View source for AEP-NRC-2016-23, Calculation RWA-1313-001, Rev. 1, AST Radiological Analysis Input Parameter Development.

{{Adams
| number = ML16169A116
| issue date = 02/12/2016
| title = Calculation RWA-1313-001, Rev. 1, AST Radiological Analysis Input Parameter Development.
| author name = Pope M
| author affiliation = Red Wolf Associates, PLLC
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000315, 05000316
| license number = 
| contact person = 
| case reference number = AEP-NRC-2016-23
| document report number = RWA-1313-001, Rev 1
| package number = ML16169A129
| document type = Calculation
| page count = 22
}}

=Text=
{{#Wiki_filter:Cook Nuclear Plant AST Radiological Analysis Input                   RWA-1313-001, Rev. 1 Parameter Development                                                 Page 1 of 6 Calculation Number:                                                                  Revision:      Total No. Pages (incl. cover):
RWA-1313-001                                                                               1                22 Calculation

==Title:==
Cook Nuclear Plant AST Radiological Analysis Input Parameter Development Calculation Quality Classification:        ~ Safety           D   Non-safety Verification Performed:          ~Review           D    Alternate Calculation         D   Testing

== Description:==

This calculation compiles plant input parameters for use in the Cook radiological analysis which is performed using the Alternative Source Term methodology.
Revision 1 is performed to change the basis for the fraction of the primary-to-secondary leakage which flashes to steam on the secondary side of the steam generators. In addition, clarifying information has been provided regarding the VCT rupture isolation time.
Digital Signature or Printed Name, Signature, and Date:                                        Applicable Sections:
Preparer: Mark Pope                                                                                      All Digitally signed by Mark Pope Date:2016.02.12 16:04:02 EST Reviewer: Nathan Block                                                                                   All Digitally signed by Nathan Block Date: 2016.02.12 16:05:58 EST Approver: Joe Sinodis                                                                                    All Digitally signed by Joe Sinodis Date:2016.02.12 16:12:48 EST

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Cook Nuclear Plant AST Radiological Analysis Input                                              RWA-1313-001, Rev. 1 Parameter Development                                                                        Page 2 of 6 Table of Contents 1     Purpose ............................................................................................................................................ 3 2     Methodology ........................................................... :....................................................................... 3 3     Inputs .............................................................................................................................................. 3 4     Assumptions .................................................................................................................................... 4 5     Calculations ..................................................................................................................................... 4 6     Conclusions ..................................................................................................................................... 4 7     References ............................. ~ ......................................................................................................... 4 Attachment A: Radiological Analysis Input Parameters ................................................................ Al-A12 Attachment B: Owner's Review Comments ..................................................................................... Bl-B4 List of Figures Figure Al: Containment Sump Temperature Profile ..........................................................:.-............... A12

Cook Nuclear Plant AST Radiological Analysis Input              RWA-1313-001, Rev. 1 Parameter Development                                        Page 3 of 6 1 Purpose This calculation compiles plant input parameters for use in the Cook radiological analysis. The calculation serves two main purposes. First, it provides a single, referenceable source for dose analysis parameters, which ensures that a consistent set of inputs are applied throughout the individual event calculations. As such, the values and sources of the specific parameters are cited and reviewed one time for the entire dose analysis project. Secondly, this calculation provides a comprehensive list of plant parameters that are important to the plant radiological consequences. This document can serve as a convenient tool to assist plant personnel in determining the impact of future design changes and in assessing the consequences of those changes when.
performing 50.59 evaluations.
2 Methodology This calculation is primarily a list of parameters with cited references. In some cases, minor mathematical computations may be performed. Some guidance is taken from Regulatory Guide 1.183 (Reference 7 .1) since the radiological analyses in which the parameters are used is performed using the alternative source term methodology. It is important. to note that the value of the parameter shown reflects the value from the applicable reference and does not necessarily indicate the value of the parameter used in the event calculations.
Any comments included with the parameter should be given equal weight with the value of the parameter itself. These comm~nts may modify the numerical value, describe how it is applied in the event calculation, identify the direction of conservatism, or cite the regulatory requirement for why it is included. Note that when a specific Cook unit is not specified, the input values apply to both Units 1 and 2, and are considered bounding values for both units.
3 Inputs Inputs to this calculation are not conventional design inputs. All parameters which are important to the consequences of the radiological analysis are identified, which may include:
* Design requirements
* Procedural controls
* System alignments
* Licensing limits
* Calculation and analysis results
* Operator actions
* Plant measurements

Cook Nuclear Plant AST Radiological Analysis Input .           RWA-1313-001, Rev, 1 Parameter Development                                       Page 4 of 6 4 Assumptions Some of the parameter values listed in this calculation are simply assumptions and are not specifically identified as such. However, once applied in the event analyses, these assumptions become design limits. For example, the backleakage rate to the RWST is somewhat arbitrarily selected in this document to create operational margin to actual measured leak rates. However, since it is defined here and applied in the LOCA analysis, this assumption becomes the leak rate limit.
5 Calculations The radiological analysis inputs are compiled and presented in Attachment A.
6 Conclusions The data provided in Attachment A provides a comprehensive set of inputs that are acceptable for use in performing safety-related dose analyses.
7 References 7.1    USNRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.
7.2    D. C. Cook Unit 1 Renewed Facility Operating License No. DPR-58 through Amendment 328.
7.3    D. C. Cook Unit 2 Renewed Facility Operating License No. DPR-74 through Amendment 310.
7.4    D. C. Cook Unit 1 Technical Specifications through Amendment 328.
7.5    D. C. Cook Unit 2 Technical Specifications through Amendment 310.
7.6    D.C. Cook Updated Final Safety Analysis Report, Revision 26.
7.7    Safety Evaluation Report, "Donald C. Cook Nuclear Plant, Unit 1 - Issuance of Amendment 273 Regarding Measurement Uncertainty Recapture Power Uprate (TAC NO. MB5498)", December 20, 2002 (ML023470126).
7.8    Indiana Michigan Power letter AEP:NRC:3902-0l, "Donald C. Cook Nuclear Plant Unit 2, Review of Draft Safety Evaluation for Measurement Uncertainty Recapture Power Uprate (TAC NO. MB6751 )'',
April 25, 2003 (ML031270262).
7.9    Safety Evaluation Report, "Donald C. Cook Nuclear Plant, Unit 2 - Issuance of Amendment 259 Regarding Measurement Uncertainty Recapture Power Uprate (TAC NO. MB6751)", May 2, 2003 (ML030990094).
7.10 Drawing OP-1-5147A-39, "Flow Diagram, Containment Ventilation, Unit No. l".
7 .11 Drawing OP-2-514 7A-46, "Flow Diagram, Containment Ventilation, Unit No. 2".
7.12 Drawing OP-12-5148-63, "Flow Diagram, Auxiliary Building Ventilation Units #1 and #2".

Cook Nuclear Plant AST Radiological Analysis Input          RWA-1313-001, Rev. 1 Parameter Development                                      Page 5 of 6 7.13 Drawing OP-1-5149-49, "Flow Diagram, Control Room Ventilation, Unit No. 1".
7.14 Drawing OP-2-5149-58, "Ffow Diagram, Control Room Ventilation, Unit No. 2".
7.15. Procedure 1-0HP-4030-127-041, "Refueling Integrity'', Rev. 30.
7.16 Procedure 2-0HP-4030-227-041, "Refueling Illtegrity", Rev. 32.
7.17 Engineering Control Package ECP 1-05-01, "Precautions, Limitations, and Setpoints, Unit 1", Rev. 20.
7.18 Engineering Control Package ECP 2-05-01, "Precautions, Limitations, artd Setpoints, Unit 2", Rev. 17.
7.19 Procedure PMP-5076-ULR-001, "Reactor Coolant System Leakage Monitoring Program'', Rev. 4.
7.20 Procedure OHI-4032, "Leakage Monitoring Program", Rev. 15.
7.21 Calculation PRA-DOSE-011, "Containment Sprayed Volumes, Unsprayed Volumes, and Average Spray Fall Heights", Rev. 0.
7.22 Letter Report RWA-L-1313-002, 

==Subject:==
Review of Calculation PRA-DOSE-011, "Containment Sprayed Volumes, Unsprayed Volumes, and Average Spray Fall Heights," July 1, 2013.
}.23 Calculation MD-12-HV-005-N, "Control Room Pressure Boundary Volume", Rev. 0.
7.24 Calculation CN-CRA-99-55, "Donald C. Cook Steam *Generator Tube Rupture T&H Analysis for NUREG-1465 Dose Project- Revised", Rev. l.
7.25 Calculation CN-CRA-99-047, "D.C. Cook Units 1 & 2 Steam Releases for Radiological Dose Calculation", Rev. 0.
7.26 Calculation TH-00-03, "D. C. Cook Unit 2 Steam Generator Tl,lbe Rupture with Operator Actions", Rev. 0.
7.27 Calculation SD-990618-003, "Containment Net free Volume", Revision 2.
7.28 Calculation MD-1-SGRP-022-N, "Cook RSG-Licensing Data for FTI B&W Replacement Steam Generators (B&W Calculation No. 222-7803-Al9, Rev. 2)", Rev. 0.
7.29 Calculation MD-12-CTS-118-N, "Containment Spray System and Recirculation Sump Minimum and Maximum pH", Rev. 4:
7.30 Design Information Transmittal DIT SGRP 99035-00 Rev, 0, "Reactor Coolant System Volumes",
November 2, 1999.
7.31 Design Information Transmittal DIT-B-01399-01, "Unit 1 SGTR Supplemental Analyses Input Assumptions", October 25, 2000.
7.32 Design Information Transmittal DIT-SGRP-00064-00, "Unit 2 Steam Generator Design Moisture Carryover", June 30, 2000.
7.33 DB-12-HVCR, "Design Basis Document for the Control Room Ventilation System", Rev. 7.
7.34 Calculation MD-12-HV-017-N, "Establish outside airflow rates for normal air conditioning system and the pressurization system for the control room", Rev. 2.

Cook Nuclear Plant AST Radiological Analysis Input        RWA-1313-001, Rev. 1 Parameter Development                                   Page 6 of 6 7.35 Specification ES-CIV-0306-QCN, "Containment Isolation System Licensing/Design Bases Requirements", Rev. 1, Change Sheet 1.
7.36 Westinghouse Letter Report AEP-88-331, "Radiation Analysis Manual, D. C. Cook Units 1 and 2'', July 26, 1988.
7.37 D. C. Cook Unit 1 Technical Requirements Manual (TRM), Rev. 64.
7.38 D. C. Cook Unit 2 Technical Requirements Manual (TRM), Rev. 65.
7.39 Westinghouse Letter Report AEP-99-277, "Safety Evaluation SECL 99-076 Revision 2- Containment Modifications Evaluation'', August 27, 1999.
* 7.40 Design Information Transmittal DIT-B-03557-00, "Core Source Term Input for Dose Reanalysis Effort", October 14, 2013.
7.41  Design Information Transmittal DIT-B-03559-00, "RCS Source Term Input for Dose Reanalysis Effort", October 15, 2013.
7.42 Design Information Transmittal DIT-B-03594-00, "Miscellaneous Input for Dose Reanalysis Effort (Contract #01559762)", May 9, 2014.
7.43 USNRC Standard Review Plan NUREG-0800, Section 6.5.2 "Containment Spray as a Fission Product
       *Cleanup System", Rev. 4.
7.44 Design Information Transmittal DIT-B-03594-01, "Miscellaneous Input for Dose Reanalysis Effort (Contract #01559762)- Corrected Lower Coi;npartment Gas Phase Mass Transfer Coefficient", June 4, 2014.
7.45 Drawing OP-1-5144-51, "Flow Diagram, Containment Spray, Unit No. l".
7.46 Vendor Technical Document VTD-BAWI-0015, "Babcock & Wilcox Canada, Operating and Maintenance Manual for Unit 1 Replacement Steam Generators, PUB. #222-7803-0&M-1".
7.47 Procedure l-OHP-4023-E-O, "Reactor Trip or Safety Injection", Rev. 38.
7.48 Engineering Evaluation EE-2005-0139, "Steam Generator Safety Valves'', Revision 0.
7.49 Engineering Change EC-0000051727, "Unit 1 Cycle 25 Core Reload", Rev. O.*
7.50 Engineering Change EC-0000052225, "Unit 2 Cycle 21 Core Reload'', Rev. 0.
7.51 D. C. Cook Unit 1 Technical Specifications Bases, Page Revision Date 06/01/05.
7.52 D. C. Cook Unit 2 Technical Specifications Bases, Page Revision Date 06/01/05.
7.53 Design Information Transmittal DIT-B-03594-02, "Miscellaneous Input for Dose Reanalysis Effort (Contract #01559762)- Revised CREV and HFAEV Filter Efficiency Inputs", July 11, 2014.
7.54 Design Information Transmittal DIT-B-03680-00, "Steam Generator Tube Recovery Time for Alternative Source Term Dose Effort (Contract #01576001)'', February 4, 2016.
7.55 Calculation TH-00-06, "D. C. Cook Unit 1 Steam Generator Tube Rupture with Operator Actions'', Rev 0.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                              Page A1 of A12 Attachment A - Radiological Analysis Input Parameters Item Parameter                                 Value                  Units                       Reference                                                Comments No.
Group A- Core Source Term Inputs A.I   Licensed Core Power                                Unit 1-3304                Mwt                Ref. 7.2, Section 2.C(l)              A single source term which conservatively hounds both units will be developed.
Unit2-3468                                    Ref. 7.3, Section 2.C(l)              Therefore, the higher Unit 2 core power should be applied.
A.2   Thermal Power Measurement Uncertainty              0.34                       %                  Ref. 7.7 Section 3.1.1 (Unit I)       Power level uncertainty is added to the licensed core power as required by Ref. 7.8 & 7.9 (Unit 2)               Section 3.1 of Reference 7.1. This value accounts for uncertainties due to power level instrumentation and is consistent with the value used to meet the requirements of IOCFR50 Appendix K per Footnote 8 of the same reference.
A.3   No. of Assemblies in Core                          193                                           Ref. 7.4 & 7.5, Section 4.2.l A.4   Maximum uranium mass per assembly                  498                        kg                 Ref. 7.40, Item #I                    Maximum value based upon Unit I 15x15 Upgrade fuel assembly with 100%
theoretical density plus 5% uncertainty.
A.5   Expected range of initial fuel enrichments         0.74 -5.00                 w/o U-235          Ref. 7.40, Item #2 A.6    Core Average Bumup                                43,000                     MWdlMTU            Ref. 7.40, Item #3                    This value is based upon a maximum projected bumup for the Unit 2, Cycle 20
                                         -                                                                                                   core design plus 10% for conservatism.
A.7   Fuel rod peaking factor limit                      1.65                                          Ref. 7.40, Item #4                    This peaking factor is based upon the historical Unit 2 Nuclear Enthalpy Rise Hot Channel Factor (FN*H) plus 4% measurement uncertainty.
A.8    Assembly radial peaking factor limit              1.65"                                         Ref. 7.40, Item #5                    Conservatively set to the rod peaking factor limit.
A.9    Number of rods in the core which exceed 6.3 kw/ft 150 rods in up to two fuel                    Assumed                               This parameter serves as a limit on the number offuel rods which exceed the above 54 GWD/MTU                                  assemblies                                                                          bumup limits of Footnote* I I of Reference 7.1 and the number of fuel assemblies which can contain these rods. These values are considered to provide margin for future core designs.
Group B- RCS Source Term Inputs B.1    Percentage of fuel rods with cladding defects     I                          %                  Ref. 7.36, Table 4-1 B.2    RCS Volume                                                                   ft.                Ref. 7.30                             From Note I of Reference 7.30, these values represent 'cold' conditions. The
       -Total RCS Volume                                 Unit I - 12,535.4 max                                                               RCS volumes include the volume of the pressurizer. From Note IO, minimum 12,204.6 min                                                               RCS volumes reflect 10% stean1 generator tube plugging: Note 4 of Reference 7.30 indicates that 'hot' volumes can be obtained by applying a hot volume Unit 2- 12,472.8 max                                                                expansion factor of 3% to the cold values.
12,144.3 min                                                               A minimum RCS inventory should be used to represent the RCS compartment in the dose analysis events which involve fuel failures to maximize the radionuclide
       - Pressurizer Volume                              Unit I - 1834.4                                                                     concentrations. A maximum RCS volume should be applied in the calculation of Unit 2- 1800.0                                                                      the iodine appearance rates for the MSLB and SGTR events.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                               Page A2 of A12 Item Parameter                        Value                    Units                      Reference                                                 Comments No.
B.3  RCS Average Temperature                                               "F                                                        RCS temperature and pressure are used to obtain a fluid density for volume-mass
     - Full Power Operation                     Unit 1 - 571.0                                 Ref. 7.41, Item #1                    units conversion.
Unit 2 - 574.0
     -Zero Load                                 547.0                                          Ref. 7.17 & 7.18, Section Ill.A B.4  RCS Normal Operating Pressure              2250                        psi a              Ref. 7.41, Item #2                    RCS temperature and pressure are used to obtain the fluid density for volume-mass units conversion.
B.5  Nominal Mixed Bed Demineralizer Flow Rate  75                          gpm                Ref. 7.41, Item #6                    This value is equal to the nominal letdown flow rate B.6  Nominal Cation Bed Demineralizer Flow Rate 800                         gal/day            Ref. 7.41, Item #7                    Approximate value since the cation bed usage varies with time of core life and other RCS chemistry considerations. A smaller value conservatively removes less Rb-86, Cs-134, and Cs-137 from the reactor coolant.
B.7  Effective Boron Removal Makeup Flow Rate   400                         gal/day            Ref. 7.41, Item #8                    Nominal makeup flow will vary from zero gpd at the beginning of cycle to several thousand gpd 'near the end of cycle when boron concentrations are low.
A smaller value is conservative since this parameter directly dilutes the radionuclide concentrations in the RCS.
B.8  Letdown Flow Rate                                                      gpm                Ref. 7.41, Item #3                    The normal letdown flow rate applies to tl1e derivation of the RCS source term.
     -Normal                                    75                                                                                   The maximum letdown flow rate should be applied in the development of tl1e
     -Maximum                                   120                                                                                  iodine appearance rates and releases from a ruptured VCT.
B.9  Letdown Fluid Temperature/Pressure         120                         "F                 Ref. 7.41, Item #4 and #5             These values are used for units conversion and should correspond to fluid 365                         psi a                                                    conditions near the location of flow measurement instrumentation.
B.10 Fission Product Escape Rate Coefficients                               sec*               Ref. 7.36, Table 4-1                  Values correspond to full power operation
     - Kr, Xe Isotopes                          6.5E-08
     - I, Br, Rb, Cs Isotopes                   l.3E-08
     - Mo, Tc, Ag Isotopes                      2.0E-09
     - Te Isotopes                              l.OE-09
     - Sr, Ba Isotopes                          l.OE-11
     - Y, Zr, Nb, Ru, Rb, La, Ce, Pr Isotopes   l.6E-12 B.11 VCT Noble Gas Stripping Fractions                                      fraction           Ref. 7.36, Table 4-3                  Table 4-3 of Reference 7.30 states that these stripping fractions are based upon
     -Kr-85m                                    0.61                                                                                 no purge flow from the Volume Control Tank (VCT).
     -Kr-85                                     7.3E-05
     -Kr-87                                     0.84
     -Kr-88                                     0.71
     -Xe-131m                                   0.017
     -Xe-133m                                   0.085
     - Xe-133                                   0.037
     -Xe-135m                                   0.95
     -Xe-135                                    0.35
     -Xe-137                                    0.98
     -Xe-138                                    0.95

                                                                                                                                                                                              *RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input. Parameter Development                                                               Page A3 of A12 Item Parameter                          Value                  Units                     Reference                                                   Comments No.
B.12  Demineralizer Decontamination Factors                                 unitless           Ref. 7.36, Table 4-3 Mixed Bed:
      - Noble Gases (Kr, Xe Isotopes)              1
      - I, Br Isotopes                             10
      - Sr, Ba Isotopes                            10
      - All Other Isotopes                         1 Cation Bed:
      - Noble Gases (Kr, Xe Isotopes)              1
      - Sr, Ba Isotopes                            1
      -Rb-86, Cs-134, Cs-137                       10
      - Rb-88, Rb-89, Cs-136, Cs-138               1
      - All Otl1er Isotopes                        1 B.13  RCS Leakage                                                           gpm                Ref. 7.4 & 7.5, Section 3.4.13         The unidentified leakage rate limit of0.8 gpm from References 7.4 and 7.5
      - Identified                                 10.0 .                                                                             applies to leakage from the pressurizer surge line. The higher value 1.0 gpm for
      - Unidentified                               1.0                                                                                leakage from other sources given in the Action Statement of the same specification is more conservative. Primary coolant leakage is used in the calculation of iodine appearance rates.
B.14  RCS Specific Activity Limits                                          &#xb5;Ci/gm             Ref. 7.4 & 7.5, Section 3.4.16         E-bar is defined in Section 1 of References 7.4 and 7.5 as the sum oftl1e nuclide
      - Nominal - Dose Equivalent 1-131            1.0DE1-131                                                                         weighted average decay energy for non-iodine isotopes with half lives greater
      - Normal - Gross Specific                    100/E-bar                                                                          than 15 minutes. The RCS source term should meet both tlie 100/E-bar and specific iodine activity limits. Noble gas activities from the RCS source term
      - Maximum FuU Power Operation - Dose         60.0                                        Ref. 7.4 & 7.5, Figure 3.4.16-1        can.be used to derive a corresponding Dose Equivalent Xe-133 value to supp01t Equivalent 1-131                                                                                                              a future Teclmical Specification RCS activity limit.
Group C- Containment Inputs C.l   Containment Volume                                                    ft                 Ref. 7.21, Tables 5.2.5.l & 5.4.4.1    In general, Reference 7.21 produces best-estimate compartment volumes that are
      - Upper Containment (Sprayed)                609,773                                     Ref. 7.22                              biased low due to the net free volume inputs from Reference 7.27. SmaUer
      - Lower Containment (Sprayed)                101,735                                                                            volumes will tend to produce higher radionuclide concentrations and higher
      - Lower Containment Fan Room (Sprayed)       47,954                                                                             activity release rates from containment. In addition, since Section 3.3 of
      - Upper Containment (Unsprayed)              120,196                                                                            Appendix A to Reference 7.1 directs calculating the natural convection mixing
      - Ice Condenser (Unsprayed)                  103,507                                                                            rate between sprayed and unsprayed regions based upon the size of the
      - Lower Containment (Unsprayed)              64,890                                                                             unsprayed volumes, smaUer compartment volumes conservatively minimizes the
      - Lower Containment Dead-Ended (Unsprayed)   18,297                                                                             containment internal mixing. However, larger volumes reduce spray effectiveness and result in lower iodine removal coefficients. Note that values from Tables 5.2.5.1and5.4.4.1 of Reference 7.21 do not include the additional reductions due to rounding that are perforn1ed in Section 6 of tl1is same reference as discussed in Reference 7.22.

RWA-1313-001,_Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                Page A4 of A12 Item Parameter                             Value                  Units                      Reference                                                 Comments No.
C.2  Containment Sump Volume                        50,955                    ft                Ref. 7.42, Item #1                     This value represents the minimum sump volume at the time of switchover to recirculation, which conservatively maximizes the radionuclide concentration for the ESF leakage outside of containment.
C.3  Containment Wall Surface Deposition Area       0                         fr                Assumed                                From the guidance of Section III.4.C.i of Reference [7.43], setting the wall deposition area lo zero will cause the elemental iodine removal by deposition onto sprayed surfaces to be conservatively ignored.
C.4  CEQ Flow Rates                                                           cfin              Ref. 7.10 & 7.11                       Values represent design flow rates for single train fan operation (HV-CEQ-1 or From Upper Containment                         39,000                                                                             HV-CEQ-2). Lower flow rates will conservatively minimize containment From Top of Containment Dome                   1,000                                                                              mixing which reduces iodine removal by containment sprays.
From Steam Generator Compartments              1,000                                                                              Section 4.19 of Reference 7.27 identifies that the Instrumentation Room is From PZR Compartment                           500 included in the Dead-End volume. Similarly, Sections 4.10 and 4.11 of From Fan Rooms                                 200                                                                                Reference 7.27 identify Volume IX and X as the Stearn Generator and From Instrumentation Room                      100                                                                                Pressurizer cubicles, respectively. Table 5.3.3.1 of Reference 7.21 shows that compartments IX and X are included in the development of the Lower Containment Net Free Volmne, which is used to calculate the size of the Lower Containment Unsprayed Volume in Equation 5.4.3 of the same reference.
C.5  CEQ Fan Start Time                             300                       seconds           Ref. 7.42, Item #3                     The response time includes actuation signal processing, EDG startup, and fan start.
C.6  Containment Leak Rate                          0.18                      Weight %/day      Ref. 7.42, Item #2                     This value represents a reduction from the current Tech. Spec. leak rate limit of 0.25%/day. Use of a value less than 0.25% in the dose analysis requires a change to Section 5.5.14.c of the Technical Specifications (Ref. 7.4 & 7.5) per Section 3.7 of Appendix A to Ref. 7.1.
C.7  Containment Purge Isolation Valve Stroke Time  5                         seconds           Ref. 7.35, pages C5 & D5               The Containment Purge Supply and Exhaust System isolation initiates on an Ref. 7.10 & 7.11                       automatic Safety Injection Signal.
Ref. 7.4 & 7.5, Table 3.3.6-1, Item #4 Ref. 7.51 & 7.52, Section B.3.3.6 C.8  Containment Purge Exhaust Flow Rate            33,000                    cfm               Ref. 7.10 & 7.11                       This value represents the design flow rate for two fan Containment Purge Exhaust fans (HV-CPX-1 & HV-CPX-2) at 16,000 cfin per fan plus the Instrmnent Room Exhaust fan (HV-CPX-1) at 1000 cfm. Maximum purge flow is conserv'ative.
C.9  Containment Spray Start Time                   300                       seconds           Ref. 7.42, Item #4                      Value includes EDG start, sequencer delays, pump acceleration, and pipe fill times.
C.10 Containment Spray Stop Time                                                                Ref. 7.42, Item #5                     Iodine removal by contaimnent sprays is not applicable when contaimnent spray
     - Duration sprays are secured for pump suction 7                         minutes                                                  is secured. Natural deposition of aerosols is not applicable when sprays are realignment for recirculation                                                                                                   active.
     - Time in event after which sprays are secured 24 hours                  hours C.11 Mass Mean Diameter of Spray Drops              - Lower Compartment - 671 &#xb5;m                Ref. 7.42, Item #6                     This parameter is used in the calculation of the elemental iodine spray removal
                                                    - Upper Compartment - 609                                                           coefficient.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                     PageA5ofA12 Item Parameter                       Value                 Units                       Reference                                                      Comments No.
C.12  Containment Spray Flow Rate                                     gpm                 Ref. 7.42, Item #7                         This parameter is used in the calculation of the elemental and aerosol iodine
      - Upper Containment                     1,466                                                                                  spray removal coefficients. Minimum spray flow rates conservatively minimize
      - Lower Containment                     660                                                                                    the removal coefficients.
      -Fan Rooms                              201 C.13  Spray Drop Fall Height                                           ft                 Ref. 7.21, Table 6.3                       The spray drop fall height is used in the calculation of the aerosol iodine removal
      - Upper Containment                     58.6                                        Ref. 7.22                                  coefficient.
      - Lower Containment                     28.5
      -Fan Rooms                              20.1 C.14  Spray Drop Fall Time                                             seconds            Ref. 7.42, Item #8                         The spray drop fall time is used in the calculation of the elemental iodine
      - Upper Containment                     11.9380                                                                                removal coefficient. Note that these values correspond to a drop diameter of 675
      - Lower Containment                     2.77329                                                                                &#xb5;m for the lower compartment and fan rooms, and a drop diameter of 625 &#xb5;m for
      -Fan Rooms                              1.95590                                                                                the upper compartment.
C.15  Gas Phase Mass Transfer Coefficient, K8                          m/sec              Ref. 7.42, Item #9                         This parameter is used in the calculation of the elemental iodine spray removal
      - Upper Compartment                     0.113640                                    Ref. 7.44                                  coefficient. These values correspond to drop diameters of 675 &#xb5;m for the lower
      - Lower Compartment                     0.155038                                                                               compartments and 625 for the upper compartment. Note that the coefficient for the lower compartment applies to the botl1 the lower containment and fan room volumes.
C.16  Time ofECCS Switchover to Recirculation 1,388.4                  seconds            Ref. 7.39, Section 3.3.1.2.l & Table 3.3-1 This value corresponds to the start of the first the RHR/CTS pump after suction transfer to the containment sump, and represents the maximum RWST drain-down case with two ECCS trains in operation and a 3-minute interruption in RHR/CTS flow. Minimum switchover time increases the amount of ESF leakage outside of contaill1)1ent.
C.17   Sump pH at Time of Spray Recirculation ?:7.0                                       Ref. 7.29, Page 71, Item 2.                The Spray Additive system provides a containment sump pH greater than or equal to 7.0 after the containment spray is transferred to recirculation. This is required by Reference 7.43, Section Il-SRP Acceptance Criteria l.Gto ensure long term iodine retention in the sump solution.
Group D - Control Room Inputs D.1    Control Room Volume                    50,616                   ft'                Ref. 7.23, p. 16                           Reference 7.23 refers to this value for the control room volume as the minimum verifiable free volume. A smaller control room volume has competing effects of higher radionuclide concentrations and increased removal by the control room filters. The sensitivity of the dose consequences to small changes in volume is negligible.
D.2   Control Room Ventilation System (CRVS)  880                      cfm                Ref. 7.13 & 7.14                           The design makeup rate is shown in References 7.13 & 7.14 as 800 cfm, with a Normal Makeup Flow Rate                                                             Ref. 7.34, Section 2.7                     normal expected operating range from Reference 7.34 of 720 - 880 cfm. Use of a maximum outside makeup flow rate is conservative. This flow enters the control room envelope at the suction of the air handler units HV-ACRA-1 and HV-ACRA-2.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                   Page A6 of A12 Item Parameter                          Value                  Units                       Reference                                                    Comments No.
D.3   Control Room Emergency Ventilation (CREV) SI input from either unit                      Ref. 7.4 & 7.5, Table 3.3.7-1 Actuation Signals D.4   CREV System Response Time                 60                         seconds             Ref. 7.33, Sections 3.3.2, 3.3.4 & 3.7.2 The response time includes actuation signal processing, EDG startup, fan start, Ref. 7.4 & 7.5, SR3.8.l.8                damper realignment. Following a LOOP, the EDGs start and are ready to load in Ref. 7.6, Section 8.4                    10 seconds. CRVS components are powered by 600 V electrical distribution J                system which is sequenced onto the safety bus in the first load block, nominally at 10 seconds. The 60 second CREV response time includes margin to address additional delay time components.
D.5   CREV Pressurization/Cleanup Flow Rates                               cfm                                                          The design filtered makeup flow is 800 cfm with an uncertainty of +/-10%.
      - Filtered Makeup                         720-880                                        Ref. 7.13 & 7.14                         Maximum makeup flow is conservative for this analysis. This flow enters the Ref. 7.34, Section 2. 7                  control room envelope at the suction of the pressurization/cleanup filter system Ref. 7.47, Attachment A, Step 10         fan (HV-ACRF-1 or HV-ACRF-2).
The total flow rate of 5400 - 6600 cfm represents single fan operation. A
      - Total System Flow                       5400-6600                                      Ref. 7.4 & 7.5, Section 5.5.9 minimum recirculation flow rate reduces the radionuclide removal by the control room filters in the event dose analyses. A maximum flow rate maximizes the activity on the control room filters in the shine dose analysis. Reference 7.47 states that following system actuation, one of the two fans is secured. This is done to keep air velocities through the charcoal adsorbers low enough to ensure a minimum iodine residence time.
D.6   Control Room Unfiltered Inleakage         40                          cfm                Ref. 7.42, Item #10 D.7   CREV Filter Efficiency                                                %                  Ref. 7.53, Item #11
      -Elemental                                95
      -Organic                                  95
      - Particulate                             99 D.8   Maximum CREV Filter Bypass Fraction       1                           %                  Ref. 7.4 & 7.5, Section 5.5.9            The filter bypass fraction is based upon the HEPA removal efficiency requirement from References 7.4 & 7.5 and is considered separately from the filter efficiencies.
Group E- Steam Generator Inputs E.l   Secondary Specific Activity               0.10                        &#xb5;Ci/gm DE I-131    Ref. 7.4 & 7.5, Section 3.7.17 E.2   Primary-to-Secondary Leak Rate                                                           Ref. 7.42, Item #12                      Section 5.1 of Appendix E to Reference 7.1 requires that the primary-to-Accident Leakage                          1.0                         gpm to all SGs                                              secondary leakage be apportioned between steam generators in such a manner 0.25                        gpm to any I SG                                             that maximizes the calculated dose. Use of a single stean1 generator leakage rate in the dose analysis requires a change to Section 5.5.7.b.2 of the Technical Specifications (Ref. 7.4 & 7.5).

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                     PageA7ofA12 Item
                          *Parameter                              Value                  Units                       Reference                                                       Comments No.
E.3    Temperature of the fluid used to assess RCS     70                         "F                  Ref. 7.19, Section 2                       Reference 7.1, Appendix E, Section 5.2 requires that the density used in leakage by the leak rate monitoring program                                                                                               converting volumetric leak rates to mass leak rates be consistent with the surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications.
E.4    Steam Generator Secondary Side Liquid Mass                                                                                                The minimum steam generator liquid mass represents Hot Full Power (HFP)
       -Minimum                                        97,515.7                    lbm/SG             Ref. 7.25, p. 16                           conditions, while the maximum value is conservatively based upon Hot Zero
       -Maximum                                        161,000                     lbm/SG             Ref. 7.25, p. 32                           Power (HZP) conditions. The minimum mass is used in combination with primary-secondary leakage rates to maximize the radionuclide concentrations in the steam generators. The maximum mass is used in the release of iodine initially present in the SG secondary.
E.5    Time to cool the RCS to 212 &deg;F and terminate    24                          hr                 Ref. 7.42, Iteni #13                       Reference 7.1, Appendix E, Section 5.3 requires that primary-to-secondary steam releases                                                                                                                            leakage be assumed to continue until the RCS is cooled to 212 "F, and radioactive releases should continue until shutdown cooling is in operation and releases from the steam generators have been terminated. The 24-hour cooldown time is based upon a single train of RHR in service.
E.6    SG Moisture Carryover Fraction                  Unit 1 - 0,045              %                  Ref. 7.46, Section 8.2.1(Unit1)            This value is used to limit particulate retention in the steam generators per Unit 2-0.15                                    Ref. 7.32 (Unit 2)                         Appendix. E, Section 5.5.4 of Reference 7.1.
E.7    SGTR break flow/tube leakage flashing fractions                                                Ref. 7.26, - Pages 42 - 44                 The fraction of the primary fluid which flashes to vapor on the secondary side of the steam generators is assumed to be released to the environment with no partitioning in the steam generators. The Unit 2 flashed break flow from Reference 7.26 exceeds the Unit 1 value from Reference 7.55. Consequently, Unit 2 inputs are applied to produce bounding flashing fractions.
E.8    Duration of intact SG tube bundle uncovered     40                          minutes            Ref. 7.54                                  Following a reactor trip, redistribution offluid in tlie SG secondary causes the following a reactor trip?                                                                                                                 water level inside the wrapper to drop below the top of the tubes until level is recovered by AFW. During periods of tube uncovery, primary-secondary leakage is assumed to be released to the environment with no partitioning in the steam generators.
Gronp F - RWST Inputs F.l    RWST Total Volume                               420,000                     gallons            Ref. 7.45 F.2    Minimum operable RWST Liquid Volume             375,500                     gallons            Ref. 7.4 & Ref. 7.5, Section 3.5.4 F.3    Maximum delivered RWST volume at time of        321,862.5                   gallons            Ref. 7.39, Sect. 3.3.1.2.1 and Table 3.3-1 This is the maximum volume delivered from the RWST from the beginning of switchover to recirculation                                                                                                              the event until after the switchover from injection to recirculation is complete, and is used to detenuine tl1e minimum remaining RWST inventory at the time of switchover.
F.4    Maximum RWST temperature                        100                         "F                 Ref. 7.4 & Ref. 7.5, Section 3.5.4 F.5    Minimum RWST pH                                 4.479                                          Ref. 7.29, Appendix 4.1                     This value corresponds to the maximum RWST boron concentration of 2600 ppm from Section 3.5.4 ofRef. 7.4 & Ref. 7.5.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                   Page AS of A12 Item Parameter                             Value                    Units                       Reference                                                    Comments No.
Group G.1-Event Specific Inputs - LOCA G.1.1 Engineered Safeguards Features (ESF) system      0.1                         gpm                Ref. 7.20, Data Sheet I                  This value applies to all sources ofESF leakage into the Auxiliary Building, both leakage limit into the Auxiliary Building                                                                                                inside and outside the ESF Ventilation system envelope .. Section 5.2 of Appendix A to Reference 7.1 requires that the value of this parameter be doubled in the evaluation of the dose consequences from ESF leakage.
G.1.2 Engineered Safeguards Features (ESF) system      0.5                         gpm                Assumed                                  This values serves as a limit on the total ESF seat leakage past isolation valves leakage limit to the RWST                                                                                                               on lines which recirculate sump fluid back to the RWST. This parameter is required to be doubled in the dose analysis per Section 5.2 of Appendix A to Reference 7 .1.
G.1.3  Containment Sump Temperature Profile            Figure Al                                      Ref. 7.6, Unit 1, Figure 14.3.4-9        Maximum sump temperatures conservatively maximize the flashing fraction of ESF fluid leakage outside of containment. In addition, higher sump fluid temperatures reduce the iodine partition coefficient of sump fluid in the RWST.
The profile from Reference 7.6 is considered to be representative ofpost-LOCA sump conditions. A bounding temperature envelope based upon this profile will provide for future analytical margin.
G.1.4  Status of Aux. Bldg HVAC following SI/LOOP      ESF Ventilatio_n - running                     Ref. 7.517.32, Section B.3.7.12 (Unit 1) Reference 7 .12 shows that the exhaust from both the normal Auxiliary Building Ref. 7.52, Section B.3.7.12 (Unit 2)     ventilation system and the ESF Ventilation system discharge through the unit vent.
G.1.5  Time of Safety Injection Signal                 5.7                         seconds            Ref. 7.6, Unit 2, Table 14.3.1-8         The SI actuation time is used to isolate contaimnent and to initiate control room ventilation realignment. This value is considered to be representative.
G.1.6  ESF Ventilation System filter efficiencies      nla                         %                  n/a                                      Omission ofESF Ventilation system filter efficiencies ensures that filtration by
       -Elemental                                                                                                                              this system is conservatively ignored in the dose analysis
       -Organic
       - Particulates Grnup G.2 - Event Specific Inputs - Main Steam Line Break G.2.1  Fuel Failure Fraction                           0                                              Ref. 7.49, Section 1.3.3 (Unit I)        Sufficient DNB margin exists to prevent fuel failures for this event.
Ref. 7.50, Section 1.4.2 (Unit 2)
G.2.2  Time of Safety Injection Signal                 <I                          seconds            Ref. 7.6, Section 7.2.3 & Unit 2 Table   The safety injection actuation oc".urs on low steam line pressure. The SI time is 14.2.5-2                                 used to initiate control room ventilation realignment. This value is considered to be represeniative, and the dose analysis results are not sensitive to small variances in this parameter.
G.2.3  Intact SG Steam Release                                                     lbm                Ref. 7.25, p. 7                          From Sections 1.1and6.2 of Reference 7.25, these values are developed 2hr                                          456,000                                                                                  assuming that the RCS is cooled to 212 'Fat the end of 8 hours with no credit 8hr                                          1,186,000                                                                               taken for heat removal by the RHR system.
24hr                                         1,347,000

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                    Page A9 of A12 Item Parameter                               Value                   Units                       Reference                                                      Comments No.
Group G.3 - Event Specific Inputs -Steam Generator Tube Rupture G.3.1  Fuel Failure Fraction                             0                                             Ref. 7.49, Section 1.3.3 (Unit I)         Sufficient DNB margin exists to prevent fuel failures for this event.
Ref. 7.50, Section 1.4.2 (Unit 2)
G.3.2  Ruptured Tube Break Flow                          146,704                    lbm                Ref. 7.24, p. 46                          The break flow from the ruptured steam generator is based upon a release duration of30 minutes per Sections 5.0 and 6.1 of Reference 7.24. Case 27 was chosen as it represents the most limiting value. Note that the additional l 0%
analytical flow margin used throughout Ref. 7.24 is not applied to this value.
G.3.3  Ruptured SG Stearn Release                        66,171                     lbm                Ref. 7.24, p. 46                          The steam flow from the ruptured steam generator is based upon a release duration of 30 minutes per Sections 5.0 and 6.1 of Reference 7.24. Case 2A was chosen as it represents the most limiting value.
G.3.4  Intact SG Stearn Release                                                     lbm                                                          From Sections 3.1 and 6.3.3 of Reference 7.24, these values are developed 30 min                                        0.75 x 264,686 = 198,515                      Ref. 7.24, p. 58                          assuming that the RCS is cooled to 212 &deg;F at the end of 8 hours and no credit for
       -30 min -2 hr                                     314,432                                       Ref. 7.24, p. 49                          heat removal by the RHR system 8 hr                                           1,367,475                                     Ref. 7.24, p. 54 24hr                                           1,347,000                                     Ref. 7.24, p. 55 G.3.5  Condenser Iodine Partition Coefficient            100                                           Ref. 7.24, Section 6.3.4 G.3.6  Full Power Main Stearn Flow Rate                  17,153,800                 lbm/hr             Ref. 7.48, Page 8                         This value represents 105% of the maximum secondary steam flow rate at 100%
rated thennal power.
G.3.7  Time of Reactor Trip                              IOI                        seconds            Ref. 7.26, Section 6.3.2                  An early trip time is conservative since the additional iodine partitioning by the condenser prior to reactor trip more than offsets the higher pre-trip break flow flashing fraction.
G.3.8  Time of Safety Injection                          334.74                     seconds            Ref. 7.24, Appendix A, Cases 25 & 25a     This is the maximum value of SI initiation time from all SGTRH cases presented in Appendix A of Reference 7.24. The SI actuation time is used to initiate control room ventilation realignment.
Group G.4 - Event Specific Inputs - Control Rod Ejection G.4.1  Fraction of Fuel Rods with Cladding Damage        10                         %                  Ref. 7.42, Item #16 G.4.2  Fraction of Core that Experiences Melting         0.25                       %                  Ref. 7.42, Item #17 G.4.3  Time of Safety Injection Signal                   60                         seconds            Ref. 7.6, Unit I Section 14.2.6.17 & Unit Th\s is an approximate value based upon the discussion in Reference 7.6. The SI 2, Section 14.2.6.2.9                     actuation time is used to initiate control room ventilation realignment, and the dose analysis results are not sensitive to small changes in this parameter.
Group G.5 - Event Specific Inputs - Locked Rotor G.5.1  Fuel Failure Fraction                             II                         %                  Ref. 7.42, Item #18 G.5.2  Steam release during plant c.ooldown                                         lbm                Ref. 7.25, p. 6                           From Sections I. I and 6.2 of Reference 7.25, these values are developed 2hr                                            460,000                                                                                 assuming that the RCS is cooled to 212 &deg;Fat the end of 8 hours with no credit 8hr                                            1,256,000                                                                               taken for heat removal by the RHR system.
24hr                                           1,347,000

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Paramete~ Development                                                               Page A10 of A12 Item Parameter                                 Value                  Units                       Reference                                                   Comments No.
G.5.3 Time CR ventilation is manually aligned             20                        minutes             Ref. 7.42, Item# 19 Group G.6- Event Specific Inputs - Fuel Handling Accident G.6.1 Fraction of rods in dropped assembly that fail      100                        %                  Ref. 7.6, Unit !,Section 14.2.1.4       Reference 7.6 supports the conservative assumption that 100% of the fuel rods in the dropped assembly fail.
G.6.2 Depth of water above the fuel assemblies             ?.-23                     ft                 Ref. 7.4 & 7.5, Section 3.7.14 & 3.9.6 G.6.3 Time CR ventilation is manually aligned             20                         minutes            Ref. 7.42, Item #20                     This is a conservatively long length of time considering that direct communication between the control room and the refueling stations is required by Section 8.9.1 of References 7.37and7.38.
G.6.4  Fuel movement time after reactor shutdown          120                        hr                 Ref. 7.37 & 7.38, Section 8.9.2         Minimum wait period after reactor shutdown before irradiated fuel can be moved G.6.5  Status of FHAEV system during refueling            One train operable and in                     Ref. 7.4 & 7.5, Section 3.7.13 operations                                         Operation G.6.6  FHAEV System filter efficiencies                                              %                  Ref. 7.53, Item #21
       -Elemental                                         90
       -Organic                                           90
       - Particulates                                     99 G.6.7  Maximum FHAEV Filter Bypass Fraction               1                          %                  Ref. 7.4 & 7.5, Section 5.5.9.a         The filter bypass fraction is based upon the HEPA removal efficiency requirement from References 7.4 & 7.5 and is considered separately from the filter efficiencies.
G.6.8  Status of building openings during irradiated fuel                                                                                       Sections 4.3 and 4.4 of References 7.15 and 7.16 establish integrity of the movement                                                                                                                                 containment and Auxiliary Building during movement of irradiated fuel. Both containment airlock doors* are permitted to be open simultaneously provided that
       - Containment                                      Op.en                                         Ref. 7.4 & 7.5, Section 3.9.3 & Bases   they can be closed subsequent to a containment evacuation. The Auxiliary Building doors must be closed to ensure operability of the FHAEV system.
       - Auxiliary building                               Closed                                        Ref. 7.4 & 7.5, Section 3. 7.13 & Bases Administrative controls which permit breaching these doors requires a dedicated individual in continuous communication with the control room who has the means to rapidly close the opening.
Group G.7 -Event Specific Inputs- Waste Gas Decay Tank Rupture G.7.1  WGDT Inventory Limit                               43,800                     Curies DE Xe-133   Ref. 7.37 & 7.38, Section 8.7.13 Group G.8 - Event Specific Inputs - VCT Failure G.8.1  VCT Liquid Volume                                  267                        ff                 Ref. 7.36, Table 4-3 G.8.2  VCT Vapor Volume                                   133                        ff                 Ref. 7.36, Table 4-3 G.8.3  Time to isolate the ruptured tank                  15                         minutes            Ref. 7.6; Section 14.2.3.l

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                       Page A11 of A12 Group H-Atmospheric Dispersion Factor Inputs H.l  Receptor Locations Unit 1 Nonna! Control Room Intake                 Onsite atmospheric dispersion factors (X/Q) are calculated from all release     Plant parameters used in the development of the atmospheric Unit 2 Nonna! Control Room Intake                 points to intakes for both units.                                               dispersion factors include release height, receptor height, arid Unit 1 Emergency Control Room Intake                                                                                              coordinates which allow the detennination of the distances and Unit 2 Emergency Control Room Intake                                                                                              directions between the release-receptor pairs.
H.2  Release Locations  Unit 1 Containment Surface (Diffuse)              A diffuse X/Q from the containment surface is generated for use in the Unit 2 Containment Surface (Diffuse)              analysis of the LOCA containment leakage onsite dose.
Unit I Containment Surface (Offsite)              Limiting X/Qs from the containment surface to the EAB and LPZ locations Unit 2 Containment Surface (Offsite)              are needed for the containment leakage offsite dose.
Unit I Containment c;Iosest Point (Onsite)        The location on the containment building surface which is closest to the Unit 2 Containment Closest Point (Onsite)         intakes provides X/Qs that are more limiting that the release from any specific containment penetration. This atmospheric dispersion factor is applicable to a Fuel Handling Accident (FHA) inside containment.
Unit I WestMSSVs//PORVs                           MSSV/PORV releases apply to all events which involve secondary releases.
Unit 2 West MSSVs/PORVs Unit 1 Steam Jet Air Ejectors                     Releases from the SJAEs apply to the Steam Generator Tube Rupture (SGTR)
Unit 2 Steam Jet Air Ejectors                     event prior to reactor trip.
Unit 1 Plant Vent                                 The plant vent release point is credited when safety related Auxiliary Unit 2 Plant Vent                                 Building ventilation is available. This includes the LOCA containment purge, FHA inside the Aux. Building, and LOCA ESF leakage in the ESF Ventilation envelope.
Unit 1 East Turbine Building Closet Point         Releases from the east side of the Turbine Building and from the West Main Unit 2 East Turbine Building Closet Point         Steam Enclosure are considered for the faulted steam generator during the Unit 1 West Main Steam Enclosure                  Main Steamline Break (MSLB) event.
Unit 2 West Main Steam Enclosure Aux. Building - North Intake                      Auxiliary Building nonnal ventilation intakes may serve* as release points for Aux. Building - South Intake                      tank ruptures when the building ventilation is not in service.
Unit 1 RWST                                       RWST releases apply to LOCA ESF seat leakage through isolated lines which Unit2RWST                                         recirculate to the RWST.
Unit 1 East Main Steam Enclosure (Offsite)        East Main Steam Enclosure and West Turbine Building atmospheric Unit 2 East Main Steam Enclosure (Offsite)        dispersion factors are needed for EAB and LPZ locations only since the west Turbine Building - NW Corner (Offsite)            steam enclosure and east turbine building locations are more limiting with Turbine Building- SW Corner (Offsite)             respect to onsite receptors.

Cook Nuclear Plant AST Radiological Analysis Input                    RWA-1313-001, Rev. 1 Parameter Development                                         Page A12 of A12 190                                                              i90 LL...
-110 e
180                                                              180 170 -
C' Q>
i.-
0 180                                                                  \60 ......,
i.-
())                                                                         ~
0..                                                                         cu Cl..
E    150                                                              150 E Q.)
1--                                                                          tl.>
l-140                                                              140 130 0         ,             2 I 1 I 3             ~   5 130 no        10            10            . 10            10  jQ Time (s)
Figure A1: Containment Sump Temperature Profile

  *rf*
Red Wall'
. ~~ab!tw. Cook Nuclear Plant AST Radiological Analysis Input RWA-1313-001, Rev. 1 Parameter Development                        Page 81 of 84
                            , Attachment.B Owner's Review Comments

RWA-1313-001, Rev.1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                           Page 82 of 84 Revision 0 AEP Acceptance/Additional AEP Comments for RWA-1313-001                                                                   RWA Response Comments Comment No.        Document Location                     Description I       Page I                       Update page numbers.                       Page numbering has been updated.                                              Acceptable. JJW-7/11/2014 2       Page 3, Methodology          Could the practice of not listing the      RWA-1313-001 is created specifically to serve as a front-end input            Acceptable. JJW-7/11/2014 parameter as it is used in the calculation document to be: referenced by the dose event calculations. This is clearly be considered a future error likely        stated in both the Purpose and Methodology sections of the document.
situation?                                 Future misuse of this calculation will require the user to ignore the stated intent of the document. However, while the 'Comments' column of the calculation is included to provide perspective on how some parameters are to be used in the dose analysis, some of the text in this column does imply a
                                                                                    'past tense' use of the listed value. Therefore, minor text changes have been                                      -
inade to project a more 'future tense' in the comments section. Note that calculation RWA-1313-015 will be prepared at conclusion of the project which summarizes the actual analysis values.
3       Page 4-6, References         The UFSAR and DB-12-XX documents           Alternate references have been provided where available. The UFSAR and        Acceptable. The input values *
(References 6.6, 6.32, 6.33, and 6.34) are DBDs continue to be cited in the following cases:                             associated with these references are not typically utilized as design input                                                                                   deemed to be appropriate and unless deemed appropriate. Alternate       Item D.2- DB-12-HVCR is used to document that the CREV dampers are            conservatism has been applied when references should be used in their place. powered from the 600 Vac Electrical Distribution System and are designed      necessary. JJW-7111/2014 to close following a LOCA. UFSAR Section 8.4 identifies that the 600 volt buses are powered from the emergency diesel generators and are loaded onto the diesel in the first load block within I 0 seconds.
Item G.1.3 - The containment sump temperature profile in Figure 14.3.4-9 if the Unit I FSAR is identified as being representative. A bounding temperature envelope will be applied in the analysis to create analytical margin.
Item G.1.5 - The timing of the SI signal for the LOCA listed in the FSAR is considered to be representative, and the dose analysis is not sensitive to small changes in this parameter. The value from the FSAR is adjusted in the analysis for future design margin.
Item G.2.2 - The timing of the SI signal for the MSLB listed in the FSAR is considered to be representative, and the dose analysis is not sensitive to small changes in this parameter. The value from the FSAR is adjusted in the analysis for future design margin.

RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                      Page B3 of B4 Item G.4.3 - The exact timing of the SJ signal for the CRE event is not specified in the FSAR. The FSAR discussion states that ECCS is actuated
                                                                  'within one minute' after the break, indicating that this value is representative. The dose analysis is not sensitive to small changes in this parameter, and the 60 second actuation time is conservatively combined with a 60 second delay time in the analysis.
Item G.6.1 - The FSAR supports the conservative assumption that I 00% of the rods in the dropped fuel assembly are assumed to fail.
4  General Input      Any time Reference 6.21 (PRA-DOSE-          Letter Report RWA-L-1313-002 has been added to the reference list and is       Acceptable. JJW-7/11/2014 011) is listed, the justification report    cited along with Reference 6.21 in the calculation.
prepared by RWA should be listed as well.
5  Page Al, Item A. I The reference should be listed as 6.2       The link to Reference 6.2 has been corrected.                                  Acceptable. JJW-7/11/2014 Section 2.C(l) instead of Section 6.1.
6  Page A3, Item B.14 Please state the purpose of deriving a      Discussion ofXE-133 is not relevant to this item and reference to it may be    Acceptable. JJW-7111/2014 corresponding value for Xe-133.             removed. However, to specifically address this comment, the text has been revised to state that the DE-133 value derived from the RCS source term may be used to support a future Tech. Spec. RCS activity limit.
7  Page A4, Item C.6  Specify that a license amendment request    The need for a license change has been added to the comment section.           Acceptable. JJW-7/11/2014 is required for the new value.
8  Page A5, Item C.15 The lower compartment mass transfer          The value of the lower compartment gas phase mass transfer coefficient has    Acceptable. JJW-7/11/2014 coefficient should correspond to a drop      been changed to correspond to a drop diameter of 675 &#xb5;m.
diameter of 675 um as opposed to 700 um.
9  Page A6, Item D.5  A statement is made in the comments          The filter shine dose methodology assumes a filter efficiency of 100% and     Acceptable. JJW-7/J 1/2014 regarding maximum*flow rates                 applies this to the control room makeup flow. As such, the sensitivity of the maximizing the activity on the control       dose to the control room recirculation flow rate is relatively weak. Use of room filters. The input values correspond    the maximum single-fan value of 6600 cfm is appropriate since plant to single fan operation. How sensitive is. operators will secure one of two fans to ensure minimum iodine residence the listed value to two fan operation?       times in the filters.
10 Page A6, Item E.2  Specify that a license amendment request     The need for a license change has been added to the comment section.          Acceptable. JJW-711112014 is reauired for the new value.
* RWA-1313-001, Rev. 1 Cook Nuclear Plant AST Radiological Analysis Input Parameter Development                                                                       Page 84 of 84 II      Page A7, Item F.3              What RWST level fonns the basis for this     From AEP 99-277, this parameter is based upon maximum pump flow rates,       Acceptable. JJW-7/11/2014 parameter? It seems low.                     delay times for operator action to perfonn pump suction realignments, and a switchover time that begins when the volume of water delivered from the RWST has reached 280,000 gallons. Since the analysis in AEP 99-277 is perfonned to demonstrate that the minimum RWST deliverable volume requirement is satisfied, the 280,000 gallon value should be consistent with the minimum initial RWST volume allowed by Tech. Specs. In the dose analysis, the maximum transfer volume of321,862.5 gallons is used to detennine the minimum amount of RWST inventory remaining in the tank at the end of switchover. This value is obtained by subtracting 321,862.5 gallons from the Tech. Spec. minimum initial inventory of375,500 gallons.
12      Page A7, Items G.1.1 and G.1.2                                              The comment section has been revised to identify that Reg. Guide 1.183       Acceptable. JJW-7/11/2014 If these values are doubled in the analysis, requires please sav so in the comments.               that olant leakage limits are to be doubled in the dose analvsis.
13      Figures Al and A3              It may be prudent to darken the part of the  Footnotes were added to Figures Al and A3 to emphasize that the figures      Acceptable. JJW-7/11/2014 plot that corresponds to the RCS pressure    contain traces for both RCS and Steam Generator pressures.
(the SG pressure is more prominent).
14      Figures A2 and A4              Add "Hot Leg" to the title of the figure. Figure titles have been revised                                              Acceptable. JJW-7/11/2014 Revision 1
* AEP Acceptance/Additional AEP Comments for RWA-1313-001, Rev. 1                                                                 RWA Response Comments Comment No.        Document Location
* Descriotion I            Page I, "Description"    Please add a statement similar to the        Text has been added to the description to identify that                      Response accepted. -JJW (02112/16) following ... " ... secondary side of the    clarifying infonnation was added to the* VCT rupture inputs steam generators as well as clarifying infonnation regarding the VCT rupture."
2               Page 6, Section 6      The title of the DIT should be changed       The title of the DIT has been corrected.                                     Response accepted. -JJW (02/12/16) from " ... Generator Recovery Time ... " to
                                           "... Generator Tube Recovery Time ... ".
3               Page A7, Item E.7      Is there a way to add a cross-reference to   TH-00-06 has been added to the reference list and the                        Response accepted. -JJW (02/12/16)
TH-00-06 (Unit I SGTR) to show that we       Item E.7 comment was revised to state that the Unit 2 flashed are using bounding values? Alternatively     flow exceeds Unit I value, and is therefore applied to produce this could be stated in the dose             bounding flashing fractions.
consequence calculations.}}