-1) upon receipt of a safety injection signal.

Key inputs and assumptions applied in the analysis of the SGTR event are shown in Table 3.4-3, and the analysis results are provided in Table 3.4-4 and Table 3.4-5.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 58 of 75 Table 3.4-1: SGTR Iodine Appearance Rate Inputs and Assumptions Input/Assumption Value Letdown Flow Rate 132 gpm Identified RCS Leakage IOgpm Unidentified RCS Leakage 1 gpm RCS Mass 607,290.6 lbm (maximum)

I-131 Decay Constant 0.000060 min" 1 I-132 Decay Constant 0.005023 min* 1 I-133 Decay Constant 0.000555 min" 1 I-134 Decay Constant 0.013176 min" 1 I-135 Decay Constant 0.001748 min* 1 Table 3.4-2: SGTR 335x Iodine Appearance 8-Hour Appearance Rate Isotope Production (Ci/min)

(Ci)

I-131 149.71 71,861 I-132 412.28 197,894 I-133 237.82 114,154 I-134 171.79 82,459 I-135 182.88 87,782

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 59 of 75 Table 3.4-3: Steam Generator Tube Rupture Inputs and Assumptions Input/Assumption Value Source Term Initial RCS Activity (Table 2.2-4)

Maximum Pre-Accident Iodine Spike Concentration 60 µCi/gm Dose Equivalent I-131 Concurrent Iodine Spike Appearance Rate 335x Equilibrium (Table 3.4-2)

Initial Steam Generator Iodine Source Term 0.1 µCi/gm Dose Equivalent I-131 Iodine Chemical Form 0% aerosol, 97% elemental, 3% organic Percentage of Fuel Rods Failed 0%

RCS Mass 466,141.5 lbm (minimum)

Steam Generator Secondary Liquid Mass 97,515.7 lbm/SG (minimum) 161,000 lbm/SG (maximum)

Intact Steam Generator Steam Release 0 - 30 min. - 198,515 lbm 30 min. - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 314,432 lbm 2 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s: 1,367,475 lbm 8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: 1,347,000 lbm Ruptured Steam Generator Steam Release 0 - 30 min. - 66,171 lbm Pre-Trip Total Steam Flow Rate Through Condenser 17,153,800 lbm/hr Time of Reactor Trip 101 seconds Primary-Secondary Leak Rate 0.25 gpm to each steam generator Density Used for Leakage Volume-to-Mass Conversion 62.3 lbm/ft3 Ruptured Tube Break Flow 146,704 lbm Duration of Ruptured Tube Break Flow 30 minutes Break Flow Flashing Fraction Pre-Trip 0-100 seconds: 19%

Post-Trip:

100- 500 seconds: 8%

500-1000 seconds: 6%

1000-1800 seconds: 5.5%

Duration of Intact SG Tube Uncovery After Reactor Trip 40 minutes Intact Tube Leakage Flashing Fraction During Uncovery 0-100 seconds: 19%

100- 500 seconds: 8%

500-1000 seconds: 6%

1000-1800 seconds: 5.5%

1800 seconds-40 min: 4%

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 60 of 75 Input/Assumption Value Time to Cool RCS to 212 °F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Steam Generator Iodine Partition Coefficient Unfiashed Leakage - 100 Flashed Leakage - 0 Condenser Partition Coeffic.ient 100 Steam Generator Moisture Carryover Fraction 0.2% (Particulate Partition Coefficient = 500)

Dose Conversion Inputs:

Atmospheric Dispersion Factors Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11&FGR12 Breathing Rates EAB/LPZ The breathing rates at the EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 104 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time* 394.74 seconds (Safety Injection)

Control Room Occupancy Factor Table 2.1-1 Table 3.4-4: Steam Generator Tube Rupture Pre-Accident Spike TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas 4.1808E-02 7.8965E-03 2.3863E-02 Pre-Accident Iodine Spike 4.1154E+OO 7.4309E.:01 3.7895E+OO Initial SG Secondary Iodine l.8818E-03 8.3936E-04 2.6022E-03 Control Room Shine 0.139 Total 4.16 0.76 3.96 Acceptance Limit 25 '

25 5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 61of75 Table 3.4-5: Steam Generator Tube Rupture Concurrent Spike TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas 4.1808E-02 7.8965E-03 2.3863E-02 Iodine Release 4.0639E-01 8.8981E-02 2.4586E-01 RCS Activity Release 2.0463E+OO 3.6688E-01 l.8432E+OO Initial SG Secondary Iodine l.8818E-03 8.3936E-04 2.6022E-03 Control Room Shine 0.139 Total 2.50* 0.47 2.26 Acceptance Limit 2.5 2.5 5

• Calculated value is 2.497 rem which is rounded up to 2.50.

3.5 Locked Rotor The Locked Rotor dose analysis is defined by the 11 % of the fuel rods which become damaged by the event.

Radionuclides released from the fuel are instantaneously and homogeneously distributed throughout the primary coolant. Noble gases are released directly to the environment, and the remaining isotopes are transported to the steam generators at a rate of 1 gpm. The core source term from Table 2.2-2 is applicable to

• this event, and the fraction of these activities available for release into the coolant are based upon the gap inventory fractions shown in Table 2.2-6 and the assembly radial peaking factor of 1.65. To acc9unt for fuel rods contained in two of the fuel assemblies which exceed the burnup limits of Footnote 11 of Reference [4.1],

the gap inventory of all of the rods in these two assemblies are assumed to be twice those listed in Table 2.2-6 as discussed in Section 2.2.5. As such, with 193 fuel assemblies in the core, the effective core-wide multiplier on the gap inventory fractions is 1 + (2/193) = 1.0104. Since the fuel failure fraction is applied to the entire core source term, 11 % of the rods in both the standard and high burnup assemblies are assumed to fail, and the assembly peaking factor is conservatively applied to all of the failed rods in the core.

• During the first 40 minutes of the event, the water level on the secondary side of the steam generators is assumed to be below the top of the tube bundles. During this time, a portion of the primary-to-secondary leakage flashes to vapor based upon the thermodynamic conditions of the reactor and secondary coolant.

Nuclides contained in the flashed tube leakage are released to the environment without mitigation. The unflashed leakage mixes. with the bulk water in the steam generators and is released as a function of the steaming rate and the partition coefficients. After 40 minutes, all of the leakage is treated as unflashed, which continues until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the RCS temperature is cooled to 212 °F.

Since the quantity of the fission products released from the failed fuel dominates the RCS activity during the event, the initial nuclide concentration in the RCS prior to the event is not considered. However, the analysis does include the dose contribution from the release of iodine initially present in the steam generator secondary side. All releases occur from the PORVs/MSSVs, which are located in the Main Steam Enclosures. For this

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Red Wolf' RWA-1313-015, Rev. 1

,A§.-nt:lat~

D. C. Cook AST Radiological Analyses Technical Report Page 62 of 75 event, the control room ventilation system remains in the normal alignment without filtration or recirculation until manually placed into the pressurization mode after 20 minutes.

Major inputs and assumptions applicable to the Locked Rotor event are listed in Table 3.5-1. The analysis results are presented in Table 3.5-2.

Table 3.5-1: Locked Rotor Inputs and Assumptions Input/Assumption Value Source Term Core Inventory (Table 2.2-2)

Fuel Rod Gap Fractions I-131 - 0.08 Kr 0.10 Other Noble Gases - 0.05 Other Halogens - 0.05 Alkali Metals - 0.12 Percentage of Fuel Rods Failed 11%

Fuel Rod Peaking Factor 1.65 No. of rods exceeding 6.3 kw/ft above 54 GWD/MTU 150 rods in two assemblies High bumup multiplier applied to gap fractions 1.0104 Initial Steam Generator Iodine Source Term 0.1 µCi/gm Dose Equivalent I-131 Iodine Chemical Form 0% aerosol, 97% elemental, 3% organic RCS Mass 466,141.5 lbm (minimum)

Steam Generator Secondary Liquid Mass 97,515.7 lbm/SG (minimum) 161,000 lbm/SG (maximum)

Primary-Secondary Leak Rate 1.0 gpm to all steam generators Density Used for Leakage Volume-to-Mass Conversion 62.3 lbm/ft3 Secondary Steam Release 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 460,000 lbm 2 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s: 1,256,000 lbm 8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: 1,347,000 lbm Time to Cool RCS to 212 °F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Duration of SG Tube Uncovery Fallowing Reactor Trip 40 minutes Intact Tube Leakage Flashing Fraction During Uncovery 0- 400 seconds: 8%

400-900 seconds: 6%

900-1700 seconds: 5 .5%

1700 seconds-40 min: 4%

Steam Generator Iodine Partition Coefficient Unflashed Leakage - 100 Flashed Leakage - 0

. ,,,,:: I Red Walt: RWA-1313-015, Rev. 1

• A~~;;~~~:

D. C. Cook AST Radiological Analyses Technical Report Page 63 of75 Input/Assumption Value Steam Generator Moisture Carryover Fraction 0.2% (Particulate Partition Coefficient= 500)

Dose Conversion Inputs:

Atmospheric Dispersion Factors Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11 & FGR 12 Breathing Rates EAB/LPZ The breathing rates at the EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 10-4 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time 20 minutes (Manual)

Control Room Occupancy Factor Table 2.1-1 Table 3.5-2: Locked Rotor TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas Dose 5.2878E-01 l.5559E-01 4.4546E-01 Non-Noble Gas Dose- Iodine 9.1900E-01 5.1347E-Ol 2.9955E+OO Non-Noble Gas Dose -Alkali 3.6098E-01 l.2094E-01 l.1508E+OO Metals Initial SG Secondar)r Iodine l.4821E-03 7.3853E-04 3.1709E-03 Control Room Shine 0.139 Total 1.82 0.80 4.74 Acceptance Limit 2.5 ' 2.5 5 3.6 Control Rod Ejection The Control Rod Ejection event involves a reactivity insertion that produces a short, rapid core power level increase which results in fuel rod damage and localized melting. For this event, a larger fraction of the core inventory is released from the damaged fuel than that identified in Table 2.2-6. Two separate release pathways are evaluated:, a release from containment and a release from the secondary system. In both cases, 10% of the

~~

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~,._.-,;n_c~~ D. C. Cook AST Radiological Analyses Technical Report Page 64 of 75 noble gases and 10% of the iodines in the core (Table 2.2-2) are available for release from the fuel gap of the damaged fuel rods. In addition, 12% of the alkali metals are also assumed to be located in the fuel rod gap.

For releases from containment, 10% of the fuel rods in the core are breached and 0.25% of the fuel experiences melting. The activity in the fuel rod gap of the damaged fuel is instantaneously and homogeneously mixed throughout the containment atmosphere. In addition, the gap inventory fractions are increased by a factor of 1.0104 to account for high bumup fuel as discussed in Sections 2.2.5 and 3.5. Moreover, 100% of the noble gases and 25% of the iodines in the melted fuel are also added to the fission product inventory in containment.

No credit is taken for removal by containment sprays or for deposition of elemental iodine on containment surfaces. Natural deposition of aerosols in containment is assumed to occur beginning 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of the event. Activity is released from containment at the proposed Tech. Spec. leak rate. The release from the containment is based upon an atmospheric dispersion factor assuming a diffuse release from the containment surface. For conservatism, the release of iodine initially present in the steam generator secondary side is also considered to address any supplemental cooldown by the steam generators for this event.

For releases from the secondary system, 10% of the fuel rods in the core are breached and 0.25% of the fuel experiences melting. Activity released from the fuel is completely dissolved in the primary coolant and is available for release to the secondary system. The gap activity is increased by a factor of 1.0104 to address fuel rods with bumups that exceed the values of Footnote 11 of Reference [4.1]. In this case, 100% of the noble gases and 50% of the iodines in the melted fuel is also released into the reactor coolant. The noble gases are assumed to be released directly to the environment, and the remaining fission products are transported to the steam generators at the Tech. Spec. steam generator program leakage limit of 1 gpm. At the beginning of the event, a portion of the primary-to-secondary leakage is assumed to flash to vapor based upon the thermodynamic conditions of the reactor and secondary coolant, and the flashed leakage is released directly to the environment without mitigation. The unflashed portion of the tube leakage mixes with the bulk fluid in the steam generator secondary and becomes vapor at a rate that is a function of the steaming rate and the partition coefficients. After 40 minutes, the water level in the steam generator is assumed to fully cover the tube bundles, and all of the primary-to-secondary leakage is treated as unflashed. The leakage continues until steam releases are terminated when the RCS temperature is cooled to 212 °F at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the large amount of fission products introduced into the reactor coolant by failed fuel, the initial activity of the RCS prior to the event is not considered. However, the dose contribution from the iodine activity initially present in the steam generator secondary is included in the analysis. All releases from the secondary system occur from the PORVs/MSSVs.

The control room ventilation system is automatically realigned into the pressurization mode following receipt of a safety injection signal. Key control room ventilation parameters applied in the analysis are shown in Table 2.1-1. Other inputs and assumptions important to the Control Rod Ejection event are provided in Table 3.6-1. The dose consequences for this event are summarized in Table 3.6-2 and Table 3.6-3.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 65 of 75 Table 3.6-1: Control Rod Ejection Inputs and Assumptions Input/Assumption Value Source Term Core Inventory (Table 2.2-2)

Fuel Rod Gap Fractions Noble Gases - 0.10 Other Halogens - 0.10 Alkali Metals - 0.12 Percentage of Fuel Rods Failed 10%

Percentage of fuel that experience fuel melting 0.25%

No. of rods exceeding 6.3 kw/ft above 54 GWD/MTU 150 rods in two assemblies High Bumup multiplier applied to gap fractions 1.0104 Fuel Rod Peaking Factor 1.65 Initial Steam Generator Iodine Source Term 0.1 µCi/gm Dose Equivalent I-131 Iodine Chemical Form - Secondary Release 0% aerosol, 97% elemental, 3% organic Iodine Chemical Form - Containment Release 95% aerosol, 4.85% elemental, 0.15% organic Containment Volume 1,066,352 ft 3 Containment Leakage Rate 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.18 %/day 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 30 days 0.09 %/day Containment Leakage Filtration 0%

Elemental Iodine - None Natural Deposition in Containment Aerosols - 0.1 hr" 1 after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Iodine/Particulate Removal by Containment Sprays None RCS Mass 466,141.5 lbm (minimum)

Steam Generator Secondary Liquid Mass 97,515.7 lbm/SG (minimum) 161,000 lbm/SG (maximum)

Primary-Secondary Leak Rate 1.0 gpm to all steam generators Density Used for Leakage Volume-to-Mass Conversion 62.3 lbm/ft3 Secondary Steam Release 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s: 460,000 lbm 2 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s: 1,256,000 lbm

. 8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: 1,347,000 lbm Time to Cool RCS to 212 °F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Duration of SG Tube Uncovery Following Reactor Trip 40 minutes

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 66 of 75 Input/Assumption Value Tube Leakage Flashing Fraction During Uncovery 0- 400 seconds: 8%

400-900 seconds: 6%

900-1700 seconds: 5.5%

1700 seconds-40 min: 4%

Steam Generator Iodine Partition Coefficient Unflashed Leakage - 100 Flashed Leakage - 0 Steam Generator Moisture Carryover Fraction 0.2% (Particulate Partition Coefficient= 500)

Dose Conversion Inputs:

Atmospheric Dispersion Factors Offsite *Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11&FGR12 Breathing Rates EAB/LPZ The breathing rates at the EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 10-4 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time 120 seconds (Safety Injection)

Control Room Occupancy Factor Table 2.1-1

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 67 of 75 Table 3.6-2: Control Rod Ejection Secondary Release TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Noble Gas, Cladding Failure 9.6153E-01 2.8289E-01 8.0333E-01 Noble Gas, Fuel Melt 2.3793E-01 7.0005E-02 l.9879E-01 Non-Noble Gas, Cladding Failure, l.1615E+OO 6.5310E-01 l.8378E+OO Iodine Non-Noble Gas, Cladding Failure, 3.2816E-01 l.0995E-01 2.0428E-01 Alkali Non-Noble Gas, Fuel Melt 1.4520E-01 8.1647E-02 2.2976E-01 Initial SG Secondary Iodine 1.4821E-03 7.3853E-04 l.9539E-03 Control Room Shine 0.139 Total 2.84 1.20 3.42 Acceptance Limit 6.3 6.3 5 Table 3.6-3: Control Rod Ejection Containment Release TEDE Dose Results EAB LPZ Control Room Release (rem) (rem) (rem)

Cladding Failure 3.8824E+OO 2.6021E+OO l.5378E+OO Fuel Melt l.8512E-01 l.1633E-01 6.8727E-02 Initial SG Secondary Iodine l.4821E-03 7.3853E-04 l.9539E-03 Control Room Shine 0.139 Total 4.07 2.72 1.75 Acceptance Limit 6.3 6.3 5 3.7 Waste Gas Decay Tank Rupture This event involves a major rupture of one of the Waste Gas Decay Tanks that causes the entire contents of the tank to be released directly to the environment. Reg. Guide 1.183 does not provide any guidance relative to the Waste Gas Decay Tank Rupture event. Guidelines for the WGDT analyses are given in Branch Technical Position 11-5 (BTP 11-5) of the Standard Review Plan (Reference [4.24]), with additional instruction available from Regulatory Issue Summary 2006-04 (Reference [4.2]). The activity in the tank is assumed to be equal to the noble gas content of the reactor coolant system during nonnal operation. This source term is derived from plant operation with 1% fuel defects which has operated for a sufficient length of time to achieve equilibrium radioactive concentrations in the RCS. The equilibrium RCS concentrations are then adjusted to 100/E as discussed in Section 2.2.3. For the WGDT rupture event, the entire noble gas inventory of the RCS is then assumed to be stripped and placed into a single tank. This inventory is conservatively calculated by taking the

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 68 of 75 noble gas specific activities in the RCS from Table 2.2-4 and multiplying by the maximum RCS mass as shown in Table 3.7-1. The total activity released is determined to be equal to 59,256.4 Curies Dose Equivalent Xe-133, which exceeds the single WGDT licensing limit of 43,800 Curies. The WGDT failure simulates a major tank rupture in which the entire contents of the tank are instantaneously released directly to the environment. No credit is taken for hold-up, dilution, or decay in the Auxiliary Building.

Section B. l .C of Reference [4.24] requires that the release to the environment occur through a pathway not normally used for planned releases and will require a reasonable time to detect and take remedial action to terminate the release. This condition is satisfied in the analysis by assuming that normal Auxiliary Building ventilation is not in service. As such, discharges to the plant vent are not ensured, and gases escaping from the ruptured tank are permitted to be released through building openings with the highest atmospheric dispersion factors. For the offsite dose, the most limiting atmospheric dispersion factor is from the north Auxiliary Building normal ventilation intake, and for the control room dose, the limiting release point is the south normal ventilation intake. The control room ventilation system is assumed to remain in the normal alignment since a safety injection signal, which is required to automatically place the system in the pressurization mode, is not received for this event.

Neither BTP 11-5 nor RIS 2006-04 require dose evaluations at the LPZ or for the control room for a WGDT failure; however, both evaluations are completed in this effort for consistency and completeness. Item 11 of RIS 2006-04 allows the use of existing current licensing basis acceptance criterion of 500 mrem whole body at the BAB when this event is not submitted as part of the AST implementation. Consequently, the results of the WGDT failure are evaluated against the existing 500 mrem acceptance criterion for both the EAB and LPZ.

Reg. Guide 1.183 does identify that the criterion for the control room dose is provided in 10 CFR 50.67, which establishes the dose limit for the control room as 5 rem TEDE. This value is applied here.

A summary of the inputs and assumptions to the WGDT rupture analysis are given in Table 3.7-2, and the dose consequences are shown in Table 3.7-3.

Table 3.7-1: WGDT Source Term RCS Activity Total Activity Nuclide

(µCi/g) (Ci)

Kr-85m 5.204E-01 l.433E+02 Kr-85 2.385E+Ol 6.570E+03 Kr-87 3.299E-01 9.087E+Ol Kr-88 9.148E-01 2.520E+02 Xe-131m l.600E+OO 4.407E+02 Xe-133m l.423E+OO 3.920E+02 Xe-133 l.037E+02 2.857E+04 Xe-135m 2.138E-01 5.889E+Ol Xe-135 3.36IE+OO 9.258E+02 Xe-138 2.292E-01 6.314E+Ol

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 69 of 75 Table 3.7-2: WGDT Rupture Inputs and Assumptions Input/Assumption Value Source Term Table 3.7-1. Equal to 59,256.4 D.E. Xe-133 RCS Mass 607,290.6 lbm (275,460,950 gm) (maximum)

Tanlc Volume 500 ft3 (arbitrary) 1,000,000 cfm (conservatively high to simulate Tanlc Release Rate an instantaneous release)

Dose Conversion Inputs:

Atmospheric Dispersion Factors I

Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11 & FGR 12 Breathing Rates EAB/LPZ The breathing rates at the. EAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

Control Room 3.5 x 104 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time Not Isolated Control Room Occupancy Factor Table 2.1-1 Table 3. 7-3: WGDT Rupture Dose Results EAB LPZ Control Room Release (rem whole body) (rem whole body) (rem TEDE)

WGDT Rupture 0.22 0.04 0.09 Acceptance Limit 0.5 0.5 5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 70 of 75 3.8 Volume Control Tank Rupture The analysis of the VCT rupture conservatively assumes a failure of the VCT just prior to venting which releases the accumulated noble gases in the liquid and vapor phases of the tank. In addition, the noble gases within the fluid of the letdown line entering the tank continue to be released for an additional 15 minutes following the tank rupture. The tank and letdown line activities are calculated from RCS equilibrium noble gas concentrations based upon 1% failed fuel with no adjustment for 100/E. This conservative source term is consistent with Section B.l.B of Branch Technical Position 11-5. Accumulated gases in the VCT vapor space are calculated using gas stripping fractions, and the entire amount of dissolved gases in the VCT liquid space and the continued letdown flow are available for release after the tank ruptures. The resulting source term is shown in Table 3.8-1.

Further guidance for the analysis of this event is taken from BTP 11-5 since Reg. Guide 1.183 does not address waste system failures. Pathways for the noble gases to escape from the VCT to the environment include those which are not normally used for planned releases. For this event, the release location is the south normal ventilation intake for the control room dose, and is the north normal ventilation intake for the offsite doses based upon the most limiting atmospheric dispersion factors. Releases I

from the normal supply vents are possible when the Auxiliary Building ventilation system is not in service. This analytical approach is consistent with the guidance of Section B. l.C of BTP 11-5. The analysis assumes that this release is made directly to the environment, without taking credit for hold-up, dilution, or decay in the Auxiliary Building.

Similarly, while the control room ventilation system filters would have no impact on this event, the control room ventilation system remains in the normal system alignment described in Section 2.1. Significant inputs and assumptions applied in the analysis ofthis event are listed in Table 3.8-2.

The acceptance criteria for this event is based upon the guidance of Item 11 of RIS 2006-04, which permits the use of the existing current licensing basis acceptance criterion of 500 mrem whole body at the BAB when AST is not implemented for this event. Without specific guidance for the LPZ, the 500 mrem whole body limit is also applied to the LPZ dose for this event. The acceptance limit of 5 rem TEDE to operators in the control room from 10 CFR 50.67 is applied based upon Section 4.4 of Reg. Guide 1.183. The dose consequences for the VCT rupture event are provided in Table 3.8-3.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 71of75 Table 3.8-1: VCT Source Term VCTGas VCTLiquid Letdown VCT Equilibrium Phase Phase Flow Total Release Nuclide RCS Activity Activity Activity Activity

(µCi/g)

(Ci) (Ci) (Ci) (Ci)

Kr-85m 1.366E+OO l.596E+02 l.021E+Ol l.012E+Ol l.799E+02 Kr-85 6.261E+Ol l.835E+04 4.679E+02 4.639E+02 l.928E+04 Kr-87 8.660E-01 3.955E+Ol 6.472E+OO 6.416E+OO 5.244E+Ol Kr-88 2.401E+OO 2.070E+02 l.794E+Ol l.779E+Ol 2.427E+02 Xe-131m 4.199E+OO 8.716E+02 3.138E+Ol 3.lllE+Ol 9.341E+02 Xe-133m

• 3.734E+OO 7.125E+02 2.791E+Ol 2.766E+Ol 7.681E+02 Xe-133 2.723E+02 5.421E+04 2.035E+03 2.017E+03 5.826E+04 Xe-135m 5.612E-Ol 5.806E+OO 4.194E+OO 4.158E+OO l.416E+Ol Xe-135 8.821E+OO l.200E+03 6.593E+Ol 6.535E+Ol 1.33lE+03 Xe-138 6.017E-01 5.770E+OO. "4.497E+OO 4.458E+OO l.473E+Ol Table 3.8-2: VCT Rupture Inputs and Assumptions In put/Assumption Value Source Term Table 3.8-1 Tank Volume Liquid Volume 267 ft3 Tank Volume Vapor Volume 500 ft3 (arbitrary) 1,000,000 cfm (conservatively high to simulate VCT Release Rate -

an instantaneous release)

Letdown Flow Rate 132 gpm Letdown Isolation Time 15 minutes Dose Conversion Inputs:

Atmospheric Dispersion Factors J Offsite Table 2.3-6 and Table 2.3-7 Onsite Table 2.3-3 and Table 2.3-7 Dose Conversion Factors FGR 11 & FGR 12 Breathing Rates EAB/LPZ The breathing rates at the BAB and LPZ are taken from Position 4.1.3 of Reference [4.1] and shown in Table 2.3-8.

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 72 of 75 Input/Assumption Value Control Room 3.5 x 104 m3/sec Control Room Ventilation System Parameters Table 2.1-1 Control Room Isolation Time Not Isolated .

Control Room Occupancy Factor Table 2.1-1 Table 3.8-3: VCT Rupture Dose Results EAB LPZ Control Room Release (rem whole body) (rem whole body) (rem TEDE)

VCTRupture 0.33 0.06 0.13 Acceptance Limit 0.5 0.5 5

RWA-1313-015, Rev. 1 D. C. Cook AST Radiological Analyses Technical Report Page 73 of 75 3 .9 Results Summary The results of the D. C. Cook radiological analyses using the AST methodology are summarized in Table 3.9-1.

Table 3.9-1: Dose Results Summary Control EABDose LPZDose Event Room (rem TEDE) (rem TEDE)

(rem TEDE)

LOCA 21.48 8.30 4.56 MSLB Pre-accident Iodine Spike 0.25 0.10 1.38 SGTR Pre-accident Iodine Spike 4.16 0.76 3.96 Acceptance Criteria 25 25 5 FHA - Containment Release 3.57 0.67 4.49 FHA - Auxiliary Building Release 0.68 0.14 0.27 Control Rod Ejection - Containment 4.07 2.72 1.75 Control Rod Ejection - Secondary 2.84 1.20 3.42 Acceptance Criteria 6.3 6.3 5 MSLB Concurrent Iodine Spike 0.84 0.29 2.99 SGTR Concurrent Iodine Spike 2.50<2) 0.47 2.26 Locked Rotor 1.82 0.80 4.74 Acceptance Criteria 2.5 2.5 5 WGDT Rupture o.2z< 1l 0.04(l) 0.09 VCTRupture 033(l) 0.06(l) 0.13 Acceptance Criteria 0.5(l) 1 o.5< l 5 (1) EAB and LPZ results and acceptance criteria for the WGDT and VCT events are whole body doses (2) Calculated value is 2.497 rem which is rounded up to 2.50.

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