GO2-13-174, Final Safety Analysis Report, Amendment 62, Appendix J - Shielding Evaluation Report

Final Safety Analysis Report, Amendment 62, Appendix J - Shielding Evaluation Report

ML14010A318
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Issue date:	12/30/2013
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GO2-13-174
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C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 J-i Appendix J SHIELDING EVALUATION REPORT

Burns and Roe, Inc., performed the analysis of radiation levels occu rring inside primary containment, assembled, edited, reviewed, and approved this technical report for Energy Northwest.

EDS Nuclear Incorporated performe d the analysis of radiation le vels occurring in the reactor building secondary containment under subcontract to Burns and Roe, Inc. Later revisions have been issued by Energy Northwest to incorporate plant changes.

Energy Northwest performed the an alysis of radiation levels occurring in areas outside the reactor building secondary containment.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Appendix J SHIELDING EVALUATION REPORT

TABLE OF CONTENTS

Section Page J-iii

SUMMARY

........................................................................................J-xv ABSTRACT........................................................................................J-xvi

J.1 INTRODUCTION..........................................................................J.1-1

J.2 REQUIREMENTS.........................................................................J.2-1 J.2.1 SHIELDING EVALUATION REGULATORY REQUIREMENTS...........J.2-1 J.2.1.1 Accident Analysis Requirements....................................................J.2-2 J.2.1.2 Source Term Assumptions...........................................................J.2-2 J.2.1.3 Vital Ar ea Access Requirements....................................................J.2-3 J.2.1.4 Systems Containing the Sources....................................................J.2-4 J.2.1.5 Safety-Related Equipment (C1E/SRM)............................................J.2-4 J.2.2 SHIELDING EVALUATION TASK DESCRIPTION............................J.2-4 J.2.3 SHIELDING EVAL UATION ITEM DELETED FROM SHIELDING ANALYSIS CO NSIDERATION....................................J.2-5

J.3 ANALYTICAL METHODOLOGY.....................................................J.3-1 J.3.1 ACCIDENT SCENARIO...............................................................J.3-1 J.3.2 CONTAMINA TED SYSTEMS.......................................................J.3-1 J.3.2.1 Systems Included fo r Primary Containment Analysis...........................J.3-2 J.3.2.2 Systems Included for Secondary Containment Analysis........................J.3-2 J.3.2.3 Systems Excluded......................................................................J.3-3 J.3.3 SOURCE TERM ASSUMPTIONS...................................................J.3-3 J.3.4 TIME PERIOD CONS IDERED FOR STUDY.....................................

J.3-4 J.4 ACCESS AND OCCUPANCY OF VITAL AREAS................................J.4-1 J.4.1 DOSE RATES OUTSIDE THE REACTOR BUILDING........................J.4-1 J.4.2 VITAL AREAS AND ACCESS ROUTES OUTSIDE THE REACTOR BUILDING.................................................................J.4-2 J.4.3 VITAL AREAS AND ACCESS ROUTES INSIDE THE REACTOR BUILDING................................................................J.4-2

J.5 METHODS..................................................................................J.5-1 J.5.1 THE USE OF CO MPUTER CODES................................................J.5-1 J.5.2 SOURCE TERM DEVELOPMENT FOR PRIMARY CONTAINMENT....J.5-2 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Appendix J SHIELDING EVALUATION REPORT

TABLE OF CONTENTS (Continued)

Section Page LDC N-9 9-0 0 0 J-iv J.5.3 SOURCE TERM DEVELOPMENT FOR SECONDARY CONTAINMENT........................................................................J.5-2 J.5.3.1 Parametric Studies for Direct Piping Dose in S econdary Containment .....J.5-3 J.5.3.2 Dose Rate and Cumulative Dose Calculation Procedure........................J.5-3 J.5.3.2.1 Calculation of Airborne Gamma Do ses Inside Secondary Containment..........................................................................J.5-4 J.5.3.2.2 Procedure for the Calcul ation of Radiation Zone Dose in Secondary Containment............................................................J.5-5 J.5.3.3 Calculation of Radiation Doses Due to Special Systems and Components Inside Secondary Containment......................................J.5-6 J.5.3.3.1 Source Term Assumptions in Secondary Containment.......................J.5-6 J.5.3.3.2 Secondary Contai nment Analysis Method......................................

J.5-8 J.5.3.3.3 Calculation of Radiation Doses Inside Secondary Containment on Generic Mechanical Equipment..................................................J.5-8 J.5.4 SOURCE TERM DEVELOPMENT FOR C1E/SRM EQUIPMENT OUTSIDE THE REACTOR BUILDING ...........................................J.5-9 J.5.5 METHODOLOGY OF BETA DOSE ANALYSIS................................J.5-9

J.6 RESULTS....................................................................................J.6-1 J.6.1 PRIMARY CONTAINMENT RADIATION RESULTS.........................J.6-1 J.6.2 SECONDARY CONTAINMENT RADIATION RESULTS....................J.6-2 J.6.3 RADIATION RESULTS IN THE VITAL AREAS AND ACCESS ROUTES...................................................................................J.6-3

J.7 REFERENCES..............................................................................J.7-1

ATTACHMENTS J.A UNISOLATED LEAKING BUILD ING PATH REPORT..........................J.A-1 J.B SOURCE TERM DEVELOPMENT AND PARAMETRIC STUDIES FOR SECONDARY CONTAINMENT................................................J.B-1 J.B.1 RADIOACTIVE SOURCE TERMS IN SECONDARY CONTAINMENT..J.B-1 J.B.2 AIRBORNE DOSE IN SECONDARY CONTAINMENT.......................J.B-2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Appendix J SHIELDING EVALUATION REPORT

TABLE OF CONTENTS (Continued)

Section Page J-v J.B.3 PARAMETRIC STUDIES FOR DIRECT PIPING DOSE.......................

J.B-12 J.B.3.1 Functional Dependence of Various Parameters on Secondary Containment Dose Rates............................................................J.B-13 J.B.3.2 Parametric Study Procedures.......................................................J.B-14 J.B.3.3 Direct Dose Para metric Study Results Inside Secondary Containment......J.B-14 J.B.3.4 Correction Factor Method of Determining Direct Doses in Secondary Containment...........................................................................J.B-15 J.C PROCEDURE FOR THE CALCULATION OF SECONDARY CONTAINMENT RADIATION ZONE GAMMA DOSES........................J.C-1 J.C.1 INTRODUCTION.......................................................................J.C-1 J.C.2 DEFINITION OF TERMS.............................................................J.C-2 J.C.3 ASSUMPTIONS, APPROXIMA TIONS, AND LIMITATIONS...............J.C-4 J.C.3.1 Basic Assumptions to be Used in the Analysis..................................J.C-4 J.C.3.1.1 Assumptions Used in the Calculation of Airborne Dose Rate Inside Secondary Containment...........................................................J.C-5 J.C.3.1.2 Assumptions Used for the Calc ulation of Shine or Streaming Dose From Primary Containment......................................................J.C-6 J.C.3.1.3 Assumptions and A pproximations Used in the Calculation of Direct Doses.................................................................................J.C-6 J.C.3.2 Limitations.............................................................................J.C-7 J.C.4 PROCEDURES FOR THE CALCULATION OF SECONDARY CONTAINMENT RADIATION ZONE DOSES..................................J.C-8 J.C.4.1 Procedure A: Radiation Zone Dose Calculation...............................J.C-8 J.C.4.2 Procedure B: Airborne Dose Calculation in Sec ondary Containment......J.C-8 J.C.4.3 Procedure C: Primary Containment Shine Dose Calculation................J.C-9 J.C.4.4 Procedure D: Direct Dose Calculation...........................................J.C-9 J.C.4.5 Procedure E: QAD-P5A Modeling Procedure..................................J.C-11 J.C.4.6 Procedure F: Streaming Dose Calculation .....................................J.C-12

J.D CALCULATION OF THE RADIATION.............................................J.D-1 J.D.1 DESCRIPTION OF THE STANDBY GAS TREATMENT SYSTEM FILTERS..................................................................................J.D-1 J.D.2 CALCULATION OF TIME-DEPENDENT FILTER ACTIVITY CONCENTRATION....................................................................J.D-2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Appendix J SHIELDING EVALUATION REPORT

TABLE OF CONTENTS (Continued)

Section Page J-vi J.D.3 CALCULATION OF RADIATION DOSE FROM THE STANDBY GAS TREATMENT SYSTEM FILTER...................................................J.D-7 J.E BETA DOSE CALCUL ATION METHOD...........................................J.E-1

J.F PRIMARY CONTAINMENT ANALYSES..........................................J.F-1 J.F.1 STATEMENT OF PROBLEM........................................................J.F-1 J.F.2 BASIC APPROACH....................................................................J.F-1 J.F.3 DRYWELL...............................................................................J.F-2

J.F.3.1 Sources.................................................................................J.F-2 J.F.3.1.1 Reactor (Normal Operation - Drywell).........................................J.F-3 J.F.3.1.2 Systems (Normal Op eration - Drywell)........................................J.F-4 J.F.3.1.3 System (Post-Loss-of-Coolant Accident) - Drywell..........................J.F-4 J.F.3.1.4 Airborne - Drywell.................................................................J.F-5 J.F.3.1.5 Plateout - Drywell..................................................................J.F-5 J.F.3.1.6 Wetwell - Drywell.................................................................J.F-6 J.F.4 WETWELL...............................................................................J.F-6

J.F.4.1 Sources.................................................................................J.F-7 J.F.4.1.1 Airborne - Wetwell................................................................J.F-7 J.F.4.1.2 Plateout - Wetwell..................................................................J.F-8 J.F.4.1.3 Suppression Pool - Wetwell......................................................J.F-8 J.F.5 QAD-CG MODEL.......................................................................J.F-8 J.F.6 CODES....................................................................................J.F-9 J.F.6.1 FSPROD...............................................................................J.F-9 J.F.6.2 ORIGEN2..............................................................................J.F-10 J.F.6.3 QAD-BR...............................................................................J.F-10 J.F.6.4 QAD-CG...............................................................................J.F-10 J.F.6.5 KAP-V..................................................................................J.F-10 J.F.6.6 ANISN..................................................................................J.F-10

J.G BETA DOSE CONTRIBUTION IN PRIMARY CONTAINMENT..............J.G-1

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Appendix J SHIELDING EVALUATION REPORT

TABLE OF CONTENTS (Continued)

Section Page J-vii J.H VITAL AREAS AND ACCESS ROUTES ANALYZED FOR POST-LOSS-OF-COOLANT ACCI DENT OPERATIONS........................J.H-1 J.H.1 SOURCE OF RADIOACTIVITY TO THE REACTOR BUILDING ELEVATED VENT.....................................................................J.

H-1 J.H.1.1 Reactor Building Air Discharge Rate.............................................J.H-1 J.H.2 POSTACCIDENT DESI GN DOSE (PADD).......................................

J.H-2 J.H.2.1 Assumptions Used in /Q Calculation Methodology...........................J.H-3 J.H.2.2 Integrated Activity Equa tions Used in this Analysis............................J.H-4

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Appendix J SHIELDING EVALUATION REPORT

LIST OF TABLES

Number Title Page J-viii J.3-1 Distribution of Fission Products in the Worst Post-Loss-of-Coolant Accident Situation for Areas In side Containment Depressurized Reactor Coolant System.......................................................J.3-5 J.3-2 Distribution of Fission Products in the Worst Post-Loss-of-Coolant Accident Situation for Areas In side Containment Pressurized Reactor Coolant System.......................................................J.3-6

J.3-3 Distribution of Fission Products in the Worst Post-Loss-of-Coolant Accident Situation for Areas Outside Containment.......................J.3-7

J.3-4 System Operation and S ource Term Assumptions........................

J.3-8 J.5-1 Generic Mechanical Equipment..............................................J.5-11

J.6-1 Six-Month Total Integrated Dose (Loss-of-Coolant Accident) to Areas Containing C1E Equipment Outside the Reactor Building................................................................J.6-5

J.6-2 Vital Areas and Access Route List of Radiation Exposure to Personnel During the Require d Post-Loss-of-Cool ant Accident Operations........................................................................J.6-6

J.A-1 System Flow Diagrams Employed to Perform the Review..............J.A-3

J.B-1 Gamma Energy Concentration (photons/sec-cm

3) in Liquid-Containing Systems....................................................J.B-17 J.B-2 Comparison of Direct Dose Rate Results...................................J.B-18 J.C-l Diameter Correction Factor (F D) for Targets in Contact With the Source Piping....................................................................J.

C-13 J.D-1 Direct Gamma Dose Rate and Integrated Dose Results for Targets in the Standby Gas Treatm ent System Room..............................

J.D-9 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November1998 Appendix J

SHIELDING EVALUATION REPORT

LIST OF TABLES (Continued)

Number Title Page J-ix J.E-1 Dose Rate Reduction Factors for the Post-Loss-of-Coolant Accident Beta Energy Groups at Finite Volumes.........................J.E-5 J.F-1 Integrated Dose in Drywell...................................................J.F-11

J.F-2 Integrated Dose in Wetwell...................................................J.F-12

J.F-3 Approximate Dose Rate Reduction Factor Ve rsus Distance from Core Mid-Plane for Reactor Integrated Dose..............................J.F-13

J.F-4 Suppression Pool and Syst em (Loss-of-Coolant Accident) Liquid Source Terms 0-6 Month Average After

Loss-of-Coolant Accident.....................................................

J.F-14 J.F-5 Airborne Source Term s 0-6 Month Average After Loss-of-Coolant Accident.....................................................

J.F-15 J.F-6 Drywell Plateout Source Terms 0-6 Month Average After Loss-of-Coolant Accident.....................................................

J.F-16 J.F-7 Time Mesh Spacing Used in Source Calculations (Minutes)............J.F-17

J.F-8 Source Energy Gr oup Structure..............................................J.F-18

J.G-1 Dose Rate Re duction Factors for the Post-Loss-of-Coolant Accident Beta Energy Groups at Finite Volumes.........................J.G-5 J.H-1 Post-Loss-of-Coolant Accident /Q Values Used for Calculations of Integrated Doses Outside the Reactor Building........................J.H-9

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Appendix J

SHIELDING EVALUATION REPORT

LIST OF FIGURES

Number Title LDCN-06-039 J-x J.5-1 Dose Model Liquid Source

J.5-2 Sixth Month Integrated Fluid Contact Dose for MS, RCIC (Steam) System, and MSLC System Upstream of the Header J.5-3 Sixth Month Integrated Fluid Contact Dose for Pi pes Containing Liquid Source Term (RHR, HPCS, LPCS, RCIC Liquid Systems)

J.6-1 Forty-Year Integrated Dose - Tu rbine Generator Building (El. 441 ft 0 in.) (Sheets 1 and 2)

J.6-2 Forty-Year Integrated Dose - Tu rbine Generator Building (El. 471 ft 0 in.) (Sheets 1 and 2)

J.6-3 Forty-Year Integrated Dose - Tu rbine Generator Building (El. 501 ft 0 in.) (Sheets 1 and 2)

J.6-4 Forty-Year Integrated Dose - Radwaste Building (El. 437 ft 0 in.)

J.6-5 Forty-Year Integrated Dose - Radwaste Building (El. 467 ft 0 in.)

J.6-6 Forty-Year Integrated Dose - Radwaste Building (El. 484 ft 0 in.)

J.6-7 Forty-Year Integrated Dose - Radwaste Building (El. 501 ft 0 in.)

J.6-8 Vital Areas and Access Routes - Radwaste Building (El. 437 ft 0 in.)

J.6-9 Vital Areas and Access Routes - Radwaste Building (El. 467 ft 0 in.)

J.6-10 Vital Areas and Access Routes - Radwaste Building (El. 484 ft 0 in.)

J.6-11 Vital Areas and Access Routes - Radwaste Building (E1. 501 ft 0 in.)

J.6-12 Vital Areas and Access Routes - Diesel Generator Building (El. 441 ft 0 in.)

J.6-13 Vital Areas and Access Routes C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Appendix J

SHIELDING EVALUATION REPORT

LIST OF FIGURES (Continued)

Number Title J-xi J.6-14 Vital Areas and Access Routes - Post-LOCA Sampling (Roof)

J.6-15 Vital Area and Access Routes - Tech nical Support Center (El. 437 ft 0 in.)

J.6-16 Vital Area and Access Routes to Reactor Building Railroad Bay (El. 441 ft 0 in.)

J.6-17 Vital Areas and Access Routes -

Reactor Building (El. 471 ft 0 in. and 501 ft 0 in.)

J.6-18 Vital Areas and Access Routes -

Reactor Building (El. 522 ft 0 in. and 548 ft 0 in.)

J.B-1 Model of the Primary and Secondary Containment

J.B-2 Time-Dependent Gamma Dose Rate for a Semi-I nfinite Cloud of Fission Products at Secondary Cont ainment Concentrations

J.B-3 Illustration of Parameters Used in the Shielding Equation

J.B-4 Standard Gamma Dose Rate Curve for Liquid Containing Systems (RCIC Liquid System and RHR System)

J.B-5 Standard Integrated Gamma Dose Cu rve for Pipes in Liquid Containing Systems (RCIC Liquid System and RHR System)

J.B-6 Standard Gamma Dose Rate Curve for Pipes in the RCIC Steam System and MSIV-LCS Steam System Before the Header

J.B-7 Standard Integrated Gamma Dose Cu rve for Pipes in the RCIC Steam System and MSIV-LCS Steam System Before the Header

J.B-8 Standard Gamma Dose Rate Curve for Pipes in the MSIV-LCS Steam System After the Header

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Appendix J

SHIELDING EVALUATION REPORT

LIST OF FIGURES (Continued)

Number Title LDCN-06-039 J-xii J.B-9 Standard Integrated Gamma Dose Curve for Pipes in the MSIV-LCS Steam System After the Header

J.B-10 Deleted

J.B-11 Deleted

J.B-12 Radial Distance Correcti on Factor for Liquid Sources

J.B-13 Pipe Length Correction Factor for Liquid Sources

J.B-14 Pipe Diameter Correctio n Factor for Liquid Sources

J.B-15 Radial Distance Correcti on Factor for Gaseous Sources

J.B-16 Pipe Length Correction Factor for Gaseous Sources

J.B-17 Pipe Diameter Correctio n Factor for Gaseous Sources

J.B-18 Parameters Used for the Calculation of Length Correction Factor

J.C-1 Calculation of Le ngth Correction Factor

J.C-2 Procedure A: Procedure for Calculating Radiation Zone Doses

J.C-3 Procedure B: Procedure for Calculating Airborne Gamma Dose Rate and Integrated Doses J.C-4 Procedure C: Procedure for the Calculation of Cont ainment Shine Dose J.C-5 Procedure D: Procedure for the Ca lculation of Direct Dose Rate and Integrated Dose

J.C-6 Time-Dependent Gamma Dose Rate for a Semi-I nfinite Cloud of Fission Products at Secondary Cont ainment Concentrations

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Appendix J

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LIST OF FIGURES (Continued)

Number Title J-xiii J.C-7 Time-Dependent, Integrated Gamma Dose Rate for a Se mi-Infinite Cloud of Fission Products at Secondary Containment Concentrations (0.5%/Day Primary Containment Leakage Rate)

J.C-8 Gamma Dose Rate at a Target 8 ft Away from Standard Pipe

J.C-9 Gamma Integrated Dose at a Ta rget 8 ft Away from Standard Pipe

J.C-10 Pipe Diameter Correction Factor

J.C-11 Radial Distance Correction Factor

J.C-12 Pipe Length Correction Factor

J.C-13 Dose Rate Versus Concrete Shield - Th ickness for Standard Pipe (8 in. Sch 40)

J.C-14 Pipe Diameter Correction Factor for Targets Located Axia lly in Line with Source Piping

J.C-15 Distance Correction Factor for Targets Located Axially in Line with Source Piping

J.D-1 Standby Gas Treatment Filter

J.D-2 Geometry of Prefilters and HEPA Filters

J.D-3 Geometry of Charcoal Filters

J.E-1 Total Integrated Beta Cloud Airborne Dose as a Function of Size J.E-2 Integrated Beta Infinite Airborne Dose for the Reactor Building

J.F-1 Geometry Examples

J.F-2 Basic QAD-CG Drywell Model

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Appendix J

SHIELDING EVALUATION REPORT

LIST OF FIGURES (Continued)

Number Title J-xiv J.F-3 Isometric of Drywell Model J.F-4 Isometric of El. 513 ft 6 in. to 520 ft 6 in.

J.F-5 Plan at El. 499 ft 6 in.

J.F-6 Plan at El. 506 ft 6 in.

J.F-7 Plan at El. 513 ft 6 in.

J.F-8 Plan at El. 520 ft 6 in.

J.F-9 Plan at El. 527 ft 6 in.

J.G-1 Total Integrated Beta Cloud Airborne Dose in Primary Containment as a Function of Size

J.G-2 Integrated Beta Infinite Airborne Dose for Primary Containment

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-000 J-xv

SUMMARY

The Three Mile Island (TM I-2) accident has gene rated a concern that during an accident in which significant core damage o ccurs, the postaccident operations requiring the use of systems containing contaminated fluid may induce abnormally high radia tion doses to safety-related equipment and components which ma ke it difficult to operate the systems. The NRC initially addressed this concern with NUREG-0578 and NUREG-0737 and recommended a design review to evaluate the functional capability of safety-related equipment and radiation exposure to personnel during the postula ted post-LOCA operations.

Radiation levels have been determined for all areas containing sa fety-related equipment, vital areas, and access routes which are require d for the postulated post-LOCA operation.

Radiation levels determined for safety-related equipment insi de primary containment. The analysis included the shadow shie lding effect of inst alled equipment and th e effect of iodine plateout were used to more accurately calculate the radiation levels inside containment.

Radiation levels were determined for safety-related equipment. The radiation source term leaking into secondary containm ent was reduced by the loss of ha logens to plateout inside primary containment.

Radiation levels calculated for safety-related equipment outsi de secondary containment are reported in Table J.6-1.

Figures J.6-8 through J.6-18 identify the vital areas which re quire personnel access on either a continuous or infrequent basi s during post-LOCA operations.

Safety-related equipment will either be qualified for the radiation level it functions in, or it will be relocated to a radiation zone it is qualified for, or it will be repl aced with comparable equipment which is qualified for the particular radiation level that has been determined.

Vital areas and access routes were evaluated for post-LOCA ope rations and are reported in Table J.6-2 and Figures J.6-8 through J.6-18. All areas and access routes are in compliance with NUREG-0737.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April2000 LDCN-99-000 J-xvi ABSTRACT

This report presents a radiation shielding design review of th e equipment and systems of the Energy Northwest Columbia Generating Station.

The original repor t was prepared in September 1982. The equipment and systems are evaluated on the ba sis of a postulated accident which in addition to normal plant radiat ion levels during its 40-year life may contain highly radioactive fluids. This design review recommended by the NRC (NUREG-0578 and

NUREG-0737) evaluates the func tional capability of safety-related equipment and personnel radiation exposure during the postaccident operations.

This design review evaluates th e postaccident radiati on conditions for personne l located in vital areas (areas which require ac cess or occupancy during the post-LOCA scenario) on either a continuous or infr equent basis.

The postulated loss-of-coolant accident (LOCA) scenarios and the operations of the safety-related systems were reviewed. Radioactive sources contained within each system were developed. Radiation levels were calculated at safety-related equipment locations, as well as at selected locations outside the reactor building to which access may be re quired for postaccident operations.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.1-1 J.1 INTRODUCTION This report presents a detailed description of the resu lts and the review of plant shielding and radiation environmental cond itions for equipment and syst ems which may be used in postaccident operations for Columbia Generating Station (CGS). The review was initiated in response to Section 2.1.6.b of NUREG-0578, "TMI-2 Lessons Learned Task Force Status

Report and Short-Term Recomme ndation," and to Part II.B.2 of NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident."

The design review determined the postaccid ent radiation environm ental conditions for equipment required for postaccid ent operations inside the prim ary containment, inside the secondary containment, and outside the secondary containment.

The 6-month total postaccident ra diation dose rate as a function of time and the integrated dose were calculated at safety-related equipment lo cations inside the CGS reactor building, inside primary containment, and at selected locations (vital areas) outside the reactor building.

Section J.2 discusses the regulatory re quirements on which this report is based and provides a description of the tasks performe d for this shielding evaluation.

Section J.3 provides the systems review and source term assumptions used as input for the definition of the postaccident radiological environment.

Section J.4 discusses the work performed during th is project relating to safety-related equipment located outside of the reactor building and the access and occupancy of vital areas. This consists of the calculation of dose rates outsi de the reactor building.

Section J.5 discusses the methods of calculation including the use of computer codes, identifying the parameters that have a significant effect on the radiation dose rates, and the dose rate and cumulative dos e calculation procedure.

Section J.6 presents a summary of the results.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.2-1 J.2 REQUIREMENTS General Design Criterion 4 (10 CFR 50 Appendix A) requires that systems and components important to safety be designed to accommodate the environmental conditions associated with accidents. The Th ree Mile Island (TMI-2) accident has generated a concern that during an accident in which significant core damage occurs, the posta ccident operations requiring the use of systems containing contaminated fluid may induce abnormally high radiation doses to safety-related equipment and components which ma y make it difficult to operate the systems.

The NRC Lessons Learned Task Force initially addressed this c oncern in Section 2.1.6.b of NUREG-0578 (Reference J.7-1) and recommended a design revi ew be performed on such systems so that the functional capability of safety-related equipment located in close proximity to the resulting high radiation fi eld will not be unduly degraded.

Described in this section is a discussion of the current regulatory requirement s and guidelines used.

J.2.1 SHIELDING EVALUATION REGULATORY REQUIREMENTS

NUREG-0578 Section 2.1.6.b requires that each licensee perfor m a radiation and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The scope of the review includes the following:

a. Identification of the locations of vital areas and safety-related equipment,

b. Evaluation of the radiation level at each location, and

c. Provision for adequate a ccess to vital area s and assurances of postaccident equipment operation through design cha nges, increased permanent or temporary shielding, or postaccident procedural controls.

To perform this review, the NRC has pr ovided guidance in the following documents

("documents of record"): a. NUREG-0578, Secti on 2.1.6.b, Reference J.7-1 , b. NUREG-0588, Revision 1, Section 1.4, Reference J.7-2 , c. NUREG-0660, Sec tion II.B.2, Reference J.7-3 ,

d. Clarification Letter to NUREG-0578, dated September 5, 1980,Section II.B.2, Reference J.7-4 ,

e. NUREG-0737, Sec tion II.B.2, Reference J.7-5 ,

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.2-2 f. IE Bulletin No.79-01B, Reference J.7-6 , and g.. IE Bulletin 79-01B, Supplement 2, dated September 30, 1980, Reference J.7-7.

The regulatory requirements in the above mentioned documents ar e summarized in the following sections.

J.2.1.1 Accident An alysis Requirements

The postaccident radiation environment should be based on th e most severe design basis accidents (DBA) during or following which equipm ent must remain functional. This includes the consideration of the entire spectrum of lo ss-of-coolant accident (LOCA) events which can lead to a degraded core condition. Thes e accident conditions include the following:

a. Loss-of-coolant accident events whic h completely depressurize the primary system, and

b. Loss-of-coolant accident events in which the primary system may not be depressurized.

J.2.1.2 Source Term Assumptions

The radioactive source terms for the postula ted accident conditions as described in Section J.2.1.1 should be equivalent to the s ource terms recommended in Regulatory Guides 1.3 and 1.7 and Standard Review Plan Section 15.6.

5. The source term assumptions consistent with current licen sing requirements used for equi pment qualification and access evaluations are summarized as follows:

a. The fission product fractions assumed to be released fr om the fuel rods during a LOCA are the following:

Noble gases 100%

Halogens 50%

Remaining fission products 1%

For the analyses, 50% of the halogens a nd 1% of the solids were assumed to be diluted into the suppression pool and li quid carrying systems. The halogens were also assumed to be in the airbor ne source while iodines were assumed in the plateout source. Thus, some care is necessary in summing calculated doses to prevent double counting of the sour ces. The post-LOCA source contribution from liquid and plateout sources are analyzed separately and the worst dose is C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 J.2-3 tabulated for that evaluation rather th an the sum of both doses. Thus, double counting of the fission product fracti ons is eliminated where possible;

b. The above release is assumed to occur and be distribut ed instantaneously at the start of the accident. The plateout is assumed to occur over an effective time of 5 hr after the accident;

c. Until depressurized, liquid in the reactor coolant system (RCS) and other systems which are not isolated from the core and which contain the reactor coolant at the start of the LOCA contain 100% noble gases, 50% halogens, and 1% of the remaining fission products. These radioactive materials are mixed homogeneously in a volume no great er than the RCS liquid space;

d. Liquid in the suppression pool and any system not isolated from the core at the start of the LOCA, and containing only liquid from a depressurized source, is assumed to contain 50% ha logens and 1% of the re maining fission products. These radioactive materials are dilute d homogeneously in a volume no greater than the combined volumes of the suppression pool and the RCS liquid space;

e. The primary containment atmosphere a nd systems which are not isolated from the primary containment atmosphere at the start of the LOCA are assumed to contain at least 100% noble gases and 50% halogens initially. These radioactive materials are diluted homogeneously in a volume no greater than the combined volumes of the drywell and suppression pool air spaces; and

f. Primary containment plateout source term is obtained by allowing the airborne halogens released (50%) to plateout on primary containment surfaces in accordance with the guidelin es presented in NUREG/

CR-0009 until the airborne elemental iodine concentration is decreased by a factor of 200.

g. Until the reactor vessel is depressurized, gases in th e steam lines and any other vapor-containing lines not isolated from the core at the start of the LOCA are assumed to contain at le ast 100% noble gases and 25%

halogens. These are diluted uniformly in a volume no greater than the RCS steam space and adjoining unisolated steam lines.

J.2.1.3 Vital Area Access Requirements

As defined in NUREG-0737 (Reference J.7-5), a vital area is an area which will or may require occupancy to permit an ope rator to help in the mitigati on of an accident or perform postaccident operations. The accident scenarios discussed in Section J.2.1.1 and the source term assumptions in Section J.2.1.2 are used for the evaluation of vital area access and occupancy. The total radiation exposure to personnel in vital areas should not be in excess of C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 J.2-4 5 rem whole body, or its equivalent, to any part of the body, for the duration of the accident.

For areas requiring continuous o ccupancy (e.g., the control r oom, onsite technical support center, etc.), the dose rate crit eria limits the total radiation exposure to less than 15 mrem/hr (averaged over 30 days).

J.2.1.4 Systems Containing the Sources

Systems considered in the shielding review are those system s that could have the potential of containing a high level of radioactivity postaccident. For those systems connected directly to the RCS or to the primary containment atmosphere and not isolated at the start of the accident, the radioactivity is assumed to be instantane ously mixed within the unisolated parts of the system.

J.2.1.5 Safety-Related Equipment (C1E/SRM)

The safety-related (C1E

/SRM) equipment list contains all equipment necessary to mitigate the consequences of an accident , bring the plant to a safe shutdown condition, and provide long-term cooling capability. This list includes equipment located inside as well as outside the primary containment.

J.2.2 SHIELDING EVALUATION TASK DESCRIPTION

The shielding evaluation tasks which have been completed to date are as follows:

a. Review all accident sc enarios and accident conditi ons that could result in a limiting radiation environment for all the pieces of safety-related equipment on the C1E/SRM (safety-related) list that are located in the reactor building;

b. Identify systems and components that could potentially c ontain radioactive materials postaccident;

c. Generate source term assumptions ba sed on regulatory requ irements discussed in Section J.2.1;

d. Calculate accident radiation service c onditions for the safety-related equipment located inside the reactor building;

e. Calculate gamma dose rates at selected locations outside the reactor building due to radioactive sources insi de the reactor building;

f. Identify vital areas and equipment to evaluate the acces s to and occupancy of the vital areas in accordance with the requirements listed in Section J.2.1; C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 1-0 0 0 J.2-5 g. Conduct a primary containment analysis of LOCA events in which the RCS may not depressurize (or may repressurize) w ith a degraded core condition. The primary containment radiation environm ent was determined with the use of 100% noble gases, 50%

halogens, and 1% of the rema ining fission products for the period of time during which the ac tivity is isolated to the RCS;

h. Calculate the radiation dose to safety-related equipment in the reactor building from post-LOCA airborne radiation a nd from normal piping sources inside primary containment streaming through the bioshield wall penetrations; and

i. The safety-related equipment list contains all equipm ent required to "mitigate" the consequences of an accident, bring the plant to a safe shutdown condition, and provide long-term cooling capability.

The completeness of the safety-related equipment list has been verified.

J.2.3 SHIELDING EVALUATION ITEM D ELETED FROM SHIELDING ANALYSIS CONSIDERATION

Columbia Generating Station has addressed all the issues ne eded to comply with the NUREG-660 II.B.2 position except as follows: Columbia Generating Station takes exception to the portion of the task that specifies that a review of "safety-related" equipment which may be degraded by radiation during postacci dent operation be pr ovided for a non-LOCA, high-energy line break source term. The pipe break/missile anal ysis described in Sections 3.5 and 3.6 addresses nonmechanistic pipe breaks inside and outside containment. These pipe breaks do not lead mechanistically to a radiation release due to fuel failures beyond those allowed in normal operation.

Hence, the source term id entified and applied outside containment is entirely hypothe tical and would be a new de sign basis beyond the scope of current regulations.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November1998 J.3-1 J.3 ANALYTICAL METHODOLOGY To develop the method used in the calculation of radiation doses, a review of all the postulated accident scenarios and system operations were performed. Source term assumptions were developed based on the results of accident analys is and system review, as well as the regulatory guidelines described in Section J.2.1. The systems and components inside the reactor building that have the potential of becoming contaminated during or following the accident were identified.

The following subsections describe these activities in greater detail. Section J.3.1 describes the accident scenario chosen fo r this analysis. Section J.3.2 identifies all the contaminated systems. Section J.3.3 describes the source term assumptions generated for each contaminated system. Section J.3.4 identifies the time period considered for this study.

J.3.1 ACCIDENT SCENARIO

The accident analyses consistent with FSAR Chapter 15 for small- and large-break loss-of- coolant accidents (LOCAs) were considered. The entire spectrum of LOCA conditions that could result in a degraded core configuration was reviewed and it was concluded that there is no single accident scenario that could result in a limiting radiation environment for all the safety-related equipment located in the reactor building. Th erefore, the accident scenario chosen here is based on a nonmechanistic LOCA in which core damage is experienced at the beginning of the accident. Prim ary containment isolation is a ssumed to be achieved prior to radioactivity transport.

A review of the postacciden t operation of the C1E/SRM (safety-related) systems was conducted. The result of this review indicated that the worst-case ac cident for the steam supply system (highest source term) was the pr essurized reactor coolan t system (RCS). For the liquid systems [the emerge ncy core cooling system (ECC S), the residual heat removal (RHR), and the reactor core isolation cooling (RCIC) systems], as well as the primary containment atmosphere and primary containment atmosphere control (CAC) system, the worst-case accident is the depressurized reactor coolant syst em with the post-LOCA core release functions dispersed with in the primary containment.

J.3.2 CONTAMINATED SYSTEMS To perform the radiation dose calculations, it was necessary to identify the systems which would or could contain highly ra dioactive materials during the postaccident period. Systems required to operate during the postaccident period are as follows:

a. Systems necessary to mitigate the consequences of a large- or small-break LOCA, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-039 J.3-2 b. Portions of systems that are in communication with systems containing radioactive liquids or gases, and

c. Defined by the NRC as being required, such as the gaseous radwaste system (see Section J.3.2.3).

J.3.2.1 Systems Included for Primary Containment Analysis

The following systems were considered:

a. High-pressure core spray (HPCS),

b. Low-pressure core spray (LPCS),

c. RHR,

d. RCIC,

e. Floor drains and equipment drains (FDR-EDR),

f. Reactor water cleanup (RWCU),

g. Main steam (MS),

h. Reactor recirculation (RRC),

i. Sample lines (PSR),

j. Automatic depressurization system (ADS), and

k. Low-pressure coolant injection (LPCI) function of the RHR system after depressurization.

J.3.2.2 Systems Included for S econdary Containment Analysis The following systems were considered:

a. RCIC, b. RHR,

c. LPCI,

d. LPCS,

e. HPCS,

f. MS, up to second isolation valve, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-039 J.3-3 g. MS line isolation valve-leakage control system (MSIV-LCS), h. Primary containment, i. Secondary containm ent atmosphere, and j. Standby gas treatment (SGT).

The following systems were also c onsidered due to their potential to affect isolation valves or extend the primary containment source terms into secondary containment.

a. Containment atmosphere monitoring (CMS), b. Containment supply purge (CSP),

c. Containment exhaust purge (CEP),

d. Blank penetrations,

e. Personnel access doors into the wetwell and drywell,

f. Instrumentation penetrations, and

g. All post-LOCA inboard a nd outboard isolation valves and their connected piping sources.

J.3.2.3 Systems Excluded

All systems required to m itigate the consequences of an accident have been included. Of those systems recommended for consideration in regulatory documents, one system (gaseous radwaste) has been excluded.

The gaseous radwaste is isolated by the primary containment and reactor vessel isolation control system and will not receive contaminat ed gas unless operation is manually initiated. The Columbia Generating Station (CGS) operating and accident procedures do not take credit for nor anticipate using this system. Since CGS philosophy is based on containment of the core releases within the primary containment, this system will not be required and was, therefore, excluded from consideration.

J.3.3 SOURCE TERM ASSUMPTIONS

Regulatory requirements specif y that source terms equivale nt to those recommended in Regulatory Guides 1.3 a nd 1.7 and Standard Review Plan Section 15.6.5 be used in the LOCA accident analysis. Additional guidan ce is given in NUREG-0588 (Reference J.7-2) and NUREG-0737 (Reference J.7-5) and is documented in Section J.2.1. Source term assumptions were generated based on the re view of the operation of the safety systems. Because a C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 J.3-4 nonmechanistic LOCA scenar io was chosen for this analysis, the worst contamin ated situation for the fluid contained within each system was conservatively assumed.

Tables J.3-1 , J.3-2 , and J.3-3 list the assumptions involved in the distri bution of fission prod ucts used in this analysis. These assumptions are consistent with the regulatory requirements discussed in Section J.2.1.

A review of the operation of each of the systems discussed in Section J.3.2 was also conducted. This review identified the source of contaminated fluid contained within each system postaccident. Using the source term assumptions discussed in Tables J.3-1 , J.3-2 , and J.3-3 , together with the results of this system re view, the limiting source term (activity divided by dilution factor) was determined for each system.

Table J.3-4 is a summary of the system operations and source term assumptions developed for each contam inated system identified in Section J.3.2.

J.3.4 TIME PERIOD CONSIDERED FOR STUDY

All systems were conservatively assumed to beco me contaminated at th e start of the accident and remain contaminated until the integrated radiation dose reached it asymptotic value. It was noted that the integrated dose becomes nearly asympt otic to a constant value beyond about 6 months. Therefore, 6 months is the time peri od chosen for accident dose qualification in this report.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.3-5 Table J.3-1 Distribution of Fission Products in the Worst Post-Loss-of-Coolant A ccid e nt Situation for A r eas Inside Con t ainment Depress u rized Reactor Coolant System

Primary Containment a Air and Steam Space Suppression Pool and R eactor Coolant System Wa t e r Volume Fission Products Fraction b Dilution Volume c Fraction b Dilution Volume c Noble gases 100% Drywell a i r plus 0% Suppression pool water and RCS

water volume Halogens 50%

d,e Suppression pool 50% Particulates 0%

Air 1% a A uniform distribution between drywell and su ppression pool atmosphere has been assumed.

b Expressed in percentage (%) of total core inventory at end-o f-life conditions (1000 days at 3556 MWt).

c Represents the total volume in which the fraction of core fissi on products is assumed to be homogeneously mixed.

d In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are conservative simultaneously. Instead , a set of mutually compatible assumptions will be used which gives the maximum dose for the location being considered. The post-LOCA source contributors are used to calculate independent doses for each contributor. The worst dose is tabulated for that system rather than the sum of all contributors (i.e., 50% ha logens airborne and 50% halogens in the water). Thus double counting of the fission produc t fractions is eliminated.

e First order iodine plateout occurs during the first 5-6 hr of the post-LOCA time frame when the elemental halogen concentration is reduced by a factor of 200. This methodology is in accordance with NUREG/CR-0009. Of the halogens released, 95.5%

is available for plateout. Virtually all of the available ha logens plateout within the initial 5 hr after the accident (0.5% remain airborne).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.3-6 Table J.3-2 Distribution of Fission Products in the Worst Post-Loss-of-Coolant Accident Situation for Areas Inside Containment Pressu rized Reactor Coolant System a Drywell Air Space a Suppression Pool Water Volume and Air Space a Reactor Coolant System Water Volume a Reactor Coolant System Steam Spac e a Fission products Fraction b Fraction b Fraction b Dilution Volume c Fraction b Dilution Volum e c Noble gases 0% 0% 100%d RCS water volume e 100%e Normal RCS steam spac e f Halogens 0% 0% 50%

g 25% Particulates 0%

0% 1% 0% a The react o r coolant system wi ll remain pressur i zed for a s hort period of time (17 hr) and then will be depressurized.

b Expressed in percentage (%) of total core inventory at end-o f-life conditions (1000 days at 3556 MWt).

c Represents the total volume in which the fraction of core fissi on products is assumed to be homogeneously mixed.

d The 100% of noble gases, present during the 17 hr of the pressurized RCS during a LOCA, are homogeneously mixed in the water a nd steam dilution volumes identified.

e The dilution volume is the RCS water volume plus the RWCU lines up to the isolation valves, RHR lines to the isolation valves, and the RRC lines during the 17 hr of the pressurized RCS scenario.

f The dilution volume is the normal RCS steam space plus the MS lines up to the isolation valves during the 17 hr of the pressurized RCS scenario.

g In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are conservative simultaneously. Instead , a set of mutually compatible assumptions will be used which gives the maximum dose for the location being considered. The post-LOCA source contributors are used to calculate independent doses for each contributor. The worst dose is tabulated for that system rather than the sum of all contributors (i.e., 50% ha logens airborne and 50% halogens in the water). Thus double counting of the fission produc t fractions is eliminated.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 1-0 0 0 J.3-7 Table J.3-3 Distribution of Fission Products in the Worst

Post-Loss-o f-Coolant Accident Situation for Areas Outside Containment Primary Containment Air Space Suppression Pool Water Volume Reactor Coolant System Steam Space a Reactor Coo l ant System Water Volume a Fission Products Fraction b Dilution Volume c Fraction b Dilution Volume c Fraction b Dilution Volume c Fraction b Dilution Volum e c Noble gases 100% Drywell 0% Suppression pool water plus RCS water 100% Normal 100% RCS Halogens 50%

d Air plus 50%e 25% RCS 50% Water Particulates 0% Suppression pool air 1% 0% Steam space 1% Volume a Based on pressurized reactor coolant system.

b Expressed in percentage (%) of total core at end-of-life conditions (1000 days at 3556 MWt).

c Represents the total volume in which the fraction of core fissi on products is assumed to be homogeneously mixed.

d 95% of the halogens released from the core are assumed to plateout within approximately 5 hr as allowed by NUREG/CR-0009. The plat eout dose was considered in the total calculation of radiation dose to equi pment inside primary containment.

e In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are simultaneously conservative. Instead , a set of mutually compatible assumptions will be used which gives the maximum dose for the location being considered. The post-LOCA sour ce contributions are used to cal culate independe nt doses for each contributor. The worst dose is tabulated fo r that system rather than the sum of total contributors (i.e., 50% ha logens airborne and 50% halogens in the water). Thus double counting of the fission of product fractions is eliminated.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 Table J.3-4 System Operation and Source Term Assumptions

System Operation Postaccident Contaminated Space Source Term Assumptions J.3-8 HPCS Suction from condensate storage tank and/or suppression pool and discharge to the reactor vessel.

Sup p ression pool (1) LPCS Suction from suppression pool and discharge to the reactor vessel.

Sup p ression pool (1) LPCI Suction from suppression pool and discharge to the core.

Sup p ression pool (1) (6)

RCIC steam

system Steam bleed-off from reactor steam space is used to drive the RCIC turbine, and exhausts into the suppression pool.

RCS steam space (2) RCIC liquid

system Suction from condensate storage tank or suppression pool and discharge to

the reactor vessel.

Sup p ression pool (1) RHR system (1) Shutdown cooling mode - suction from reactor recirculation system suction line and discharge into the reactor recirculation discharge line. (2) Alternate shutdown cooling mode - suction from suppression pool and discharge to core recirculates and

cools the water in the suppression pool. (3) Containment spray cooling mode - suction from suppression pool and discharge into the drywell and suppression pool. (4) Reactor steam condensing mode.

RCS liquid space

Suppression pool

Suppression pool

System mode deleted from plant Note a (1) Note b (1)

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007

Table J.3-4 System Operation and Source Term Assumptions (Continued)

System Operation Postaccident Contaminated Space Source Term Assumptions LDCN-06-039 J.3-9 Main steam supply (MS) Stagnant steam from the reactor vessel terminates at the second MSIV. RCS steam space (2)

MSIV-LCS (MSLC) Steam bleed-off from main steam line, diluted, and discharged into the SGTS. RCS steam space (2)

Note c SGT filters (SGTS) Process the halogens from primary containment leakage and MSIV-LCS. Primary containment and secondary

containment

atmosphere (3) Primary containment (PCN) Primary containment is isolated postaccident. Primary containment

atmosphere (4) Suppression

pool The primary function of the suppression pool is to contain and condense the blowdown from the RCS

postaccident. Suppression pool liquid (1) Secondary

containment (SCN) The primary function of the secondary containment is to contain all the leakage from the primary containment

postaccident. Primary containment

atmosphere (5) Sample lines Actuated to obtain primary containment atmosphere samples per NUREG-0737 (Reference J.7-5). Primary containment

atmosphere (2) Sample lines Actuated to obtain liquid samples per NUREG-0737. RCS liquid space (1)

Reactor water

cleanup (RWCU) Reactor water cleanup system isolated during post-LOCA. Liquid up to the second isolation valve is considered

contaminated. RCS liquid (1)

Reactor recirculation (RRC) Suction from RRC system suction line

and discharge into the reactor recirculation discharge line. RRC liquid; RCS

liquid (1) Floor drains and

equipment

drains (FDR/EDR) Liquid from ruptured pipes or leaky seals discharged into the suppression

pool. RCS liquid (1)

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007

Table J.3-4 System Operation and Source Term Assumptions (Continued)

System Operation Postaccident Contaminated Space Source Term Assumptions J.3-10 Automatic depressurization

system (ADS) Automatic or manual depressurization of the reactor vessel by blowdown of the RCS into the suppression pool. RCS steam (2)

Automatic depressurization

system (ADS) Alternate shutdown cooling mode with reflood of reactor vessel and discharge into suppression pool. Suppression pool (1)

Containment

monitoring system (CMS) Continues to monitor primary containment atmosphere conditions.

Isolation of primary

containment into secondary containment (4) Containment supply purge (CSP) Isolated - no action required. Isolation of primary containment into secondary containment (4) Containment exhaust purge (CEP) Isolated - no action required. Isolation of primary containment into secondary containment (4) Blank penetrations None Isolation of primary containment into secondary containment (4) Personnel access doors to primary

containment None Isolation of primary containment into secondary containment (4) Instrumentation

penetrations None Isolation of primary containment into secondary containment (4) All post-LOCA inboard and outboard isolation valves As defined per Columbia Generating Station system requirements post-

LOCA Isolation valves and their connected piping

which extends into secondary containment Note d Source Term Assumptions

(1) 50% halogens and 1% solid fission produc ts diluted with suppre ssion pool water plus RCS water. (2) 100% noble gases and 25%

halogens diluted with the RCS steam space.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 Table J.3-4 System Operation and Source Term Assumptions (Continued)

J.3-11 (3) 50% halogens leaked from the primary c ontainment is assumed to be deposited in the SGT filters at the rate of 0.67% per day. See Section J.5.3.3.1 for justification. 100% noble gases pass through al so but are not absorbed. (4) 100% noble gases and 50% hal ogens diluted with the primar y containment air space.

First order iodine plateout (0-95% elemen tal iodine) inside primary containment was considered. (5) Assumptions involved in the calculation of source terms for s econdary containment atmosphere are discussed in Section J.5.3.2.1. (6) Based on a pressurized reactor coolant system.

a According to accident mitigation procedures, this mode of operation is not used after a degraded core condition is identified.

b Full discussion of source term assumptions fo r alternate shutdown cool ing are presented in Section J.5.3.3.1. c For the portion of system after the distribution header, credit is take n for dilution by clean air. See Section J.5.3.3.1 for justification.

d For all isolated systems the source term for th e isolation valves will be primary containment atmosphere unless the penetration is filled with water that remains during the post-LOCA scenario. All penetrations and their associated isolation valves which contain a flowing fluid during post-LOCA operations are analyzed with the post-LOCA source term of that flowing fluid.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.4-1 J.4 ACCESS AND OCCUPANCY OF VITAL AREAS NUREG-0578 initiated the requirement for a design review to identify the location of vital areas in which personnel occ upancy may be unduly limited by the radiation fields during postaccident operations. It requi red that each licensee provide adequate access to vital areas through design changes, increased permanent or temporary shielding, or postaccident procedural controls. NUREG-0737 further make s the point that the purpose of this design review is to determine what ac tions can be taken over the short-term to reduce radiation levels and increase the capability of operators to control and mitigate the consequences of an accident.

This shielding evaluation includes the calculati on of gamma dose rates at selected locations outside the reactor building due to radioactive sources inside. The radioactive source terms obtained from ORIGEN computer calculations coupled with recommenda tions from Regulatory Guide 1.109 were the basis for the assumptions used in evaluating vital areas and access routes outside the reactor building.

J.4.1 DOSE RATES OUTSIDE THE REACTOR BUILDING

An analysis was conducted to dete rmine the dose rates at selected locations outside the reactor building for personnel access purpos es. The radiation level in the various areas outside the reactor building is defined by the following three radioactive sources:

a. Direct gamma ray dose from radioactive piping located inside the reactor building and attenuated through the walls of the reactor building,

b. Gamma shine dose from airborne activ ity inside the reac tor building, and

c. Gamma dose from airborne activ ity outside the reactor building.

Radiation levels outside the reactor building were determin ed by the zone dose method as discussed in Section J.5.4. Representative zones were chosen at selected locations outside the reactor building such as ground level outside th e railroad bay, sampling room, etc. The worst point in a zone was chosen to be the point direc tly outside the reactor bu ilding wall, at a height of 6 ft above floor elevation, at a lateral point determined by inspection to receive the highest dose along that wall.

The zones outside the reactor building are indi cated by the letters Y and Z in the various elevations. The shine dose contribution to areas outside the r eactor building (Zones Y and Z) were included in the dose c a lculations shown in Figures J.6-11 through J.6-18.

Attachment J.H presents the methodology used to calcula te the radiation dos es for the various vital areas.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.4-2 J.4.2 VITAL AREAS AND ACCESS ROUTES OUTSIDE THE REACTOR BUILDING Radiation calculated for the access routes we re based on the assump tion that no individual would be in an access route longer than 30 minutes for the first 8 hr af ter the postulated LOCA before reaching the vital area of interest.

The assumption was also made that no individual would occupy an infrequent occupied vital

area longer than 30 minutes for the fi rst 8 hr after the postulated LOCA.

All integrated radiation doses calculated for time spent in the access routes and vital areas were less than the guidelines presented in NUREG-0737.

J.4.3 VITAL AREAS AND ACCESS ROU TES INSIDE THE REACTOR BUILDING

Analysis has been completed to take credit for a vital area in the reactor building railroad bay and on the west side of the 522-ft el. The anal ysis of reactor building zones is discussed in Section J.5.3. The access route to the reactor building is discussed in Sections J.4.1 and J.4.2. See Section J.6.3 for a description of the access to th e 522-ft el. of the reactor building.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.5-1 J.5 METHODS Due to the large number of C1E/SRM components in primary containment, it was decided to calculate the worst point dose from each of the major sources in the drywell and wetwell, and then sum the doses for a conservative es timate of the total integrated dose.

The secondary containment radi ation dose assessment portion of the shielding evaluation was initiated by dividing the reactor building into radi ation zones. Because of the large number of radioactive piping and safety-related equipment in the building, the division of the various regions of the secondary containment into radiation zones permits a precise, detailed calculation of the total integrated dose at th e "worst target" locati on. The methods for performing the calculations are discusse d in detail in the following sections.

The radiation dose assessment of safety-related equipment outside of the reactor building was done by calculating the radiation dose of each vital area where safety-rel ated equipment was located. The as sumptions and methodology used to perform these cal culations are discussed in detail in the following se c tions and in Attachment J.H.

J.5.1 THE USE OF COMPUTER CODES

The two computer codes used in the primary containment shielding evaluation were ORIGEN2 and QAD-CG. Descriptions of the two codes are found in References J.7-8 , J.7-9 , J.7-10 , and J.7-17. ORIGEN2 was used to compute the radioactive source terms (inside containment) used by QAD-CG to calculate the radiation dos es from piping and various pieces of equipment.

The three computer codes used in the original secondary containment radiation shielding review were ORIGEN, SCAP-BR, and QAD-P5A.

Descriptions of the codes are found in References J.7-10 , J.7-11 , and J.7-18. ORIGEN computes the radioactive source terms used by QAD-P5A to compute the radiation from piping and other source configur ations to pieces of equipment. SCAP-BR computes th e radiation dose contribution to safety-related equipment in the reactor building from primary containmen t airborne radiation streaming through the bioshield wall penetrations.

ORIGEN and ORIGEN2 are fission product source term codes which solve the equations of radioactive growth and decay for large numbers of isotopes.

The codes have been used to calculate the radioactivity of fission products and fuel materials that were assumed to be released from the reactor core during the pos tulated loss-of-coolant accident (LOCA) to become the primary containment source terms for the dose rate calculations. SCAP-BR is similar to QAD-CG with the added capability to determine the radiati on dose contribution due to scattering.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.5-2 J.5.2 SOURCE TERM DEVELOPMENT FOR PRIMARY CONTAINMENT

The radiation level at any gi ven location inside the primar y containment of Columbia Generating Station (CGS) following the postulated LOCA such as that described in Section J.3.1 is determined from the follo wing major source contributors:

a. Gamma ray dose from airborne radioactive sources suspended in the drywell and wetwell inside primary cont ainment (airborne gamma dose), b. Gamma ray dose from piping and/or e quipment containing contaminated fluids which are recirculated inside prim ary containment (direct gamma dose), c. Gamma and beta ray dose from iodines plated out inside primary containment (iodine plateout), and

d. Beta ray dose emitted by airborne radioactive sources suspended in the drywell and wetwell inside primary cont ainment (airborne beta dose).

The initial phase of this analysis was concerne d with the determination of radioactive source terms for the liquids and gases inside primar y containment. The OR IGEN2 computer code was used for this calculation.

The fission product inventory at th e end of fuel life (1000 days irradiation at a power level of 3556 MWt) was assumed to be av ailable for release immediately following the accident. The re lease fractions and resulting c oncentrations of noble gases, halogens, and other fission prod ucts in the gaseous and liqui d fluids were computed.

A detailed description of the analysis including the assumptions used is provided in Attachment J.F.

J.5.3 SOURCE TERM DEVELOPMENT FOR SECONDARY CONTAINMENT

The radiation level at a given location inside the secondary c ontainment of CGS following an accident such as that described in Section J.3.1 is defined by the following major source contributors:

a. Gamma ray dose from airborne ra dioactive sources inside secondary containment (airborne gamma dose),

b. Gamma ray dose from radioactive sour ces suspended in the drywell and the wetwell inside primary contai nment (containment shine dose),

c. Gamma ray dose from piping and/or e quipment containing contaminated fluids which are recirculated inside the reactor building (direct gamma dose),

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.5-3 d. Beta ray dose emitted by airborne radioactive sources inside secondary containment (airborne beta dose), and

e. Gamma ray dose from liquid piping and airborne radioactive sources inside primary containment which stream thro ugh bioshield wall pe netrations into secondary containment (bioshield penetration streaming dose).

The initial phase of this analysis was concerne d with the definition of radioactive source terms for the liquid and gas containing systems. The ORIGEN comput er code was used for this calculation. The fission products at the end of fu el life (1000 days irra diation at a power level of 3556 MWt) were assumed to be available for release immediately following the accident.

The released fractions of noble gases, halogens, and other fission products to the gaseous and liquid sources were computed.

Subsequent fission product depletion and daughter product generation were then calculat ed for 20 time periods, covering a total period of 1 year.

A detailed description of the an alysis, including the assumptions us ed, as well as results of the source terms, is found in Attachment J.B and Reference J.7-12.

J.5.3.1 Parametric Studies for Direct Piping Dose in Secondary Containment

The purpose of the parametric study was to iden tify the parameters which have a significant effect on the radiation dose rates inside secondary containment.

The computer code QAD-P5A was used to develop a correlati on scheme for the significant parameters such that a simplified procedure for calculating radiation dose rates fo r complex source and receptor geometries can be developed. The dose rate at a target distance of 8 ft radially outwards from the centerline of an 8-in. schedule 40 pipe, infinitely long (standard pipe), was first calculated and defined as the standard dose rate. The resu lts of this parametr ic study were then co rrelated as a set of correction factors to the standard dose rate.

A simplified procedure was developed to calculate the dose rates and cumulate doses for complicated source-target configur ations by using these correction factors. The deve lopment of these correction fa ctors and the result of the parametric study inside secondary containmen t is discussed in Attachment J.B.

J.5.3.2 Dose Rate and Cumulativ e Dose Calculation Procedure

The results of the source term calculations and those of the parametric study were used to generate and cumulate doses for complicated source target configurations in side secondary containment. The following steps were taken to define the radiation service conditions for the pieces of safety-related equipment:

a. Based on the accident scenarios, contam inated systems, and assumptions defined in Section J.3, the radioactive source term s for liquid-containing and gas-containing system s were developed;

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 J.5-4 b. Radiation zones were selected and the radiation zone boundaries were carefully defined based on shield wall locations, contaminated piping locations, and locations of safety-relat ed C1E/SRM equipment;

c. The radiation environment in each secondary containment zone (zone dose) was calculated (see Attachment J.B for the procedure). A z one dose is the radiation dose (gamma) that bounds the magnitude of dose received by all the pieces of safety-related C1E/SRM equipmen t located within that zone;

d. The zone dose as calculated in step c was used, as a first cut, to qualify all the pieces of safety-related C1E/SRM equi pment located within that zone; and

e. For the pieces of safety-related C1E/SRM equipment that could not be qualified for the conservative radiation environment calculated in step c, the integrated dose for that piece of equipment was re defined based on a more realistic and refined approach.

J.5.3.2.1 Calculation of Airborne Gamma Doses Insi de Secondary Containment

The time-dependent post-LOCA acti vity levels as calculated by the ORIGEN computer code were used as input for the calculation of the airborne gamma dose rate s and integrated doses inside the cubicles in the seconda ry containment. The assumptions used in th is analysis are as follows:

a. Activity that leak s into the secondary containmen t is homogeneously mixed with the secondary containment atmosphere prior to its removal from the atmosphere

through the standby gas tr eatment system (SGTS);

b. The SGTS flow rate of 2430 scfm was assumed to be the flow rate of the effluent air. This is e quivalent to one reactor bu ilding air change per day;

c. Air that leaks out of the primary contai nment flows directly and totally into the secondary containment. Bypa ss leakage was not considered;

d. Geometric factors were us ed to convert the semi-infinite cloud gamma dose to a finite gamma dose; and

e. Primary containment leakage rate of 0.5 wt %/day was considered.

Justifications of the above assumptions are stated in Attachment J.B. The equations that were used for the gamma dose calculations are described in Attachment J.B. Primary containment airborne beta dose results are discussed in Attachment J.G.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November1998 J.5-5 J.5.3.2.2 Procedure for the Calc ulation of Radiation Zone Do se in Secondary Containment

As discussed previously, the ga mma radiation level at a given location inside the secondary containment of CGS following a LOCA is dete rmined for four types of radioactive source distributions:

a. Fission products suspended in the atmosphere of the secondary containment (airborne gamma dose),

b. Gamma irradiation from the pr imary containment (shine dose),

c. Direct gamma irradiation from the radioactive fluid contained inside recirculating pipes (direct dose), and

d. Gamma ray dose from liquid piping and airborne radioactive sources inside primary containment which stream thro ugh bioshield wall pe netrations into secondary containment (bioshield penetration streaming dose).

The dose contributed by each of these sources is determined by the location of the equipment, the time-dependent distribution of the source, and the effects of shielding.

A step-by-step procedure for calculating radioactive zone doses is shown in Attachment J.C. The methods presented in that procedure make it possible to calculate the worst case gamma dose from the above-mentioned source contributor s inside radiation z ones of the secondary containment. In general, this procedure for determining zone doses consists of a correction factor method for calcula ting direct dose rates.

As discussed in Attachment J.B , the correction factor met hod for calculating dose rates provides a convenient and fairly precise way of determ ining direct dose ra tes due to generic pipe segments. For radioactiv e fluid contained within components of geometry other than generic pipe segments, such as residual heat removal (RHR) heat exchangers, SGTS filters, hydrogen recombiners, etc., special QAD-P5A computer modeling was performed to calculate the gamma dose contributi on due to those systems. A brief description of the guidelines used in modeling special co mponents is found in Attachment J.B.

An evaluation of beta dose is nece ssary for qualification of safety-related equipment that is beta sensitive and not adequately protected against beta radiation. The beta dose analysis for secondary containment is presented in Section J.5.5. Beta dose is discus sed in more detail in Attachment J.D as related to secondary cont ainment radiation contributors.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-039 J.5-6 J.5.3.3 Calculation of Radia tion Doses Due to Special Sy stems and Components Inside Secondary Containment

As discussed in Attachments J.B and J.C , the correction factor me thod for calculating gamma dose rates and integrated doses is involved with the a pplication of the dose correction factors (pipe diameter, pipe length, and radial distance correction factors) to a standard dose rate curve. A standard dose is defined as the gamma radiation measured at a target distance of 8 ft and emitted by radioactive s ources contained within th e suppression pool liquid and recirculated within infinitely long 8-in. schedule 40 piping.

The systems that contain such radioactive fluids are the reactor coolant system, high-pressure core spray, low-pressure core spray, and residual heat removal systems. Other systems which contain fluids of different source terms and dilutions are considered special sources. The systems that need to be considered for special s ources are the following:

a. Standby gas treatment system filters,

b. Main steam system, and c. Main steam isolation valve l eakage control syst em (MSIV-LCS).

J.5.3.3.1 Source Term Assumptions in Secondary Containment

The assumptions for the calculati ons of source terms inside sec ondary containment for special source systems are listed as follows:

Standby Gas Treatment System Filters

a. The SGTS filters will be loaded by halogens at th e rate of 0.67% primary containment free volume per day. This consists of 0.5% per day of primary containment leakage and 0.17% per day of leakag e due to the MSIV-LCS system. No holdup of this activity in the secondary containment is assumed;

b. The released halogen fraction is 50% of the core halogen inventory. This halogen fraction is assumed to be compos ed of 95.5% elemental, 2% organic, and 2.5% particulate halogens; and

c. The particulate halogens are assumed to be homogene ously distributed within the prefilters and the particulate filters, while the elem ental and organic halogens are assumed to be homogeneously distribut ed within the charcoal filters.

Assumption a is consistent with the assumptions used in the accident analysis (Reference J.7-13 and Section J.3.1).

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-002 J.5-7 Assumption b is the NRC recommended assumption for the distribution of halogen inventory (Reference J.7-14).

Assumption c is necessary because the time-dependent distribution of activity within a filter is unknown. The homogeneous assumption, therefore, is considered appropriate and conservative for zone dose assessment.

Containment Atmosphere Control System The function of the CAC system was to process the primary containment atmosphere to remove oxygen after a LOCA accide nt. Therefore, this system wa s assumed to be filled with gaseous source containing 2.5%

halogens and 100% noble gase s diluted with the primary containment free volume, although it is now deactivated.

Main Steam System

The main steam lines are located inside and outside the primary containment; they include the main steam lines in the stea m tunnel and the RCIC turbine supply and exhaust lines. The radioactive source term for this system is assumed to be com posed of 100% noble gases and 25% halogens, distributed thro ughout the reactor coolant system (RCS) steam space.

Alternate (Suppression P ool) Shutdown Cooling

To prevent failure of the RHR pumps due to excessive radiation exposure, the alternate shutdown cooling mode is the onl y allowable mode for shutdown cooling once a degraded core condition has been identified.

A small pipe-break accident will take approxima tely 6 hr to depressu rize from 1000 psi to 150 psi through automatic depressu rization system (ADS) valve ac tuations. On ce a degraded core is identified and the reactor is sufficiently depressurized, w ithin 17 hr afte r the accident, the ADS valves actuation will be maintained to dilute the prim ary coolant source concentration with the suppression pool since the alternate shutdown cooling mode will be used for decay heat removal.

For the large pipe-break acciden t the primary coolant source c oncentration will be diluted by the water in the suppression pool due to blowdo wn of the vessel through the large break or automatic actuation of the ADS valves. Once the vessel has been depressurized the water level in the vessel will be maintained with the emergency core cooli ng system while decay heat is removed by suppressi on pool cooling.

Thus, in all degraded core scenarios the primary coolant is diluted with the suppression pool prior to initiating the suppression pool shutdown cooling mode.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-039 J.5-8 Main Steam Isolation Valve Leakage Control System

The MSIV-LCS system is a vacuum-type system which collects leakage between and downstream of the closed isolati on valves and then re leases it to the atmosphere through the SGTS. Leakage through the valve stems (maximum leakage of 11.5 scfh as described in Reference J.7-15) is directed to a distribution header or low-pressure manifold where clean air is brought in to dilute the contaminated steam before exhausting to the SGTS filter unit at a rated flow rate of 50 scfm. Thus the source term in the portion of piping system before the

distribution header is conservatively assumed to be the same as that of the main steam system. For the portion of the piping after the header, credit is taken for the dilution by the clean air.

This assumption is consistent with that recommended in Reference J.7-16.

J.5.3.3.2 Secondary Cont ainment Analysis Method

The correction factor method is used for the cal culation of the direct dose contribution due to the piping systems described in Section J.5.3.3, with the exception of the SGTS filter system.

A description of the analysis of the SGTS filter is documented in Attachment J.D. Generic piping dose rate and integrated dos e (dose at a target distance of 8 ft away from the centerline of an infinitely long 8-in. sc hedule 40 pipe) for each system we re developed using the source term assumptions discussed in Section J.5.3.1 and are shown in Attachment J.B. Parametric studies were also performed to i nvestigate the variation of dose ra tes due to pipe diameter, pipe length, and target distance for pipe segments containing source terms. The gaseous source

term correction factors derived as a result of this parametric study (described in Attachment J.B), together with the generic dose rate curves generated for each system, were used to calculate the direct ga mma dose contribution on a target.

J.5.3.3.3 Calculation of Ra diation Doses Inside Secondary Containment on Generic Mechanical Equipment

Table J.5-1 is a sample list of generic mechanical equipment that are on the safety-related equipment list. For conservatism, the dir ect dose on the containment pieces of generic mechanical equipment is assumed to be the fluid contact dose.

Figure J.5-1 is an illustration of the point where the direct dose is calculated on a piping segment.

The secondary containment source term assumptions developed in Section J.5.3 are used for the calculation of radioactive source terms for different systems, and the fluid contact dose was calculated using QAD-P5A by following the guidelines set forth in Attachment J.C. Figures J.5-2 through J.5-3 are 6-month integrated fluid cont act doses versus pipe diameter.

These curves are intended to gi ve conservative, uppe r-bound direct gamma dose estimates for the qualification of the pieces of generic mech anical equipment and co mponents in the various systems. To use these curves to calculate the direct doses on generic mechanical equipment, the following steps should be taken.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 J.5-9 a. Identify the system on which the equipment or component is located, b. Identify the diameter of the contam inated pipe on which the equipment is located, and

c. The 6-month integrated dose for that piece of equipment or component can be determined by reading th e appropriate curve.

J.5.4 SOURCE TERM DEVELOPMENT FOR C1E/SRM EQUIPMENT OUTSIDE THE REACTOR BUILDING The radiation level at any give n location outside the reactor building following the postulated LOCA as described in Section J.3.1 is determined from the following major source contributors:

a. Direct gamma dose from radioactive piping located inside the reactor building and attenuated through the walls of the reactor building,

b. Gamma shine dose from airborne activ ity inside the reac tor building, and

c. Gamma ray dose from airborne activity outside the reactor building.

A detailed description of the method of analysis, including the assumptions used, as well as results of the source terms is found in Attachment J.H.

J.5.5 METHODOLOGY OF BETA DOSE ANALYSIS

The finite source volume used for the beta dose an alysis in secondary c ontainment is a sphere surrounded by a shell of sufficien t thickness to stop all outside be ta particles from entering the source volume. This finite s pherical source volume is conserva tive for any generalized source shape (the dose at the center of the sphere is higher than the dose at any point of any generalized source shape of equal total volum e). A discussion of this beta analysis methodology is presented in Attachment J.D.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.5-11 Table J.5-1 Generic Mechanical Equipment Valve packing Lubricants Seals Expansion joints Pressure relief valve Flow element Rupture disk Gasket material Conductivity element

Valve Strainers Steam traps Filters (piping)

Temperature elements

Tanks Moisture separators Evaporator Heat exchanger Air washer (scrubber)

Pumps

Dose Model Liquid Source 970187.23 J.5-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.105cmTarget Location Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Sixth Month Integrated Fluid Contact Dose for MS, RCIC (Steam) System, and MSLC SystemUpstream of the Header 960222.63 J.5-2 Nominal Pipe Diameter (inches)

Interated Dose (10 6 Rads)0 4 8 12 16 20 24 35 30 25 20 15 10 5 0 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Sixth Month Integrated Fluid Contact Dose forPipes Containing Liquid Source Term (RHR, HPCS, LPCS, RCIC Liquid Systems) 960222.61 J.5-3 Nominal Pipe Diameter (inches)

Integrated Dose (10 6 Rads)0 4 8 12 16 20 24 14 13 1211 10 9 8 7 6 5 Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.6-1 J.6 RESULTS All loss-of-coolant accident (LOCA) scenarios and accident c onditions that could result in a limiting radiation environment for all the Columb ia Generating Station (CGS) safety-related equipment on the C1E

• list were reviewed and analyzed accordingly. Shielding (shield doors) was constructed for zones 522D, 572N, 572D, and 572H due to the radiation exposure of safety-related equipm ent in these zones.

In addition a shield wall was designed and installed on the southeast portion of the 501 ft el.

against the bioshield wall to protect C1E

• equipment from RRC pi ping radiation sources (normal operation) which stream throug h penetrations X-100A, X-105A, and X-100B.

The completeness of the safety-related equipment list has been verified. The safety-related equipment list contains all equipment required to "mitigate the c onsequences of an accident, bring the plant to safe shutdown conditions and provide long-term cooling capability."

Systems that could potentially contain radioactive material during and following the accident have been identified as listed in Sections J.3.2.1 and J.3.2.2.

The accident radiation dos es indicated in Section J.6.1 and Table J.6-1 generated as a result of this analysis, are intended solely for the purpose of the qualification of safety-related equipment.

J.6.1 PRIMARY CONTAINMENT RADIATION RESULTS Due to the large number of safe ty-related components it was deem ed impractical to calculate the integrated dose to each piece of equipment. Therefore, th e worst point dose from each of the major sources in the drywell and wetwell was calculated, and then summed for a conservative estimate of the total integrated dose. The dose sum of the worst-case source contributors in the drywell is 7.6 x 10 7 rads, but 7.4 x 10 7 rads is used as the worst-case dose for the equipment qualification program. All of the worst-case contribut ors cannot be present for a particular accident. Thus, the largest worst-case dose is calculated for the depressurized reactor coolant system. The worst-case dose is applied to safety-related equipment with an elevation within 5 ft of core midplane. Safety-related equipment in th e drywell outside this elevation span is assi gned a dose of 7.0 x 10 7 rads. In the wetwell, the maximum gamma dose above the suppression pool is 9.5 x 10 7 rads (see Section J.F.3 for discussion on photon energy and anticipated dose reduction of the above results).

These results include the contributions from all major gamma sources within primary containment during normal operation as well as the 6-month period contribution fo llowing a postulated LOCA. Tables J.F-1 and J.F-2 give a breakdown of the integrated dose contribution from each of the

• Environmental qualification (EQ) of safety-related mechanical (SRM) equipment has been eliminated from the overall CGS EQ program.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.6-2 major gamma sources to the drywell and the wetwell. The 40-year in tegrated gamma doses due to normal operation ar e taken from Reference J.7-20. This methodology for determining a worst-case dose for equipment in the drywell is not valid for the region inside the sacrificial shield wall or under the react or pressure vessel. A point-specific radiation dose calculati on is required for all components present in either of these two regions.

Specific calculations have been performed for equipment that was evaluated individually for total integrated dose. Results of these calculations are summarized in Reference J.7-26.

In accordance with Section 1.4(8) of Reference J.7-2 , only the gamma dose need be considered for "shielded components." Since beta radiation is so readily atte nuated, virtually any enclosure of sensitive components will be sufficient to classify the component as "shielded." A review of all safety-related equipment located inside primary containment determined that most C1E* equipment is sufficiently shielded against beta radi ation. Thus, the beta dose contribution is excluded from the total integr ated radiation doses compiled for equipment qualification purposes unless a beta-sensitive component is not ad equately protected from the airborne beta environment. Wh en required to include beta dose contributi ons, a finite source volume is used. The source volum e is a sphere surrounded by a sh ell of sufficient thickness to stop all beta particles from ente ring the source volume. This fi nite spherical source volume is conservative for any generalized source volume shape (the dose at the center of the sphere is higher than the dose at any poin t of any generalized source sh ape of equal total volume). A discussion of the results is presented in Attachment J.G.

J.6.2 SECONDARY CONTAINMENT RADIATION RESULTS

The integrated direct gamma dose (40 years and 6 months LOCA -

direct gamma, gamma shine, and airborne gamma) was evaluated for the worst target of all C1E

• equipment in each zone and is used for qualifi cation of all the other C1E

• equipment in that zone. The 40-year integrated gamma doses (Figures J.6-1 through J.6-10) are taken from References J.7-20 and J.7-21. The direct gamma dose contribution outside primary containment due to sources inside the primary containment was investigated. Safety-related equipment located in the direct shine path through the penetrations was evaluated in Reference J.7-23. All post-LOCA radiation dose contributions to safety-related equipment from streaming through the bioshield wall penetrations were included in the radiation doses. Evaluation of bioshield wall penetrations

identified radiation dose problems associated wi th some of those penetrations (Reference J.7-24). The post-LOCA evaluation of safety-related equipment assumed the C1E

• equipment was shielded for 40-year normal opera tions. To adequately protect C1E

• equipment a concrete wall was designed and installed for penetrations X-100A, X-105A, and X-100B. The

• Environmental qualification (EQ) of safety-related mechanical (SRM) equipment has been eliminated from the overall CGS EQ program.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-000 J.6-3 remaining penetrations evaluated (Reference J.7-25) were surveyed during plant startup to confirm radiation analysis calculations.

Airborne beta doses outside containment were evaluated in accordance with the methodology described in Section J.5.5 and Attachment J.D. The beta dose contri bution is excluded from the total integrated radiation doses compiled fo r equipment qualificati on purposes unless a beta sensitive component is not adequately protect ed from the airborne beta environment.

J.6.3 RADIATION RESULTS IN THE VITAL AREAS AND ACCESS ROUTES

Figures J.6-8 through J.6-16 present the vital areas and ac cess routes located outside the reactor building. Figures J.6-17 and J.6-18 present the vital areas and access routes located inside the reactor building. The doses indicated on each figure are also the 6-month LOCA integrated gamma doses to be used for C1E

• (safety-related) equipment qualification purposes.

Table J.6-1 also presents a summary of the 6-m onth LOCA integrated gamma doses on all C1E* equipment located in vital areas.

Figures J.6-17 and J.6-18 show the access route in the reactor building for operation of SW-V-75AA and SW-V-75BB, the manual isolation valves for the service water to fuel pool cooling makeup water supply.

Radiation levels of vital areas and access routes were determined at selected locations outside the reactor building due to radioactive sources inside the reactor building and release of radiation activity from the reactor building elev ated vent. The vital areas and access routes analyzed are consistent with those discus sed in NUREG-0737, Item II.B.2 (Reference J.7-5). The radiation levels determined for the vital areas and access routes identified in Figures J.6-8 through J.6-18 are summarized in Table J.6-1. All of the vital areas and access routes have radiation levels less than the guidelines presented in NUREG-0737.

The total dose received at a vital area during a post-LOCA scenario is obtained by summing the exposure dose enroute to the v ital area and the radiation dose at the vital area.

These doses are listed in Table J.6-2.

The analysis completed for vital areas and access routes assumed that except for the reactor building railroad bay and on the west side of the 522-ft el. there would be no access to equipment or areas located with in the reactor building during the post-LOCA scenario. The exceptions are shown in Figures J.6-17 and J.6-18 and Table J.6-2. Access to the reactor building railroad bay for 3 hr is allowed to provide the ability to fill or exchange N 2 bottles. The entry to the west side of the 522-ft el. is to allow SW-V

-75AA and/or SW-V-75BB to be opened (see Section 9.1.3.2.3). These valves are readily accessible and the entire opening

• Environmental qualification (EQ) of safety-related mechanical (SRM) equipment has been eliminated from the overall EQ program.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December2005 LDCN-05-005 J.6-4 evolution for one of these valv es would take 2.17 mi nutes and could be pe rformed at 9.7 hr with the resulting exposure of 3.8 rem. Under worst-case conditions, at least one of these valves would need to be opened by 10 hr. Once a manual valve is opened, the spent fuel pool level can be controlled with the motor-operated valve from the main control room.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.6-5 Table J.6-1 Six-Month Total Integrat ed Dose (Loss-of-Coolant Accident) to Areas Containing C1E Equipment Outside the Reactor Building

Vital Area Description Radiation Level a Direct Gamma Shine +Airborne Gamma (rads)

Control room (el. 501 ft) 0.21 Technical support center 0.21 Sale area (el. 487 ft)

6.5 Nitrogen

supply to ADS accumulators (el. 437 ft)

3.9 Standby

service water pump valves

1.7 Remote

shutdown room (el. 467 ft)

3.9 Switchgear

room 1 (el. 467 ft)

3.9 Switchgear

room 2 (el. 467 ft)

3.9 Radwaste

control ro om (el. 467 ft)

3.9 Battery

racks, dc battery chargers, two motor control

centers (MCCs) (el. 467 ft)

3.9 Three

MCCs and three sw itchgears (el. 437 ft)

3.9 Direct

current battery charge r and rack (el. 437 ft)

3.9 Diesel

oil tanks (el. 437 ft)

3.9 Solid

radwaste contro l panel and decontamin ation station control panel (el. 437 ft) 3.9 a Volume correction factors for a semi-infinit e cloud were applied to the control room and technical support center. If the vol ume correction factors were to be applied to all areas, the integrated dose would be redu ced by a minimum of fivefold.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-005 J.6-6 Table J.6-2 Vital Areas and Access Route List of Radiation Exposure to Personnel During the Required Post-Loss-of-Coolant A ccident Operations Radiation Exposure Vital Area Description Gamma Whole Body (rem) Thyroids (rem)a) Beta Skin (rads) Control room (el. 501 ft) b Technical support center b 0.21 0.21 0.21 c 0.21 c 0.95 0.95 Security center b Auxiliary security center b 3.1 1.7 13.4 d 4.8 2.7 Sample analysis area (EOC) b 0.0013 - -

Standby service water pump valves (cooling ponds) e 0.3 0.94 d 0.46 All infrequently occupied vital areas inside the radwaste and diesel generator buildings b 0.13 f 1.6 d 0.48 Sampler for elevated release duct (roof turbine

building)e 2.5 8.0 d 3.8 Reactor building railroad bay (N 2 bottles)g 0.4 - - Reactor building 522-ft el. (SW-V-75AA

and/or SW-V-75BB h 3.8 - -

Postaccident sample area (el. 487 ft) e 0.36 3.2 d 0.96 Access Routes All access routes inside the radwaste and

diesel generator buildings e 0.13 f 1.6 d 0.48 All access routes i outside the radwaste and diesel generator buildings e 0.53 1.6 d 0.8 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-000 J.6-7 Table J.6-2 Vital Areas and Access Route List of Radiation Exposure to Personnel During the Required Post-Loss-of-Coolant Accident Operations (Continued) a If self-contained respiratory equipment (SCBA) is used, the thyroid dose will essentially equal the whole-body dose.

b Area of continuous occupancy.

c Assumes self-contained respiratory equipment was used by pe rsonnel during 0-3 hr post-LOCA situation.

d No respiratory equipment was assumed.

e Area occupied 0.5 hr at times after 1 hr into the LOCA.

f A volume correction factor for the semi-inf inite cloud was included in the calculation.

g Assumes entry after 12 days pos t-LOCA for 3-hr occupancy with respiratory equipment for railroad bay portion of reactor building only. h Assumes entry after 9.0 hr post-LOCA for a 2.17-minute evolution to open SW-V-75AA or SW-V-75BB with respiratory equipment and in full PC gear following access routes shown in Figures J.6-17 and J.6-18. The 2.17 minutes consists of a 1.

83-minute transit (1.5 minutes in 522K and 0.33 minutes in 522H) and a 0.33-minute occupancy time in 522H.

i Extremely conservative anal ysis since the plume of ai rborne radioactivity cannot simultaneously cover all access routes.

Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December2007 LDCN-06-000 J.7-1 J.7 REFERENCES

J.7-1 NUREG-0578, "TMI Lessons Lear ned Task Force Status Report and Short-Term Recommendations."

J.7-2 NUREG-0588, Rev. 1, "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."

J.7-3 NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident."

J.7-4 Clarification letter to NUREG-0578, September 5, 1980.

J.7-5 NUREG-0737, "Clarification of TMI Action Plant Requirements."

J.7-6 IE Bulletin No.79-01B, "Environmental Qualification of Cl ass 1E Equipment."

J.7-7 Supplement No. 2 to IE Bulle tin 79-01B, September 30, 1980.

J.7-8 Oak Ridge National Laboratory, "ORIGEN2, Isotope Generation a nd Depletion Code - Matrix Exponential Me thod," ORNL Report No. CCC-371.

J.7-9 J. F. Perkins, U.S. Army Missile Command, Redstone Arsenal, Alabama, Report No. RR-TR-63-11 (July, 1963).

J.7-10 Oak Ridge National Laboratory, "Modifications of the Point-Kernel QAD-P5A," ORNL-4181, July 1968.

J.7-11 Oak Ridge National Laboratory, "ORIGEN-79, Isotope Generation and Depletion Code - Matrix Exponentia l Method," ORNL Report No. CCC-217.

J.7-12 EDS Report No. 01-0740-1138, Revision 0, "Source Term Report,"

December 1980.

J.7-13 Columbia Generating Station FSAR, Section 15.6.5.5.1.1, Amendment 27, November 1982.

J.7-14 Regulatory Guide 1.3 "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant A ccident for Boiling Water Reactors," Revisi on 2, June 1974.

J.7-15 Columbia Generati ng Station FSAR, Secti ons 6.7.2 and 6.7.3.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November1998 J.7-2 J.7-16 Standard Review Plan , Section 15.6.5, "Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary."

J.7-17 Oak Ridge National Laboratory, "QAD-CG, A Combinitorial Geometry Version of QAD-P5A, A Point-Kernal Code," ORNL Report No. CCC-307.

J.7-18 Byoun, Yoon and Babe l, Paul J., "Descripti on of Computer Program (SCAP-BR)," User's Manual.

J.7-19 Impell Calculation

1. 0740-004-019, "Fluid Cont act Dose Calculation."

J.7-20 Letter BRWP-RO-81-288, dated July 29, 1981, "For ty-Year Integrated Dose for Radiation Zones in the Reactor Building."

J.7-21 Letter BRWP-RO-81-181, dated September 29, 1981, "Forty-Year Integrated Dose for Radwaste and Turbine Buildings."

J.7-22 Burns and Roe calculation 5.01.78.

J.7-23 Burns and Roe calculati on 5.01.030, "Total Accide nt Integrated Dose to Equipment in the Reactor Building From Primary Containment Penetration Streaming."

J.7-24 Letter BWRP-RO-82-186, dated June 18, 1982, "Bioshield Wall Penetrations with Unresolved Shielding Concerns."

J.7-25 Burns and Roe Techni cal Memorandum, TM-1295, "B ioshield Wall Penetration Shielding Requirements," dated April 6, 1983.

J.7-26 Calculation 1140-006-CLIST, "C1E/SRM Containment List."

J.7-27 Letter EDSWP-81-015, dated February 18, 19 81, "SGTS Filter Modeling Assumption."

J.7-28 Murphy and Campe "Nuclear Power Pl ants Control Room Ventilation System Design," 13th AEC Air Cleaning Conference.

J.7-29 Letter EDSWP-80-031, "Shielding Design Input Data."

J.7-30 Theodore Rockwell III, "Reactor Sh ielding Design Manua l," U.S. AEC, TID-7004, March 1956.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.7-3 J.7-31 David A. Slade "Meteorology and Atomic Energy," U.S. AEC, July 1968.

J.7-32 Record of Convers ation, D. A. Wert to J. A. Ogawa, dated September 22, 1980, "Shielding Source Term."

J.7-33 Letter BREDS-RO-81-17 dated December 11, 1981, A. N. Kugler to E. A. Daugherty.

J.7-34 NUREG/CR-009, Technical Basis for Models of Spray Washout of Airborne Contaminants in Containmen t Vessels, October 1978.

J.7-35 Columbia Generat i ng Station FSAR, Amendment 11, Table 6.2-12.

J.7-36 Columbia Generating S t ation FSAR, Table 6.2-1.

J.7-37 Not used

J.7-38 EDS Nuclear Inc., Calcul ation 0740-004-00 4, Rev. 0.

J.7-39 EDS Nuclear Inc., Calcul ation 0740-004-00 6, Rev. 0.

J.7-40 Columbia Generating S t ation FSAR, Section 15.6.5.

J.7-41 Standard Review Pl an, Section 6.5.3, "Fissi on Product Control Systems," Nuclear Regulatory Commission, June 1975.

J.7-42 Columbia Generat i ng Station FSAR, Secti on 6.2 , Table 6.2-2.

J.7-43 Columbia Generat i ng Station FSAR, Secti on 6.2 , Table 6.2-11.

J.7-44 WPBR-R0081-143, Holmberg to Forrest, "Plant Shielding Analysis - WNP-2,"

August 8, 1981.

J.7-45 Burns and Roe calculation 5.01.59, "P rimary Containment 6 Month Accident Source Terms ORIGEN2."

J.7-46 Burns and Roe calculation 5.01.60, "Six Month Accident Integrated Dose in Primary Containment - Depressurized Case - QAD-CG."

J.7-47 Burns and Roe calculation 5.01.62, "Dose Rate for 26" Main Steam Lines Inside Primary Containment."

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.7-4 J.7-48 Burns and Roe calculation 5.01.63, "RWCU Piping Sources in Containment (Normal Operations)."

J.7-49 Burns and Roe calculation 5.01.

64, "RWCU, RRC, and RHR Lines in Containment (Shielding - Normal Operation)."

J.7-50 Burns and Roe calculation 5.01.69, "Total Integrated Dose, Accident &

Operation, from Systems and Primary Containment Atmosphere to Safety Related Equipment Located EDS - Pressurized Case."

J.7-51 Burns and Roe calculation 5.01.74, "Six Month Accident Integrated Dose in Primary Containment -Depressurized Case - QAD."

J.7-52 Burns and Roe calculation 5.01.75, "Accident Integrated Dose for Systems -

Depressurized Case."

J.7-53 Burns and Roe calculation 5.01.76, "Total Integrated Dose, Accident & Operation, to Safety-Related Equipment in Containment Located by EDS Due to Systems."

J.7-54 Burns and Roe calculation 5.01.77, "Total Integrated Dose, Accident &

Operation, from Systems and Primary Containment Atmosphere to Safety-Related Equipment in Containment Located by EDS - Depressurized Case."

J.7-55 Burns and Roe calculation 5.01.20 , "WPPSS Sacrificial and Bioshield Calculations."

J.7-56 Burns and Roe calculation 5.45.07, "Post-LOCA MSIV Leakage Calculation."

J.7-57 Calculation 0740-004-VLIST, "Reactor Building SRM Active Valve, Pump, Fan, and Turbine List."

J.7-58 0740-073-001 Parameter Penetration Study.

J.7-59 Ianni, P. W., "Effect iveness of Core Standby Cooling Systems for General Electric Boiling Water React ors," APED-5458 , March 1968.

J.7-60 "Design and Performance of General Electric Boiling Water Reactor Main Steam Line Isolation Valv es," APED-5750, General Electric Company, Atomic Power Equipment Depa rtment, March 1969.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.7-5 J.7-61 "Power Uprate with Extended Load Line Limit Safety Analysis for WNP-2,"

NEDC-32141P, General Electric Company.

J.7-62 "General Evaluations of General Electric Boiling Wa ter Reactor Power Uprate -

Volume 1," NEDC-31984P, Ge neral Electric Company.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.A-1 Attachment J.A UNISOLATED LEAKING BUILDING PATH REPORT

A basic assumption to the plant shielding analysis is that the reactor isolates such that there is no radiation leakage path to the outside. A le akage path investigati on was done verifying the above assumption. While performing this investigation, th e total number of lines (69) penetrating the RB boundary, the associated sy stem components and interface systems were reviewed.

The assumption eliminating the consideration of leakage is consistent with NUREG-0737, Clarification 2. This investigation assumed that containment isolation occurred prior to the egress of highly radioactive flui ds. Additionally, it assumed that all safety-related equipment was available, and that all safety systems were pressurized.

Therefore, at any interface, such as a heat exchanger, no potentia l leakage was considered if th e nonradioactive sy stem was at a higher pressure than the radioactive system. This investigat ion has not considered leakage from equipment seals, closed valves, or pipe rupture, except in the evaluation of the equipment and floor drain systems. The systems considered are tabulated by drawing number in Table J.A-1.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.A-3 Table J.A-1 System Flow Diagrams Employed

to Perform The Review

Drawing Number Revision D r awing Number Revision M501 10 M536 12 M502 17 M537 25

M503 5 M538 9 M504 25 M539 28 M505 14 M540 15

M506 23 M541 13

M507 27 M542 4 M508 25 M543 17

M509 10 M544 10

M510 30 M545 15

M511 15 M546 10

M512 8 M547 9 M513 33 M548 14

M514 13B M549 14A

M515 17C M550 9 M516 20 M551 8 M517 25 M552 12

M518 14 M553 10

M519 18 M554 11

M520 15 M555 7 M521 20 M556 10

M522 6 M557 4 M523 29 M607 Sheet 1 7

M524 19 M607 Sheet 2 5

M525 19 M607 Sheet 3 3

M526 25 M527 18 M528 15 M529 21 M530 18 M531 24 M532 20 M533 Sheet 1 1

M533 Sheet 2 1

M533 Sheet 3 1

M534 16 M535 Sheet 1 26 M535 Sheet 2 21 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-1 Attachment J.B SOURCE TERM DEVELOPME N T AND PARAMETRIC STUDIES

FOR SECONDARY CONTAINMENT

The major tools used in the development of source terms and parame tric studies inside secondary containment were the ORIGEN and QAD-P 5A computer codes.

Descriptions of the codes are in References J.7-11 and J.7-10. ORIGEN was used to compute the activities and energies of fission products released from th e reactor core. The output of ORIGEN [the time-dependent energies and ac tivity of radioactive fission products following loss-of-coolant accident (LOCA)] was used as input to QAD-P5A to calcula te the airborne, shine, and direct doses for standard geometrics as well as the basis of dir ect dose parametric studies.

J.B.1 RADIOACTIVE SOURCE TERMS IN SECONDARY CONTAINMENT

The ORIGEN computer code (Reference J.7-11) was used to calculate the radioactive source terms inside secondary containment for liquid-containing and gas-containing systems. The fission products at the end of fuel life were assumed to be available for release immediately following the accident. The concentrations of noble gases, haloge ns, and other fission products released to the gaseous and liquid sources were co mputed. Subsequent fission product decay and daughter pr oduct generation were then calculated for 20 time periods, covering a total period of 1 year.

The assumptions used in determin ing the initial distribution and l eakage of radioactivity in the primary containment air and liquid space are as follows:

a. 100% of the noble gases and 50%

of the halogens are distributed homogeneously within the primary containment free volume immediately following the postulated accident;

b. 50% of the halogens and 1% of the re maining fission products in the core are mixed instantaneously and homogeneously with the primary containment liquid

space. The primary containment liquid space is defined as the sum of the suppression pool liquid and the reactor coolant system (RCS) liquid; and

c. The fission products available for rele ase are defined as the total inventory generated in the equilibrium core after 1000 days at reactor power of 3556 MWt.

Assumptions a and b are NRC recommended assu mptions for defining radioactivity release fractions for the qualification of safety-related equipment (Reference J.7-2) and are detailed in References J.7-32 and J.7-34.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-2 Assumption c represents the maximum burnup level in the core and the fi ssion products at the end of fuel life prior to radioac tivity release and is conservative.

Table J.B-1 shows the gamma activity concentrati on at selected time periods for the liquid-containing system. The results of Table J.B-1 were used as input in the dose parametric study. Due to rapid decay of the high-ener gy isotopes, the averag e gamma energy for the gas-containing system varies fr om 0.8 MeV at the beginning of the accident to 0.3 MeV at 1 year after the accident.

J.B.2 AIRBORNE DOSE IN SECONDARY CONTAINMENT

The time-dependent post-LOCA acti vity levels as calculated by the ORIGEN computer code were used as input in the calculation of the ai rborne beta and gamma dos e rates and integrated doses inside the cubicles in the secondary contai nment. The assumptions used in this analysis are as follows:

a. Activity that leak s into the secondary containmen t is homogeneously mixed with the secondary containment atmosphere prior to its removal from the atmosphere

through the standby gas treatment system (S GTS). This is c onsistent with the NRC-recommended assumptions used for calculation of doses inside primary containment (Reference J.7-2 and J.7-34);

b. An SGTS flow rate of 2430 scfm was assumed to be the flow rate of the effluent air. This is the designed minimum accident flow rate (Reference J.7-35) based on one reactor build ing airchange per day;

c. Air that leaks out of the primary contai nment flows directly and totally into the secondary containment. Bypass leakage is not considered. This is conservative when considering dose in the secondary containment, since it maximizes the buildup of radioactivity in the secondary containment;

d. Geometric factors are used to convert the semi-infinite cloud gamma dose to a finite gamma dose. This assu mption is used in Reference J.7-28 , and is based on an average gamma ray energy of 0.733 MeV. The effect of time dependence of average gamma ray energies has been proven to be negligible; and

e. Primary containment activity leakage rate is 0.5%/day. This is consistent with the assumptions established in Reference J.7-29.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-3 A model of the primary and secondary contain m ent atmosphere is shown in Figure J.B-

1. The activity concentration of a certain isotope i n side the contai n m ent is changing due to the following three mec h anisms: a. Transport of activ i t y due to air leakage,

b. Depletion of activ i t y due to radioact i v e decay a n d plateout of e l emental halogens inside pri m ary containmen t , and c. Increases in activity levels due to daughter p r oduct g e neration from fission product decay.

According t o References J.7-2 and J.7-34 , plateout may be model e d by an expotential removal process:

At A t p () ()e x p ( )0 Where p is the removal constant due to plateout.

The fir s t s t ep in this ca l culation is t o m odel the decay and transport of the airborne radionuclides.

General airborne activity balance in containment:

d dt CV Q C CV CV C V li l l i l i pi li j j lj i ()11 1 1 le a ka g e d e c a y plat e out g r o w th (J.B-1) where 1i C = concentration of isotope "i" 1 Q = leakage rate from primary containment 1 V = volume of primary containment i = radioactive decay constant of isotope "i" pi = plateout removal c onstant of isotope "i" The term j j1i1 CV reflects the growth of a given nuclide as the result of decay of parent nuclides.

The original release of nuclides consists only of ha logens and noble gase

s. Since fission products are neutron-rich, decay of fission products proceeds toward higher atomic numbers.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-4 In this manner, halogens will decay into noble gases, and t h en to higher a t omic-numbered

ele m ents. Since the decay chain reaches a s t a b le isotope af t e r only a few decays, it can be seen that upon release of these airborne nucli d es, the halogens have no significant airbo r ne parent nuclides. This term may be neglected in the case of halogens.

Case 1 - Containment Halogens

Elemental iodine undergo plateout (Reference J.7-34) so equation (J.B-1) becomes:

d dt (C V)= -Q C - CV - CV 1i 1 1 1i i 1i 1 pi 1i 1 (J.B-2) Solving (J.B-2) with the initial condition;

at t = 0,

1i 1i C = C (0), 1i 1i 1 1ipi C (t) = C (0) e x p (-Q V + + t) (J.B-3) Particulate and o r ganic iodine are a s sumed unaffec t ed by plateout (Re f erence J.7-34).

Equations (J.B-3) for particula t e and org a nic iodine may then be shown to be 1i 1i 1 1 i C (t) = C (0) e x p (Q V + t) (J.B-4) One can note, at this po i nt, that all three iodine species have factors of

1i i C (0) exp ( - t) in the equations. This term may be defined as

i1 i i 1 S (t) = C (0) e x p ( - t) V (J.B-5) i S(t) is seen to be the total ac tivity released into the system as a result of decay. S i(t) is independent of the transport of the nuclides. The following definitions will be made.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-5 e f = fraction of total iodine that are elemental p f = fraction of total iodine that are particulate o f = fraction of total iodine that are organic Equations (J.B-3) and (J.B-4) can be combined to get 1iH e p op 1 1 iH 1 C (t) = f exp (-t) + (f + f) e Q t/V S(t)V xp (J.B-6) where

1iH C(t) = total iodine concentratio n in primary containment iH S(t) = total iodine activity p = plateout constant for elemental iodine

At this point, Reference J.7-2 allows only a factor of 200 reduction for elemental iodine plateout effects.

So when exp(-t) =

1 200 , pp th e n becomes zero.

(J.B-7) Defining:

p p t = Ln (200) Equation (J.B-6) may be rewritten as 1iH iH 1 1 1 C(t) = S(t)V e -Q t/V xp H f (t) (J.B-8) Where f H(t) is defined as

(a) He ppop f (t) = f exp (-t) +

f + f t t (J.B-9)

(b) Hepop f (t) = (f/200) + f + f t t

Case 2 - Containment Noble Gases

Noble gases do not undergo plateout. Daughter products are al so conservatively assumed to act as noble gases.

Equation (J.B-1) for noble gases becomes C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-6 d dt (CV)= -Q C - CV + CV1i1 1 1i i1i1 j j1j1 (J.B-10) Integrating equati on (J.B-10) gives 1i 1 1 i i C(t) = e -Q t/V exp- t (B +

f (t))xp (J.B-11) where i j j ij f (t) = C e/xpQV 11 + i d t (J.B-12) and B is a constant to be determined.

All daughter products of plated-out iodine are conservatively assu med to be re-released into the containment atmosphere as if th e iodine were airborne. For th e first isotope in a series (no parent nuclide), j = 0 and f o (t) = 0.

Since C 1j(t) has the same form as C 1i (t), equation (J.B-12) becomes i j j j( - )t f (t) = (B +

f (t)) e ij dt (J.B-13)

Equation (J.B-13) shows th at the only dependence on Q 1/V 1 is that carried over from the parent isotope is f n(t). Since f o (t) is independent of Q 1/V 1 , f i (t) is independent of Q 1/V 1. Equation (J.B-11) can thus be rewritten as 1i-( / ) t i1 C (t) = e Q V S (t)/V 1 1 (J.B-14) where i S (t) = e (B +

i f (t))1 V xp i t (J.B-15) It can be seen that S i (t) is the solution to equation (J.B-10) without the leakage term. S i (t) is the activity for the total inventory of nuclides releas ed from the reactor core. S i (t) values are determined by the use of ORIGEN. S i(t) includes radioactive decay and daughter product growth.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-7 For a general airborne activity balance in the reactor building (secondary containment):

d dt (CV) = + Q C - Q C - CV + CV2i2 i 1i 2 2i i2i2 j j2j2 leakage leakage decay growth in out (J.B-16) where C 2i = concentration of isotope "i" in the reactor building Q 2 = leakage rate from reactor building V 2 = volume of the reactor building

Plateout inside secondary contai nment is conservatively neglected.

Case 3 - Iodine Inside the Reactor Building

As in Case 1, the growth term of equation (J.B-16) is neglig ible. Equation (J.B-16) can be integrated to give

2i-(/ + )t 1 2 1i C (t) = e Q VB + Q V e C(t) dt 2 2 i Q V(/ + )t 2 2 i (J.B-17) From equation (J.B-8), C 1i (t) is substituted into (J.B-17) 2i(/ + )t 1 2 C (t) = Be Q V Q V e 2 2 i Q V e xp (Q / V + )t-(/ + )t 2 2 i 2 2 i (S (t)V e -1 Q t/1 V f(t)) dt iH 1 H (J.B-18)

Substituting equation (J.B-5) in to (J.B-18) results in 2i 1 2 1i C (t) = Bexp -(2 Q/2 V + i)t + Q V C (0) e-(2 Q/2 V + i)t xp (J.B-19) exp (Q/V - Q/V)t ()2 2 1 1ftdt H f H (t) is a complex function of time (equation J.B-9). C 2i(t) must be solved in a series of solutions to equation (J.B-19).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-8 For simplification, the f o llowing factors are defined x = Q V - Q V 2 2 1 1 y = x - p Equation (J.B-19) becomes

2i 1 21 iH ()C (t) = B e x p - (2 Q / 2 V + i)t + Q V V S (t) e -( 2 Q / 2 V + i)t xp ef d t xt H t (J.B-20) Integrating (J.B-20) for 0 t t p with the initial condition; C 2i (0) = 0 gives (For S (t) = S (0) e - t ): iH iH i 2i 1 12 iH 1 1 e p po C (t) = Q V V S (t) (e -Q t/V (f y e -t + f + f x)xp xp (J.B-21) e-Q t/V (f y + f + f x))2 2 e po Defining 1 e po K = (f y + f + f x); (J.B-21) becomes (for 0 t t p): (J.B-22) 2i 1 12 iH e p po 1 C(t) =Q VV S(t) f y e -t +f + f x e +K e xpxp-Q t/Vxp-Q t/V 1 1 2 2 And (for t ³ t p): (J.B-23) 2i 2 2 i 1 12 iH 2 2 xt e po C(t) = Bexp -(Q/V +)t + Q VV S(t)e -Q/V t e (f200 + f + f) dt xp C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-9 Solving (J.B-23) gives (J.B-24) 2i 2 2 i 1 12 iH 2 2eop xt C(t) = Bexp (Q/V +)t + Q VV S(t)e -Q/V t (f/200 + f + f x)e xp

At t = tp (from J.B-22) ):

(J.B-25) 2ip 1 12iHp 1 2 2p e op C (t) = Q V V S (t) Ke -(Q/V)t+ (f y e + (f + f x))e xp -t xp-Q t/V p p 1 p1

By definition of t p [eq. (J.B-7)]: exp -t =1 200 p p Combining (J.B-24) a nd (J.B-25) at t =

t p gives (J.B-26) Bexp -(Q/V +)t= Q VV S (t) e -(Q/V)t (K + f200 (1 y - 1 x) e p)2 2 i p 1 12iHp 2 2p 1 e xp xt and B Q VV SK iH1 12 2 0 () (J.B-27) where 21 e K = K + f 200 (1 y - 1 x) e p xt (J.B-28) So (J.B-24) becomes (for t p t) 2i 1 12 iH 2-t/eop-t/C(t) = Q VV S(t) (K e Q V + (f/200 + f + f x)e Q 2 2 1 1 (J.B-29) Equations (J.B-22) and (J.B-29) may be combined to form a general solution as follows:

2iiH 2H 2 C (t) = S (t) F (t) / V (J.B-30)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-10 where (for 0 t t p) 2H 1 1 1 2 2 e po 1 1 F (t) = Q V (K e -Q t/ V + (f y e + f + f x)e -Q t/ V xp xp - t xp ) p for (t t p) (J.B-31) 2H 1 1 2 2 2eop 1 1 F(t) = Q V (K e-Q t/V + (f/200 + f + f x)e-Q t/V)xp xp Case 4: Noble Gases I n side the Reactor Building

Equation (J.B-16) for noble gases m a y be rewritten as

d dt (C) - (Q/V) C - (Q/V + ) C + C 2i 121i 2 2 i 2i j j 2j (J.B-32)

Integrating (J.B-32) gives (J.B-33) 2i 2 2 i 1 2 1i j j 2j C (t) e (Q/V +)tB +e (Q V C(t)+ C(t)) dt xpxp(Q/V + )t 2 2 i C 1i (t) is found from equation (J.B-14) to be 1i 1 1 i1 C (t) = e -(Q/V)t S (t)/V xp S i (t) cannot be found analytically; hence equation (J.B-33) cannot be found analytically through this method. However, in deri ving equation (J.B-14), it was show n that if all parent nuclides are transported identic ally, then the so lution of equations cons isting of transport and radioactive decay can be separated. Since the halogens are not transpor ted in the same manner as noble gases, this is not strict ly true. However, the assumpti on of daughter growth as if the halogens were transported will be conservative, due to the nonconsideration of the physical holdup in primary to secondary l eakage of daughters of halogens.

Equation (J.B-14) may be rewritten as

11i iN 1N VC (t) = S (t) F (t) (J.B-34)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-11 where 1N 1 1 F (t) = e-Q t/V xp (J.B-35) S iN (t) is the noble gas total ac tivity term, as before. F 1N(t) is the fraction of that activity remaining in primary containment.

Equation (J.B-16) may be modified to show the fractions of activ ity, rather than total isotopic activity, in secondary containment to give

d dt (F) = Q V F - Q V F 2N 1 1 1N 2 2 2N (J.B-36) Integrating equation (J.B-

36) with initial conditions:

at t=0, F 2N = 0; gives 2N 1 1 1 1 2 2 F(t) = Q V x (e -Q t/V -e -Q t/V )xp xp (J.B-37)

C 2i(t) is then found from C 2i (t) = S i N (t) F 2 N (t)/V 2 (J.B-38) p is found in Reference J.7-2 to be determined p = K g A 1/V 1 (J.B-39)

K g is conservatively assumed to be equal to 0.05 cm/sec (Reference J.7-34).

A 1 is the surface area inside the drywell = 3.2x10 7 cm 2 (Reference J.7-33).

V 1 =5.68x10 9 cm 3 (Reference J.7-36) p = 1.01 hr

-1 To calculate the airborne gamma dose rate inside the secondary containment, the method as described in Reference J.7-28 is used:

D = 0.25 E i (C noble gas +

C halogen)i=1 n2i2i (J.B-40)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-12 D = D GF (J.B-41) GF = 1173 V 0.338 (J.B-42) where D = semi-infinite gamma cl o ud dose rate (rads/sec)

E i = average gamma energy of the isotope (MeV)

C 2i = activity concentration inside secondary contain m ent (Ci/m 3)

GF = geometric factor used to scale the semi-infinite gamma cloud dose to a finite cloud dose

V = volume of the finite cloud (ft

3) By taking S i (t) from ORIGEN output and using equations (J.B-31) a nd (J.B-37) to calculate F 2 (t), the total gamma dose in secondary containment can be computed by using equations (J.B-40) through (J.B-42).

The airborne semi-infinite cloud gamma dose rates are shown in Figure J.B-2. As can be observed from the figures, the gamma doses inside secondary containment reach their peaks at around three days after the accide nt, and decay slowly thereafter due to the depletion of radioactivity by radioactive decay and removal through the SGTS.

The geometric factor in equation (J.B-42) is devel oped in Reference J.7-28 for average gamma energies of 0.733 MeV. There has been a concern that this geometric factor may vary appreciably with time due to the faster decay rate of the hi gh energy isotopes. The average gamma energy during various tim e periods following the accident were computed and the results show that the average gamma energy varies from 0.3 Me V to 0.8 MeV. As discussed in Reference J.7-31 , the geometric factor changes by less th an 5% within that energy range. It is therefore concluded that the change in the geometric factors with time is negligible, and that equation (J.B-42) can be used to calculate the finite cloud gamma dose inside the secondary containment.

J.B.3 PARAMETRIC STUDIES FOR DIRECT PIPING DOSE

The purpose of the parametric study was to iden tify the parameters which have a significant affect on the radiation dose rates. The com puter code QAD-P5A was used to develop a C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-13 correlation scheme for the signi ficant parameters such that a simplified procedure for calculating radiation dose rates for complex source and receptor ge ometries can be developed. The dose rate at a target dist ance of 8 ft radia lly outwards from the centerline of an 8-in. schedule 40 pipe, infinitely long (standard pipe) was first calculated and defined as the standard dose rate. A parametr ic study was then performed to investigate the effects of the variation of parame ters such as pipe lengt h, pipe diameter, shield thickness, and target locations on the dose rate. The resu lts of this parametric study were then correlated as a set of correction factors to the standard dose rate.

A simplified procedure was developed to calculate the dose rates and cumulate doses for the multitude of source-target conf igurations by using these correction factors.

J.B.3.1 Functional De pendence of Various Parameters on Secondary Containment Dose Rates

The gamma ray energy flux from a line source "S L" to a detector point "P" (see F igure J.B-

3) is shown in Reference J.7-30 as = L BS 4r exp -1 b Sec exp -

1 b Sec d 0 12d - (J.B-43) where

= uncollided gamma ray flux (photons/c m 2 - sec) b 1 = total attent u a tion through shield S L = source strength of line source (pho t ons/cm sec) B = buildup factor = angle subtended by the l e ngth of the line source (see Figure J.B-3)

The source strength "S L" is a function of the volume of li quid inside the pipe segments, which is also a function of the diameter and volume of the pipe. The angles " 1" and "2" are also functions of "a/r" and "b/r," respectively (see Figure J.B-18 for definition of "a/r" and "b/r" respectively). Therefore, th e functional dependen ce of gamma ray dose rates on the various parameters can be represented by the following equation:

= FFF (a/r, b) + F (b/r, b) oDRL1L1 (JB-44) where o = base gamma ray f l ux for standard pipe F D = pipe diame t er correcti o n factor F R = radial dis t ance correcti o n factor F L = (a/r, b 1) = pipe length correction factor

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-14 J.B.3.2 Parametric Study Procedures The procedure for performing this parame tric study is documented as follows:

a. Calculate the dose rate at a target distan ce of 8 ft from the cen terline of an 8-in.

schedule 40 pipe infinite ly long (standard pipe);

b. Perform parametric studies on the variation of dose rates with

1. Radial distance from the pipe centerline,

2. Length of the pipe,

3. Nominal pipe diameter,

4. Time, and

5. Axial position along the pipe;

c. Correlate the results of the parametric study by a se t of geometric correction factors;

d. Develop a procedure for calculating dose rates by using the correction factors; and

e. Verify the correlation scheme by calcula ting the dose rates at different target locations due to source piping of varied geometries through the use of QAD-P5A computer code, and compare th e results to those obtained by using the procedure developed in step d.

J.B.3.3 Direct Dose Para metric Study Results Inside Secondary Containment

The standard pipe gamma dose rate and integrated dose curves for the different systems having different source term assump tions (defined in Section J.5.3.2) are shown in Figures J.B-4 through J.B-11. The various correction factors were calculated by the following correlation.

F R (r) = Dose rate at a radial distance "r" from an infinitely long 8-in. sch 40 pipe Dose rate at a radial distance of 8 ft from an infinitely long 8-in. sch 40 pipe F L () = Dose rate at a radial distance of 8 ft from an 8-in. sch 40 pipe of length "2" Dose rate at a radial distance of 8 ft from an infinitely long 8-in. sch 40 pipe F D (d) = Dose rate at a radial distance of 8 ft from an infinitely long sch 40 pipe of nominal diameter "d"Dose rate at a radial distance of 8 ft from an infinitely long 8-in. sch 40 pipe The above mentioned correction factors for liquid system source terms are shown in Figures J.B-12 , J.B-13 , and J.B-14. The correction factor curves for gaseous source terms are shown in Figures J.B-15 , J.B-16 , and J.B-17.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-15 J.B.3.4 Correction Factor Met h od of Determining Dir e ct Dos e s in Secondary Containment

Using the parametric curves from Section J.B.3.3 , one obtains dose rates at varied radial distances (between 2 ft to 40 ft) from varied pipe diameters (between 2 in. to 24 in.) of varied lengths (between 2 ft to infinity) at any given time period within 1 year. The step-by-step procedure for calculating direct dose is as follows:

a. Identify a/r, b/r para meters and obtain pipe le ngth correction factor F L from Figure J.B-13 or J.B-16 , depending on the system being considered. (See Figure J.B-18 for definition of "a/r" and "b/r");

b. Obtain the standard dos e rate from the standard dose rate curve for time "t" desired;

c. Obtain the pipe diameter correction factor F D (d);

d. Obtain radial distan ce correction factor F R (r); and

e. The dose rate for the given pipe segment can be computed by Dose Rate = (Standa rd Dose Rate) (F R)(F D)(F L).

Table J.B-2 compares the results for dose rate of 17 different pipe ge ometry and target locations as calculated using the correction factor method to those calculated by using the computer code QAD-P5A. It was observed that the biggest difference in results between the two methods is less than 10%.

It is concluded that the correc tion factor method is adequate for calculating direct dose.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-17 Table J.B-1 Gamma Energy Concentration (photons/sec-c m 3) in Liquid-Containing Systems Gamma Energy (MeV)

Time 0.30 0.63 1.10 1.55 1.99 2.38 2.75 3.25 3.70 4.22 4.70 5.25 0 min 1.32E+09 7.25E+09 2.33E+09 1.63E+09 1.28E+08 1.00E+

08 1.33E+08 3.61E+07 1.90E+07 2.82E+07 4.58E+07 3.39E+05 2 min 1.17E+09 7.17E+09 2.03E+09 6.17E+08 1.25E+08 4.54E+

07 4.88E+07 2.42E+07 1.03E+07 7.17E+06 1.02E+07 2.10E+05 6 min 1.06E+09 6.92E+09 1.85E+09 5.63E+08 1.22E+08 1.99E+

07 1.55E+07 1.62E+07 7.34E+06 8.50E+05 5.84E+05 8.09E+04 20 min 9.71E+08 6.21E+09 1.69E+09 5.00E+08 1.14E+08 1.30E+

07 8.21E+06 9.17E+06 5.25E+06 2.03E+04 3.55E+03 2.86E+03 1 hr 8.84E+08 4.75E+09 1.41E+09 3.84E+08 1.01E+08 7.71E

+06 3.36E+06 2.60E+06 2.17E+06 1.43E+00 3.66E-01 2.01E-01 3 hr 8.50E+08 2.65E+09 9.92E+08 2.28E+08 7.75E+07 2.77E

+06 3.61E+05 3.74E+05 1.58E+05 3.26E-03 1.55E-03 9.71E-04 9 hr 9.04E+08 1.29E+09 5.00E+08 1.05E+08 3.88E+07 7.71E

+05 9.04E+03 2.35E+04 6.17E+01 3.26E-03 1.55E-03 9.71E-04 1 day 8.09E+08 7.17E+08 1.39E+08 3.73E+07 8.84E+06 6.17E

+05 1.30E+03 6.29E+01 5.17E-03 3.26E-03 1.54E-03 9.71E-04 3 days 5.54E+08 2.71E+08 1.79E+07 1.91E+07 9.34E+05 5.71E

+05 1.10E+03 3.56E+01 5.17E-03 3.26E-03 1.54E-03 9.67E-04 9 days 3.16E+08 1.22E+08 4.58E+06 1.36E+07 5.71E+05 4.29E

+05 1.09E+03 3.46E+01 5.13E-03 3.22E-03 1.53E-03 9.59E-04 30 days 5.42E+07 6.46E+07 1.73E+06 4.38E+06 2.67E+05 1.48 E+05 1.05E+03 3.31E+01 4.96E-03 3.12E-03 1.48E-03 9.29E-04 60 days 8.17E+06 4.42E+07 9.29E+05 1.03E+06 1.49E+05 3.94 E+04 9.92E+02 3.13E+01 4.71E-03 2.98E-03 1.41E-03 8.88E-04 90 days 3.99E+06 3.45E+07 7.13E+05 3.60E+05 1.14E+05 1.75 E+04 9.38E+02 2.96E+01 4.54E-03 2.86E-03 1.35E-03 8.50E-04 120 days 3.25E+06 2.77E+07 6.25E+05 2.19E+05 1.00E+05 1.26 E+04 8.88E+02 2.80E+01 4.38E-03 2.75E-03 1.30E-03 8.17E-04 150 days 2.90E+06 2.27E+07 5.79E+05 1.83E+05 9.17E+04 1.11 E+04 8.38E+02 2.64E+01 4.21E-03 2.66E-03 1.26E-03 7.92E-04 180 days 2.64E+06 1.89E+07 5.42E+05 1.69E+05 8.50E+04 1.04 E+04 7.92E+02 2.50E+01 4.08E-03 2.57E-03 1.22E-03 7.67E-04 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.B-18 Table J.B-2 Comparison of Direct Dose Rate Results Target Location and Pipe G e ometry Dose Rate Results Pipe Pipe Target Location Time After Correction Computer Diameter Length r a b Accident Factor Method Results Difference (cm) (cm) (cm) (cm) (cm) (hr) (rad/hr) (rad/hr) (%) 6 800 548.6 570 230 24 52.7 53.3 -1.1 6 800 91.4 720 80 24 484 479 +1.0 6 800 1006.8 650 150 24 15.3 16.1 +5.0 8 800 548.6 570 230 24 77.0 80.4 -4.23 8 800 391.4 720 80 24 105.0 110.0 -4.5 8 800 1066.8 650 150 24 22.4 24.4 -8.2 2 700 100.0 600 100 720 5.36 5.32 0.75 2 700 1066.8 600 100 720 0.159 0.146 8.9 2 700 100 -900 1600 720 0.0126 0.0123 2.3 2 700 1066.8 200 900 720 0.128 0.124 3.2 12 400 1066.8 -400 800 720 1.14 1.21 -5.8 12 400 100 350 50 720 72.5 71.4 1.5 12 400 609.6 350 50 720 4.66 4.93 -5.5 12 400 1066.8 350 50 720 1.57 1.73 -9.3 10 600 304.8 -243.8 548.

6 0.0333 554 539 2.7 10 600 121.9 450 150 0.0333 7617 7396 -3.0 10 600 1005.8 450 150 0.0333 258 280 -7.9 Model of the Primary and Secondary Containment 970187.24 J.B-1 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev. Q 2 = Air Leakage Rate from Secondary Containment (m 3/sec) Q 1 = Air In-leakage Rate from Primary Containment (m 3/sec) V 2 = Volume of Secondary Containment (m

3) V 1 = Volume of Primary Containment (m

3) i = Nuclide Index C 2i (t) = Activity Concentration in Secondary Containment (Ci/m 3)C 1i (t) = Activity Concentration in Primary Containment (Ci/m

3) Q in = Clean Air In-leakage Rate (m 3/sec)Primary Containment Standby Gas Treatment System Q in V 2 V 1 C 1i C 2i (t)Q 2 Q 1 Secondary Containment (t)Columbia Generating Station Final Safety Analysis Report Time-Dependent Gamma Dose Rate for a Semi-Infinite Cloud of Fission Products at Secondary Containment Concentrations 970187.25 J.B-2 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev. 0.5%/Day Primary Containment Leakage Rate 10-1 10 0 10 1 10 2 10 3 10 4 10 1 10 2 10 3 10 4 Time (Hrs)

Dose Rate (Rads/Hr)

Columbia Generating Station Final Safety Analysis Report Illustration of Parameters Used in the Shielding Equation 970187.26 J.B-3 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Line Source b i = i t i iTarget Location P a b r 0 1 0 2 Columbia Generating StationFinal Safety Analysis Report Standard Gamma Dose Rate Curve for Liquid Containing Systems (RCIC Liquid System and RHR System) 970187.27 J.B-4 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 1 10 2 10 3 10 4Time After Accident (Hrs)

Standard Dose Rate (Rads/Hr) 10-2 10 0 Columbia Generating StationFinal Safety Analysis Report Standard Integrated Gamma Dose Rate Curve for Pipes in Liquid Containing Systems (RCIC Liquid System and RHR System) 970187.28 J.B-5 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 3 10 4 10 5 10 6Time After Accident (Hrs)Total Integrated Dose (Rads) 10-2 10 2 Columbia Generating StationFinal Safety Analysis Report Standard Gamma Dose Rate Curve for Pipes in theRCIC Steam System and MSIV-LCS SteamSystem Before the Header 970187.29 J.B-6 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 1 10 2 10 3 10 4Time (Hrs)

Dose Rate (Rads/Hr) 10-2 10 0 10 5 Columbia Generating StationFinal Safety Analysis Report Standard Integrated Gamma Dose Curve for Pipesin the RCIC Steam System and MSIV-LCS SteamSystem Before the Header 970187.30 J.B-7 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 4 10 5 10 6 10 7Time (Hrs)

Integrated Gamma Dose (Rads) 10 3 10 8 Columbia Generating StationFinal Safety Analysis Report Standard Gamma Dose Rate Curve for Pipes in theMSIV-LCS Steam System After the Header 970187.31 J.B-8 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 0 10 1 10 2 10 3Time (Hrs)

Dose Rate (Rads/Hr) 10-2 10-1 10 4 Columbia Generating StationFinal Safety Analysis Report Standard Integrated Gamma Dose Curve for Pipesin the MSIV-LCS Steam System After the Header 970187.32 J.B-9 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 2 10 3 10 4 10 5Time (Hrs)

Integrated Gamma Dose (Rads)

Columbia Generating StationFinal Safety Analysis Report Amendment 59December 2007 970187.33 J.B-10 Figure Form No. 960690FHDraw. No.Rev.LDCN-06-039 Columbia Generating StationFinal Safety Analysis Report Deleted Amendment 59December 2007 970187.34 J.B-11 Figure Form No. 960690FHDraw. No.Rev.LDCN-06-039 Columbia Generating StationFinal Safety Analysis Report Deleted Radial Distance Correction Factor for Liquid Sources 970187.35 J.B-12 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.1.0 0.1 Radial Distance of Target from Source Pipe (Ft) 0.2 0.3 0.4 0.5 1.02.03.04.05.010.020.030.040.060.0 2.0 3.0 4.0 5.0 10.0 Radial Distance Correction Factor (F R)Columbia Generating Station Final Safety Analysis Report Pipe Length Correction Factor for Liquid Sources 970187.36 J.B-13 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.0.1 L/r (Dimensionless)

Pipe Length Correction Factor F L 1.0 10.0 20.0 0.03 0.1 1.0 Columbia Generating StationFinal Safety Analysis Report Pipe Diameter Correction Factor forLiquid Sources 970187.37 J.B-14 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Pipe Diameter Correction Factor F D Pipe Diameter (Inches)24681012141618202224 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

3.2 Columbia

Generating StationFinal Safety Analysis Report Radial Distance Correction Factor forGaseous Sources 970187.38 J.B-15 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.1.0 0.1 Radial Distance Correction Factor F RRadial Distance of Target from Source Piping (Ft) 0.2 0.3 0.4 0.5 1.02.03.04.05.010.020.030.040.050.0 2.0 3.0 4.0 5.0 10.0 9.0 8.0 7.0 6.0 9.0 8.0 7.0 6.0 Columbia Generating StationFinal Safety Analysis Report Pipe Length Correction Factor forGaseous Sources 970187.39 J.B-16 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Pipe Length Correction Factor (F L)( /r)1 2 Pipe Length Correction Factor (F L)3 4 5 6 7 8 9 10 11 12 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Columbia Generating StationFinal Safety Analysis Report Pipe Diameter Correction Factor for GaseousSources 970187.40 J.B-17 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Normal Pipe Diameter (Inches)

Pipe Diameter Correction Factor (F D)2 4 681012 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 Columbia Generating StationFinal Safety Analysis Report Amendment 58December 2005Parameters Used for the Calculation of LengthCorrection Factor 970187.50 J.B-18 Figure Form No. 960690FHDraw. No.Rev.LDCN-05-000 Columbia Generating StationFinal Safety Analysis Report a b Source Pipe Configuration 1 a b Source Pipe Configuration 2TargetTarget r F L =F L (r ) - F L (r )2 F L =F L (r ) + F L (r )2a 2 r 2b 2a 2b C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-1 Attachment J.C PROCEDURE FOR THE CALCULATION OF SECONDARY

CONTAINMENT RADIATION ZONE GAMMA DOSES

J.C.1 INTRODUCTION

Three Mile Island Lessons Learned S hort Term Recommendations (NUREG-0578)

Section 2.1.6.b, requires all nuc lear power plant licensees to calculate post-loss-of-coolant accident (LOCA) environmental conditions for all safety-related equipment. This procedure is specifically concerned with the definition of the postaccident radiological environments in the

secondary containment of Columbia Generating Station (CGS), a BWR.

The assumptions used in this procedure are based on a nonmech anistic LOCA scenario in which core damage is experienced at the be ginning of the accident and primary containment isolation is achieved prior to radiation transport.

The radiation level at a given location inside the secondary containment of CGS during and

following such an accident is defined by the following major source contributors.

Airborne gamma dose Gamma ray dose from airborne ra dioactive sources inside secondary containment

Containment shine dose Gamma ray dose fr om radioactive sources suspended in the drywell and the wetwell inside primary containment

Direct gamma dose Gamma ray dose fr om piping containi ng recirculating radioactive fluids

Bioshield penetration

streaming dose Gamma ray dose from liqui d piping and airborne

radioactive sources inside primary containment which

stream through bioshield wall pe netrations into secondary containment

The methods presented in this procedure make it possible to calculate the worst-case gamma ray dose due to the above mentioned source of contributors inside radiation zones (see Section J.C.2 for the definition of radiation zones) of the secondary containment of CGS. The radiation zone dose calculated by using this procedure is appli cable solely for the purpose of environmental qualification of safety-related equipment.

The following sections of this procedure describe the nomenclature, assumptions, and methods

used in calculating radiation dose ra tes and cumulative doses. Section J.C.2 defines the terms and nomenclature found in this procedure. The assumptions and approximation used in C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-2 developing the dose rate calculation method, as well as limitations to this method, are stated in Section J.C.3. Section J.C.4 provides a step-by-step procedur e for determining the worst-case gamma dose rate and cumulative dose inside a pa rticular radiation zone. The calculation of airborne beta dose is defined in a separate calculation procedur e and is not included in this procedure (see Attachment J.E

). J.C.2 DEFINITION OF TERMS

This section contains the definition of the te rms and symbols as used in this procedure:

CIND: Cumulative integrated dose (rads) Cumulative dose due to exposure to the decaying radioactive sources.

D a: Airborne gamma dose rate (rads/hr) Gamma dose rate resulting from radioisotopes suspended in the atmosphere of the secondary containment.

D d: Direct dose rate (rads/hr) Gamma dose rate resulting from the radioactive fluid contained inside recirculating pipes.

D s: Shine dose rate (rads/hr) Gamma dose rate in the secondary containment resulting from radioisotopes suspended and deposited insi de primary containment.

D B: Bioshield penetration streaming dose (rads/hr) Gamma dose rate contributed by th e liquid piping and airborne radioactive sources inside primary containment which stream through the bioshield wall.

D t: Total gamma dose rate (rads/hr) Gamma dose rate contributed by the sum of airborne, direct, and shine from penetrations into s econdary containment.

D t = D a + D d + D s + D B GF: Geometric factor Scaling factor used to convert semi-inf inite airborne gamma dose to finite dose inside enclosed air spaces.

D D a GF a ,

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-3 GF V11730338. (Reference J.7-39) F L: Length conversion factor A scaling factor dependent on the s ource pipe segment length and spatial orientation relative to a target (see Figure J.C-1 for the calculation of this factor). F L is used to convert the standard dose to the dose emitted by a pipe segment of finite length.

F D: Diameter conversion factor A scaling factor dependent on the source pipe diameter. F D is used to convert the standard dose to the dose emitted by a pipe of specified diameter.

F R: Radial distance conversion factor A scaling factor dependent on the radial distance of the target from the source

piping. F R is used to convert the standard dose to the dose at a target of specified radial distance from the source piping.

F t: Total dose contribution correction factor A scaling factor used to convert the sta ndard dose to the dose at a target from a pipe segment of specified geometry and orientation.

F t = F D F R F L F s: Sum of dose contribution correction factor A scaling factor used to convert the sta ndard dose to the radi ation zone dose due to all the significant pipe sources in the zone.

F n F ti il S Radiation zone: A region in the secondary containment defined to be such that gamma radiation calculated in the zone bounds the magnitude of dose received by the pieces of safety-related e quipment located in that zone.

Source term

The total radiated gamma ener gy associated with a specified quantity of radioactive material released from the reactor as the result of a postulated accident.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-4 Special sources

Radioactive source of such ge ometry or concentration that cannot be approximated by pi pe segments of diameters 2 in. through 24 in. and containing contaminated liquid of ac tivity concentration established in Section J.C.3.1. This can be a heat exchanger, standby gas treatment filter, pump, etc.

Standard dose

Gamma dose at a target having a radial distance of 8 ft from the centerline of an infinitely long, 8-in.- diameter schedule 40 pipe.

Target: The point in space chosen to repr esent the location of an object for which a dose rate and/or cumula tive dose is being calculated.

Worst case target

Location of the piece of safety-related equipment inside a radiation zone which will experience the highest gamma dose among all the pieces of safety-related equipment in that zone.

J.C.3 ASSUMPTIONS, APPROXIMATIONS, AND LIMITATIONS

J.C.3.1 Basic Assumptions to be Used in the Analysis

Gamma doses and dose rates insi de radiation zones will be de termined for four types of radioactive source distribution:

Major Source Contributors Airborne gamma dose Isotope s suspended in the atmosphere of the secondary containment

Shine dose Gamma irradiation fr om the primary containment Direct dose Direct gamma irradiation from the radioactive fluid contained inside recirculating pipes

Streaming dose Gamma irradiation from liquid piping sources inside primary containment and primary contai nment atmosphere streaming through bioshield wa ll penetrations

The dose contributed by each of these sources is determined by the location of the equipment, the time dependent distribution of the s ource, and the effects of shielding.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-5 The assumptions used in determining the initial di stribution and leakage of radioactivity in the primary containment are as follows:

a. 100% of the noble gases and 50% of the halogens initially in the reactor core will be distributed homogeneously within the primary containment free volume immediately following the postulated accident. Plateout of 95% of the elemental iodines is allowed to occur in accordance with Reference J.7-34;

b. 50% of the halogens and 1% of the remaining fission products in the core will be mixed homogeneously with the primary containment liquid space instantaneously. The primary containmen t liquid space is defined as the sum of the suppression pool liquid and the react or coolant system (RCS) liquid.

Assumptions a and b are NRC-reco mmended assumptions for defining radioactivity release fractions for the qua lification of safety-related equipment (Reference J.7-2) and are consistent with the accident analysis (Reference J.7-13);

c. The core fission product source term is defined as the total product generated in the core after 1000 days at a reactor power of 3556 MWt. This represents the

maximum burnup level in the core prio r to radioactivity release and is conservative; and

d. Primary containment leakage of 0.

50% volume/day was considered and is consistent with the assumptions established in Reference J.7-13.

J.C.3.1.1 Assumptions Used in the Calculati on of Airborne Dose Ra te Inside Secondary Containment

a. Activity that leaks into the secondary containment is homogeneously mixed with the secondary containment atmosphere prio r to its removal from the atmosphere by the standby gas treatment system (SGTS) exhaust fans. This is consistent

with the NRC-recommended assumptions us ed for calculation of doses inside primary containment (Reference J.7-2);

b. The SGTS flow rate of 2430 scfm is assumed to be the flow rate of the effluent air and is based on one reactor building air change per day;

c. Air that leaks out of the primary cont ainment flows directly into the secondary containment. Bypass leakage is not c onsidered. This is conservative when considering dosage in the secondary cont ainment, since it maximizes the buildup of radioactivity in the s econdary containment; and

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-6 d. Geometric factors provide a good appr oximation to convert the semi-infinite cloud dose to a finite cloud dose and is based on the results presented in Reference J.7-28 and based on average gamma ra y energy of 0.733 MeV. The effect of variation of this parameter due to difference in gamma ray energies have been proven to be negligible (see Attachment J.B for justification).

J.C.3.1.2 Assumptions Used for the Calculati on of Shine or Streaming Dose From Primary Containment

a. No depletion of activity due to l eakage is assumed to maximize the source activity and is conservative;

b. The airborne source is assumed to be uni formly distributed in the drywell and in the wetwell air space. The effect of the pl ateout of iodine is not considered in secondary containment;

c. Activity in the wetwell water volume is assumed to be uniformly distributed in the sump water. Assumptions b and c are based on the plateout modeling and source term assumption cont ained within References J.7-2 and J.7-34;

d. The dosage at a point inside the region closest to the source is considered to be representative of the gamma dose in th e region which maximizes the gamma ray dose at the region and is conservative; and

e. The liquid piping sources inside pr imary containment are assumed to be uniformly distributed in the RCS for the first 17 hr post-LOCA. The liquid piping sources inside primary contai nment are assumed to be uniformly distributed in the RCS plus the s uppression pool after the first 17 hr post-LOCA. This is consistent w ith the CGS operations procedure to depressurize and utilize the alternate s hutdown cooling mode within 17 hr post-LOCA once a degraded core condition is identified.

J.C.3.1.3 Assumptions and A pproximations Used in the Ca lculation of Direct Doses

a. No valve leakage is assumed, which is consistent with Reference J.7-5 , Item II.B.2, Clarification (2);

b. Schedule 40 piping is assu med, which is a conserva tive simplification of the calculation process. Because the majority of the pipe segments considered are schedule 40 piping, and because increases in pipe schedule can only decrease the dose rate at the targets, this approximati on is considered to be conservative and appropriate;

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-7 c. Heat exchangers and pumps can be a pproximated as pipe systems. The volume of radioactive liquid in the component and its length are used to determine an equivalent volume of liquid. This is a crude approximation for dose rates contributed by complex geometries.

Because the pump and heat exchanger walls are thicker than the pipe walls of schedule 40 piping, this assumption is conservative; and

d. Radioactive piping with diameters 2-1/2 in. or less was not modeled unless it was determined that such a pipe wa s a major source contributor. A major source contributor is defined as the only radioactive pipe in a target area or the radioactive pipe of closest proximity to the target. This is made because the dose contributions due to pipe segments of diameter less than 2-1/2 in. are generally negligible, unless they are major source contributors.

J.C.3.2 Limitations

The following limitations apply to the use of th is procedure for the calculation of radiation zone doses.

a. This procedure is only applicable to the calculation of ra diation zone gamma doses in the secondary containment of CGS;

b. The assumptions stated in Section J.C.3.1 are basic to the methodology used in this procedure. Changes in any of th e assumptions will affect the accuracy of the results generated us ing this procedure;

c. The calculation of direct doses using the generic curves in this procedure is limited to liquid sources in schedule 40 pipe segments or equivalent pipe segments with nominal pipe diameters ranging from 2 in. to 24 in. Any

deviation from these pipe geometries should be modeled as special cases.

Note: Schedule 40 piping is used because the majority of the pipe segments to be considered are standard pipes (sche dule 40). Increases in the pipe schedule only introduces conservatism in the results;

d. The results for direct dose calculated us ing the generic curves were found to be accurate to within 10% (see Reference J.7-39 for error study); and

e. Source piping located 40 ft or furthe r from the target is generally an insignificant dose contributor. If its contri bution is not found to be negligible, it should be considered as a special source.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-8 J.C.4 PROCEDURES FOR THE CALCULATION OF SECONDARY CONTAINMENT RADIATION ZONE DOSES This procedure describes the method used in calculating the gamma radiation doses inside radiation zones. For equipment located inside a zone, the follo wing four sources contribute to the total dose level.

a. Airborne dose (gamma),

b. Direct gamma dose from sources within pipes,

c. Direct gamma shine dose fr om drywell and wetwell, and d. Gamma streaming dose fr om drywell and wetwell.

A step-by-step procedure is discussed in the following sections for the calculation of the

maximum total gamma dose a nd dose rates for each zone.

J.C.4.1 Procedure A: Radia tion Zone Dose Calculation

The first step in preparing a zone dose calculation is to identify all the parameters to be used.

This includes the identification of all the potentia l sources and targets, both inside and outside the zone, and the identification of the dimensions of the zone.

Figure J.C-2 is a step-by-step flowchart of the calculation procedure. When identifying sources outside the zone, sources at the upper and lower elevations in the review process are included. A conservative dose estimate is used to determine whether a source out side a zone is a significant contributor. For example, if the closest pipe segment in the zone is a few feet away from a target, then the dose estimate will show that a pipe segment outside th e room at 30 ft is insignificant by comparison.

Conversely, if a target is located near a wall with several pipes on the other side of a wall, then those pipes may become significant source contri butors and are included in the final evaluation for the target.

J.C.4.2 Procedure B: Airborne Dose Calculation in Secondary Containment

Because the semi-infinite airborne dose and dos e rates are already calculated and shown in Figures J.C-6 and J.C-7 , the only calculation involved in dete rmining the airborne dose is the conversion of the semi-infinite cloud dose at react or building concentrations to a finite cloud dose inside the cubicles in which the radiation zones are defined. The first step in this calculation is to determine the volume which defi nes the air space (or zone) of interest. An enclosed air space is defined as a cubicle, at least 95% shielded by c oncrete (or equivalent shielding) at least 1 ft thick.

To convert a semi-infinite cloud dose (calculated in Reference J.7-38) to a finite cloud dose, a geometric factor is used.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-9 Dt Dt GF a a ()(), (J.4-1) whereGF V11730338. (Reference J.7-39) (J.4-2) GF = geometric factor (dimensionless)

V = volume of the enclosed air space (ft

3)

Similarly, CINDtCINDt GF a a ()(), (J.4-3)

Figure J.C-3 is a step-by-step flowchart of the procedure for calculating airborne gamma doses.

J.C.4.3 Procedure C: Primary C ontainment Shine Dose Calculation

Containment shine doses are calculated using th e QAD-P5A computer code. Guidelines for preparing input parameters are docum ented in Procedure E and Reference J.7-10. The modeling procedure and the accuracy of the results are highly dependent on the geometry to be modeled, specification of the source volume, and the selection of a buildup factor.

Figure J.C-4 is a step-by-step procedure for calculating containment shine doses.

J.C.4.4 Procedure D: Direct Dose Calculation

The first step in the direct dose calculation (from Reference J.7-39) is the identification of the "worst-case" target. Normally, the worst-case ta rget is the piece of equipment that is closest to the major source piping and can be selected by in spection. However, if situations arise such that the worst-case target cannot be chosen by simple inspection, order-of-magnitude calculations are performed for each potential worst-case target in the zone. These calculations are illustrated in Steps 3a through 3c of Figure J.C-5.

The next step is to identify special sources.

Special sources are defined as source geometries that cannot be represented by liquid pipe segments between 2 and 24 in. in diameter. Example

special sources are SGTS filters, reactor core isolation cooling (RCIC) steam pipe, turbines, and heat exchangers larger than 24-in. diamet er. Other components such as pumps and small heat exchangers should be mode led as pipes. The pipe cro ss-sectional area is calculated by dividing the total fluid volume by the effective length of the component.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-10 The contribution due to sources with sh ield walls is investigated next.

Figure J.C-13 is used for this evaluation. If these sources are dete rmined to be significant contributors, special QAD-P5A modeling procedures as desc ribed in Procedure E are followed.

It is unlikely that all sources under considera tion will contribute significantly to the dose at a specific target. If all source contributions we re to be calculated, the time involved in performing the calculation would be unnecessa rily long without making a substantial improvement in the accuracy of the results.

Hence, as the sources are being identified, good judgment is used to distinguish between sources which contribute significantly to the target dose and those sources which do not.

An insignificant source is determined by compar ing its dose contribution to the source making the largest dose contribution. The comparison is facilitated by arranging sources in decreasing order of importance and assigning rank numbers to the sources.

The largest dose contributor is given a ranking number of 1.

The largest dose contributor is determined by inspection of the sketches and drawings being used. The la rgest dose contributor is generally the longest segment with the largest pipe diameter and th e least amount of intervening shielding between the target and source. All sources which are in the radiation zone and ha ve been assumed to be insignificant contributors are listed as such to indicate that those sources have been considered.

Equations Used in the Calculation of Dose Rates

The following procedure is followe d for the calculation of correc tion of dose rates factors of dose rates (Step 9 through Step 12 of Figure J.C-5

): a. Identify the radial distance of the pipe segment from the target; read F R from Figure J.C-11. If the target is in contact with the source piping, read F D from Table J.C-1 and set F R and F L equal to 1.

(Note: dose rate is not a function of pipe length and radial distance.)

If the target is geometrically in line w ith the source pipe segment, as shown in configuration 3 of Figure J.C-1 , set F L=1 and read F D and F R from Figures J.C-14 and J.C-15 , respectively.

(Note: F L is defined here because dose rate is not sensitive to pipe length variation.)

b. Identify the pipe diameter; read F D from Figure J.C-10.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-11 c. Determine F L from Figure J.C-12

use equations in Figure J.C-1 to calculate this factor.

d. The total dose contribution factor for a given pipe segment (I) is given as

F t (I) = F D (I) F R (I) F L (I) e. When all the significant contributions have been calculated, sum the total dose contribution factors.

F n F n I st1 ()

f. To determine if a source is negligible , the following test should be performed:

When N source segments are being cons idered and the dose contribution of ranking I is less than 1/10 of the dose ra te calculated from the largest source divided by (N-I), the sources remaining shoul d not contribute more than 10% to

the total source contribution. This le vel of accuracy should be adequate for most calculations.

The total integrated direct dose and dose rate can be calculated.

D D (t) = D Do (t) F s + D D (t) (Special Sources)

CIND D (t) = CIN D Do (t) F s + D D (t) (Special Sources) where D Do (t) and CIND Do (t) are dose rates and cumulative doses for standard pipe segments and are found on Figures J.C-8 and J.C-9.

J.C.4.5 Procedure E: QAD-P5A Modeling Procedure

Direct dose contribution due to special sources and/or sources with shield walls should be calculated using the QAD-P5A computer code. This computer code is three-dimensional and calculates dose rates at specified target locati ons from radioactive volume, line, and point sources. Attenuation due to shield materi als, if applicable, is also applied.

The accuracy of the results is highly affect ed by the manner by which the source volume is divided, and the position of the target relative to the source poi nt. Therefore, a sensitivity study on the specification of the source volume shoul d be performed. This can be achieved by C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-12 increasing the number of source volume divisions until the dose rate results converge to within 5%.

Another factor to be considered is the specification of the buildup factor. As a general rule, aluminum buildup factor should be used when c oncrete shield is encountered, and iron energy buildup factor should be used when cons idering attenuation through steel shield.

J.C.4.6 Procedure F: Streaming Dose Calculation

Containment streaming doses through the bioshiel d wall penetrations are calculated using the SCAP-BR and QAD-CG computer codes. The modeling procedure and the accuracy of the

results are highly dependent on the geometry to be modeled, specification of the source volume, and the selection of a buildup factor.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.C-13 Table J.C-1 Diameter Correction Factor (F D) for Targets in Contact With the Source Piping

Nominal Pipe Diameter Pipe Diameter Correction Factor (in.) (F D) 2 18.4 4 24.4 6 54.6 8 33.3 10 35.3 12 35.3 14 35.5 16 33.7 20 32.0 24 29.6 Calculation of Length Correction Factor 970187.51 J.C-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.a b Source Piping Configuration 1 a b Source Piping Configuration 2TargetTarget r F L =F L (r ) - F L (r )2a 2 2b F L =F L (r ) + F L (r )2a 2 2b r Source Piping Configuration 3Target F L = 1 r Columbia Generating StationFinal Safety Analysis Report PLC T Total Dose Rate and Integrated Dose StartLocate Targets Inside Zone on the Zone SketchIdentify All Dose Contributing Sources Inside Zone and Locate on Zone SketchIdentify All Possible Dose Contributing Sources Outside Zone Calculate Direct Dose Gamma Dose Rate and Integrated Dose (Procedure D)Procedure A: Procedure for Calculating Radiation Zone Doses 970187.52 J.C-2 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Identify Dimensions of ZoneCalculate Airborne Gamma Dose Rate and Integrated Dose (Refer to Procedure B)

Calculated Shine and Streaming Gamma Dose Rate and Integrated Dose (Refer to Procedure C)Calculate Total Gamma Dose and Dose Rate D t= D a+ D d+ D s + D B CIND t = CIND a + CIND d + CIND s+ CIND BComplete Table 1 Problem Completed Step 1 Step 2 Step 3 Step 4 Step 5 Step 6 Step 7 Step 8 Step 9 Step 10 Columbia Generating StationFinal Safety Analysis Report Procedure B: Procedure for Calculating Airborne Gamma Dose Rate and Integrated Doses 970187.53 J.C-3 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Identify Shield Wall Thickness and Vent Openings Surrounding Zone Step 1 Step 2 Step 3 Step 4 Step 5 Step 6 Calculate Volume of Enclosed Air Space Yes No Identify Dimensions of the Enclosed Air Space in which the Zone is Located Step 2A Calculate Geometric FactorGF = 1173 Using Figures (6) and (7)Calculate Airborne Gamma Dose Rate and Integrated Dose D a (t) + Da (t)CIND a (t) = CINDa, (t)Return to Step 5 of Procedure A Is Criteria for Enclosed Air Space Satisfied?

From Step 5 of Procedure A GF V 0.338 GF Columbia Generating Station Final Safety Analysis Report Procedure C: Procedure for Calculation of Containment Shine Dose 970187.54 J.C-4 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Select Point Inside Zone which is Expected to have the Highest Shine Dose Step 1 Step 2 Step 3 Step 4 Step 5 Return toStep 6 of Procedure AFrom Step 6 of Procedure A Model the Primary Containment, Pertinent Reactor BuildingWalls and Floors According to the Schemes Established in Procedure EDiscretize the Airborne and Sump Sources Inside Containment Perform Sensitivity Study onSpecification of Volume Source Points until Results Converge to within 5%Run QAD-P5A to Establishthe Time Dependent Shine Dose Columbia Generating StationFinal Safety Analysis Report Procedure D: Procedure for Calculation of Direct Dose Rate and Integrated Dose 970187.55 J.C-5 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Step 11 Step 13Yes NoSET I = I +1Calculate Time Dependent Direct Dose Rate and

Integrated Dose by Summing

All Dose Contribution.

Return toStep 7 of Procedure A Ft(I) < 0.1 (Ft(1) / (n - I))

Sum up all Dose Contribution

Correction Factors

Fs = Fs + Ft(I)Calculate Total Dose Contribution

Correction Factor

F t (I) = F D (I)

• F R (I)

• F L (I)Step 10 For the I TH Pipe Segment, S1, of Order of Importance "I", Calculate F D (I), F R (I) and F L (I) by using Table 2 and Figures J.C-10 , 11 and 12. (note)Step 9 For N Pipe Segments inside Zone, Set I = 1 Step 8 Calculate Dose Contribution

due to Special Sources, Refer to Procedure E for

this Calculation.Yes No Is Dose Contribution due to Special Sources Significant?

Step 5 Calculate Dose Rate for

Sources Outside Zone (Refer to Procedure E)Yes Is contribution due to Sources Outside Zone Important? Use Figure J.C-13 for this Evaluation.

Step 3 Evaluate Sources; Identify by

Inspection their Order of

Importance, (I), and Complete Table 2.Step 2Yes No Can a"Worst" Target be Selected by Inspection?

From the Sketch Generated in

Step 3 of Procedure A and Taking into Consideration

all Significant Sources

Outside the Zone, Identify the "Worst" Target(s).From Step 7 of Procedure AM = # Worst Targets

Selected, Evaluate, for each Target, the Order of

Importance of the Sources and Complete Table 2 for each Target.Step 3A Step 1 Step 12 Following the Guidelines of

Step 4 through Step 12, Roughly Calculate Total Dose

Contribution Correction Factors for each Target.

Step 3BCompare Total Dose for each Target and Select the Worst Target.Step 3C Step 7 Step 6 Step 4Note:If Target is in contact with the Source, Read F D from Table 3, and Set F R and F L Equal to1. If Target is in line with the Axis of the Pipe Segment, as shown in Configuation 3 of Figure J.C-1

,

Set F L = 1 and Read F D and F R from Figures J.C-14 and J.C-15 respectively.

No Columbia Generating StationFinal Safety Analysis Report Time-Dependent Gamma Dose Rate for a Semi-Infinite Cloud of Fission Products at Secondary Containment Concentrations 970187.56 J.C-6 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev. 0.5%/Day Primary Containment Leakage Rate 10-1 10 0 10 1 10 2 10 3 10 4 10 1 10 2 10 3 10 4 Time (Hrs)

Dose Rate (Rads/Hr)

Columbia Generating Station Final Safety Analysis Report Time-Dependent, Integrated Gamma Dose Rate for a Semi-Infinite Cloud of Fission Products at Secondary Containment Concentrations 970187.57 J.C-7 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.10 0 10 1 10 2 10 3 10 4 10 2 10 3 10 4 10 5 Time (Hrs)

Integrated Gamma Doe (Rads) 10-1 10 1 10 6 0.5%/Day Primary Containment Leakage Rate Columbia Generating Station Final Safety Analysis Report Gamma Dose Rate at Target 8 ft Away from Standard Pipe 970187.58 J.C-8 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 1 10 2 10 3 10 4Time After Accident (Hrs)

Standard Dose Rate (Rads/Hr) 10-2 10 0 Columbia Generating StationFinal Safety Analysis Report Gamma Integrated Dose at a Target 8 ft Awayfrom Standard Pipe 970187.59 J.C-9 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10-1 10 0 10 1 10 2 10 3 10 4 10 3 10 4 10 5 10 6 Total Integrated Dose (Rads)

Time After Accident (Hrs) 10-2 10 2 Columbia Generating StationFinal Safety Analysis Report Pipe Diameter Correction Factor 970187.60 J.C-10 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Pipe Diameter (Inches)

Pipe Diameter Correction Factor (F D)24681012141618202224 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.8 3.0

3.2 Columbia

Generating Station Final Safety Analysis Report Radial Distance Correction Factor 970187.61 J.C-11 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.10 0 10-1 Radial Distance of Target from Source Pipe (Ft)

Radial Distance Correction Factor (F R)10 0 10 1 6x10 1 10 1 Columbia Generating Station Final Safety Analysis Report Pipe Length Correction Factor 970187.62 J.C-12 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.10-1 Pipe Length Correction Factor (F L)L/r (dimensionless) 10 0 10 1 2x10 1 3x10-2 10-1 10 0 Columbia Generating Station Final Safety Analysis Report Dose Rate Versus Concrete Shield-Thickness for Standard Pipe (8 in. Sch 40) 970187.63 J.C-13 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.1.0 2.0 3.0 1.0 10.0 100Shield Thickness (ft)

Dose Rate at Target Distance of 8 ft. (Rads/hr) 0 0.2Note: This Figure is to be used for estimation purposes only.

Refer to Procedure E for calculating Dose rates outside shield walls Columbia Generating StationFinal Safety Analysis Report Pipe Diameter Correction Factor for Targets Located Axially in Line with Source Piping 970187.64 J.C-14 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.10-1 10 0 10 1 Pipe Diameter Correction Factor (F D)Pipe Diameter Inches 3x10-2 10 1 5x10 1 Configuration 3 of Figure J.C-1 10 0 Columbia Generating Station Final Safety Analysis Report Distance Correction Factor fo r Targets LocatedAxially in Line with Source Piping 970187.65 J.C-15 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.10 1 7x10 1 10-2 10-1 10 0 Distance Correction Factor (F R)Axial Distance of Source from Target (Ft) 10-3 Configuration 3 of Figure J.C-1 10 0 Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-1 Attachment J.D CALCULATION OF THE RADIATION

The standby gas treatment system (SGTS) filters are located in the reactor building (el. 572 ft) and function to process the radioactive contamin ated gaseous effluent from the primary and secondary containment. In the event of a loss-of-coolant accident (LOCA) in the primary containment the SGTS will be actuated. The gase ous contaminants that leak out of the primary containment will be filtered by the SGTS. It will adsorb the iodines in th e charcoal filters and the particulates in the prefilters and high-efficien cy particulate air (HEPA) filters. Plateout in the primary containment of the iodines released from the core was considered in the radiation assessment of the SGTS. Depending on the radi oactive source distribution and the primary containment leakage rate, the radioactive iodine concentration in the filters will be increasing with time as more and more is deposited on the filters. Main steam isolation valve (MSIV) leakage is also considered in the radiation dose calculations.

The purpose of this study was to evaluate th e time dependent gamma radiation level for safety-related equipment locat ed near the SGTS filters a nd in adjacent rooms post-LOCA.

The time-dependent activity concentration in each of the filters is first calculated. The time and energy-dependent gamma activity levels on the SGTS filters is developed by a combination of computer runs and hand calcula tions and is used as input to the QAD-P5A computer code to calculate the gamma radiation levels for the pieces of safety-related e quipment located in the room. A discussion of the analysis follows.

J.D.1 DESCRIPTION OF THE STANDBY GAS TREATMENT SYSTEM FILTERS

Figure J.D-1 is a drawing of the SGTS filter train.

The SGTS consists of two fully redundant filter trains, each of which consists of the following components in series:

a. A demister to remove entrained wate r particles in the incoming air stream;

b. Two banks of electrical coil heater s designed to limit the humidity of the incoming air to 70% at design fl ow during post-LOCA conditions;

c. A bank of prefilters to remove la rge particles from the airstream (Figure J.D-2

);

d. A bank of HEPA filters to remove the remaining particulates from the airstream (Figure J.D-2

);

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-2 e. Two 4-in. deep beds of charcoal adsorber filters, arranged as shown in Figure J.D-1 , are designed to capture the elem ental and organic halogens from the airstream. The dimensions of the charcoal filters are shown in Figure J.D-3

and f. A second bank of HEPA filters, identical to that described in item d above. The function of this second HEPA filter bank is to capture contaminated charcoal dust which may escape from the charcoal filters.

Both SGTS filter units are locat ed in reactor building el. 572 and are automatically actuated and become fully operational within 34 sec of the event of any of the following three isolation

signals:

a. High radiation in the reactor building ventilation exhaust duct,

b. High drywell pressure, and

c. Low water level in the reactor vessel.

J.D.2 CALCULATION OF TIME-DEPENDENT FILTER ACTIVITY CONCENTRATION

The analysis of the time-dependent transport of the radioactivity from the primary containment to the SGTS filters and the activity concen tration on each filter is based on the following assumptions:

a. The SGTS filters are assumed to be loaded by iodine at a rate based on atmospheric leakage from primary contai nment of 0.67 wt %/day. This is composed of 0.5% direct from primar y containment leakage and 0.17% via the MSIV. This is based on the primary c ontainment rated leakage flow rate and

the calculated MSIV leakage (Reference J.7-40). The containment rated leakage flow rate is 0.5%/day. The MSIV leak age was originally determined to be 0.23%/day, but a reevaluation has resulted in a revision of the MSIV leakage to 0.17%/day as referenced in J.7-40. Since the revision resulted in a lower value the original analysis with MSIV leakage of 0.23%/day is conservative. Thus, the radiation zone calculati ons were not revised to reflect the MSIV leakage of 0.17%/day since the original analysis was conservative;

b. Straight exhaust through the filters, with no mixing or holdup in the secondary containment atmosphere, is assu med based on an NRC recommended assumption for the analysis for fission product control systems (Reference J.7-41);

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-3 c. The elemental iodine in primary cont ainment plateout on primary containment surfaces until one part in 200 of the elemen tal iodine remain airborne (0.5% of the total iodine). This is consistent with Reference J.7-14; d. The released halogen fraction is 50%

of the core inventory. This halogen fraction is assumed to be composed of 95.5% elemental, 2% organic, and

2.5% particulate iodines. This is consistent with Reference J.7-14;

e. The particulate halogens will be homoge neously distributed within the prefilters and the HEPA filters, while the elem ental and organic iodines will be homogeneously distributed within the two ch arcoal filters of the filter train.

This is conservative and necessary b ecause the time-dependent collection of iodines in the filters has not been defined. The homogenous assumption is reasonable; and

f. Leakage past the MSIVs discharges dir ectly to the inlet of the operating SGTS filter unit. Therefore, it bypasses the s econdary containment volume. This is conservative and necessary because the tim e dependent collection of iodines in the filters has not been defined.

The homogenous assumption is reasonable.

The time- and energy-dependent gamma activity c oncentration in the SGTS filters was first investigated as discussed in Section J.5.3.3. This analysis was perf ormed by a combination of computer analysis and hand calculations. The activ ity concentration of a halogen isotope inside a SGTS filter is changing with time due to the following three mechanisms:

a. Transport of activity from the primar y containment and deposition of the filters due to air leakage,

b. Depletion of activity due to radioactive decay and plateout of elemental halogens inside primary containment, and

c. Increases in activity levels due to daughter fission product generation from radioactive decay of other isotopes.

The activity balance on the SGTS filters can be described by (from equation J.B-16, Attachment J.B) d dt (A) = Q C(t) - A + A i 1 1i ii j j j (J.D-1) leakage - decay + growth in C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-4 where A i = activity (iodine) deposited on the SGT filters C 1i(t) = airborne concentrati on of iodine isotope "i" Q 1 = flow rate (volume) from the primary containment

As in Attachment J.B (equation J.B-1, J.B-2) the gr owth term is negligible.

C 1i (t) is given by equation J.B-8 of Attachment J.B as C 1i (t) = (S iH (t)/V 1) f H (t) exp (-

Q 1 t/V 1) (J.D-2)

V 1 is the volume of primary containment

f H (t) is defined by He-tpop f(t) = f e p + f + f where t t (J.D-3) H e po p f(t) = (f 200) + f + f where t t

Integrating (J.D-2) gives the following, wh ere B is a constant to be determined:

i -t -t t 1 li A(t) = Be i + e i e i Q C(t) dt (J.D-4)

C 1i (t) is substituted into (J.D

-4) from (J.D-2) to give i -t -t 1 1 iH H A(t) = Be i + e i V Q S(t) f(t) e/dti t QV 11 (J.D-5)

Substituting the definition of S iH (t) from equation J.B-5 of Attachment J.B , where A 1i (0) is the original activity in primary containment

(S(t)) = C (0) e i V = A (0) e i iHli -t 1 li -t i -tli -t 1 1 H 1 1 t A(t) = Be i + A (0) e i Q V f(t) e-(Q/V) dt xp (J.D-6) f H (t) consists of three chemical species: or ganic, particulate, and elemental iodine.

Equation (J.D-6) must be solved for each species, so the species w ill be separated at this point:

o (t) = f o C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-5 p (t) = f p (J.D-7) e (t) = f e p 0 t t p e-t e e p(t) = f 200 where t t f H (t) = o (t) + p (t) + e (t) To clarify the solution of (J.D-6), the following definitions are made:

X = p q = Q V --q 1 1 Since e (t) has step-function changes, solutions to (J.D-6) require a series solution - one for both of the time bands:

0 t p Organic Iodines

Equation (J.D-6) for all times t becomes

i -tlio -t-qt A(t) = Be i + A (0) q f e i e-q (J.D-8) At t=0, A i=0, so io l i-t-qt A(t) = f A (0) e i 1-e (J.D-9)

Since A 1i (0)e- i t = S iH (t) from equation J.B-5 (from Attachment J.B

), we define

A i (t) = S iH (t) o (t) (J.D-10)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-6 where o o-q t (t) = + f (1- e ) o(t) = fraction of organic halogens on the SGTS filters

Particulate Iodines

Particulate halogens are obtained in the same manner as organic haloge ns. The only difference is that f o is replaced by f

p.

Elemental Iodines For 0 t t p , equation (J.D-6) becomes i -t li -t-t e-qt A(t) = Be i + A (0) e i q (e p f) e dt (J.D-11) since at t = 0, A i = 0 i e li-t xt A(t) = q f x A(0) e i (e -l) (J.D-12) For t p t, equation (J.D-6) becomes i -t li -t e-qt A(t) = Be i + A (0) e i q (f 200) e dt (J.D-13) Integrating, with initial condition of t = t p i e li -t xt A = qf x A (0)e i p (e-1)p iel i i xt-qt-qt A(t) = f A (0) et q x (ep - 1) + x200q (ep - e) xptt (J.D-14) The activity on the SGTS filter may then be generally described by A i (t) = S iH (t) (t) (J.D-15) where (t) is the fraction of released iodines located on the filters and is defined by (t) = o (t) + p (t) + e (t) (J.D-16)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-7 where o o-q t(t) =f (1-e) p p-q t(t) = f (1-e) e(t) = e xt p e xt-1-qt-e-qt p F q x (e -1) (F o r 0 t t )

F q x (e p) + x 200 q (e p ) ( o r t t) F J.D.3 CALCULATION OF RADIATION DOSE FROM THE STANDBY GAS TREATMENT SYSTEM FILTER After the activity concentration in each filter segment is determ ined, the gamma radiation dose for safety-related equipment lo cated in the SGTS filter room is determined by the use of computer code QAD-P5A (Reference J.7-10). The QAD-P5A modeling procedure as described in Attachment J.C is followed for this analysis.

The following modeling assumptions were used:

a. Self-shielding of the filters is conservatively neglected because the density of the charcoal dust or the wire mesh (prefilter a nd HEPA filters) in the filters is low.

Neglecting the self-shielding effect of the filters will not add too much conservatism to the results; and

b. Shielding due to the sheet metal filter housing is conservativel y neglected due to computer code stability considerations.

The shielding effect of the thin sheet metal filter housing is negligible.

One zone and four subzones we re evaluated for the SGTS system and the five C1E/SRM components evaluated are

a. SGT-DV-1A3,

b. FPC-LIS-1A,

c. SGT-EHO-1B1,

d. SGT-MO-5B1, and

e. SGT-TE-6A1/7A1.

These targets are evaluated according to their proximity to the SGTS filters.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-8 The time-dependent, gamma ray activity con centration as calculated using the method described in Section J.D.2 was used as input to the QAD-P5A model described in Attachment J.C. The dose rate results of this analysis were integrated numerically to give time-dependent, integrated doses.

Table J.D-1 shows the direct gamma dose rate and integrated results for each of the five targets.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.D-9 Table J.D-1 Direct Gamma Dose Rate a nd Integrated Dose Results

for Targets in the Standby Gas Treatment System Room FPC-LIS-1A SGT-EHO-1B1 SGT-MO-5B1 SGT-TE-6Al/7A1 Time (h r) Dose Rate (r ad/h r) Integrated Dose (r ad) Dose Rate (r ad/h r) Integrated Dose (r ad) Dose Rate (r ad/h r) Integrated Dose (r ad) Dose Rate (r ad/h r) Integrated Dose (r ad) 0 8.6E+02 4.3E+01 2.7E+02 1.4E+01 9.9E+02 5.0E+01 4.8E+04 2.5E+03 1 4.1E+03 2.3E+03 1.3E+03 7.3E+02 4.7E+03 2.6E+03 2.3E+05 1.4E+05 3 3.8E+03 1.0E+04 1.2E+03 3.3E+03 4.4E+03 1.2E+04 2.1E+05 5.8E+05 9 2.3E+03 2.9E+04 7.6E+02 9.3E+03 2.6E+03 3.3E+04 1.2E+05 1.6E+06 24 1.6E+03 5.8E+04 5.1E+02 1.9E+04 1.7E+03 6.6E+04 7.7E+04 3.1E+06 72 1.2E+03 1.2E+05 3.8E+02 4.0E+04 1.3E+03 1.4E+05 5.4E+04 6.3E+06 216 1.2E+03 3.0E+05 3.9E+02 9.5E+04 1.3E+03 3.3E+05 5.5E+04 1.4E+07 720 5.7E+02 7.5E+05 1.7E+02 2.4E+05 6.1E+02 8.1E+05 2.5E+04 3.4E+07 1440 7.6E+01 9.8E+05 2.3E+01 3.1E+05 8.1E+01 1.1E+06 3.3E+03 4.4E+07 2160 7.7E+00 1.0E+06 2.5E+00 3.2E+05 8.2E+00 1.1E+06 3.3E+02 4.6E+07 4320 1.0E-02 1.0E+06 6.4E-03 3.2E+05 1.0E-02 1.1E+06 2.3E-01 4.6E+07 Standby Gas Treatment Filter 970187.66 J.D-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.46'-3" 5'-5 1/4" (G) Hepa Filter (F) Deep Bed Carbon Filter (E) Deep Bed Carbon Filter (D) Hepa Filter (C) Prefilter (B) Main Heater (A) Moisture Eliminator Columbia Generating StationFinal Safety Analysis Report Geometry of Prefilters and HEPA Filters 970187.67 J.D-2 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.24" Note A 24" Four Filters Note A: Prefilter 8" HEPA Filter 11 1/2 " Columbia Generating Station Final Safety Analysis Report Geometry of Charcoal Filters 970187.68 J.D-3 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.63" 42" 120" 4" (typ)Not to Scale Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.E-1 Attachment J.E BETA DOSE CALCULATION METHOD

The source volume used for the beta dose analysis in secondary containment is a sphere

surrounded by a shell of sufficient thickness to stop all outside beta particles from entering the source volume. This spherical source volume is conservative for any generalized source

volume shape. The dose at the center of the sphere is higher than the dose at any point of any generalized source of equal total volume.

The assumptions used in this analysis are as follows:

a. Atmosphere inside the equipment casing is identical to the atmosphere in the reactor building which is conservative becau se there will be some actual delay in transport of the gaseous fission products into the equipment;

b. The initial beta source term used wa s 100% of core noble gases and 50% of core halogens based on NUREG-0588, Revision 1 and NUREG-CR/0009 (References J.7-29 and J.7-34);

c. Daughter products of the airborne noble gases and halogens are included in the calculation of the airborne dose. This is conservative and was required by the

use of ORIGEN2 as a source code (Reference J.7-8);

d. Plateout of halogens inside primar y containment was utilized as allowed in accordance with Reference J.7-34. The dose contribution of fission products plated out on equipment casings was ne glected. This is based on the NRC recommended assumptions (Reference J.7-34). The deletion of dose contributions from fission products pl ated out on equipment casings is acceptable, since equipment surface areas are small relative to the available containment surface area. In addition, th e beta radiation emitted from plated out fission products would be absorbed in th e equipment casing and, hence, would not affect internal components;

e. The primary to secondary leak rate is 0.5% of primary containment, wt %/day is consistent with the assump tions established in Reference J.7-2;

f. The standby gas treatment system (SGT S) operates at the minimum flow of 2430 scfm based on the SGTS flow rate assumption of one reactor building air change per day;

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.E-2 g. Primary to secondary leakage is homogeneously mixed in the secondary containment atmosphere consistent with the NRC-recommended assumptions used for the calculation of doses in side primary containment (Reference J.7-2); h. No halogen plateout in the secondary containment was assumed; and

i. A spherical volume and equipment casi ng will be used which is conservative.

The beta dose to equipment is dependent on the in ternal volume size of the piece of equipment.

The beta dose is determined through the use of any energy dependent ge ometry factor and a ratio of the internal equipment volume to an infinite cloud. The beta dose contribution is excluded from the total integrated radiation doses shown on the radiation zone maps and tables for the C1E

• equipment in the reactor building. If de termination of a beta dose contribution to a C1E* component is required then a calculation to determine the internal volume size and perhaps the angle of incidence of the beta cloud to the sensitive component is performed. The results of the beta calculation are then included in the equipment qualification files for that beta sensitive equipment.

The beta calculation is determined by the airborne dose at the center of the spherical source as a function of the volume of the sphere.

The variation of beta dose rate from a typical beta energy distribution in a one-dimensional absorbing medium can be a pproximated by the formula:

D(X) = A exp (- E X) (J.E-1)

where

D(X) = dose at a point X A = constant X = position in the material E = a parameter that depends on beta energy This relationship holds approximately up to the point where all beta particles are absorbed.

This point is called the range of the beta particles. The range of a beta particle is dependent upon the energy of the beta pa rticle and is denoted r E. Both of the parameters E and r E may be determined by empirical formulas given below, based on the maximum energy of the beta particles, a nd approximately independe nt of the absorbing medium.

• Environmental qualification (EQ) of safety-related mechanical equipmen t has been eliminated from the overall Columbia Generating Station EQ program (SRM).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.E-3 E = 1 7 (E ma x)-1.14 (J.E-2) r E = (0.412

/) E n for 0.01 E 3 (J.E-3)

= (0.530E- 0.106)/ for 2.3 E 20 (J.E-4) is material density (in g/cm

3) E is energy of beta particle (in MeV) E is in cm

-1 r E is in cm n is 1.265 - .0954 LnE The dose at a given point from a single beta source is now transformed into a dose from a uniform concentration of airborne sources which extend from radius zero to radius r. K is a constant.

D(r) = K(1 - exp

( E r)) (J.E-5)

This relationship is valid for r r E. At r r E , none of the beta par ticles originating beyond r E reach the target point. Hence, at this radius, an effective infinite medium for airborne beta radiation has been reached. The dose from a volume such that r r E is equal to the dose from an infinite volume, which is denoted D. The dose as a function of volume radius is t hus found to be given by the dual relation:

DrD r r rr E EE E ()(exp())(exp())1 1 0 (J.E-6) This relation may be transformed to a function of volume by noting that V = 4 r 3/3. Since E and r E vary for each beta energy, this equation cannot be solved analytically for the case of a mixture of many beta energies - wh ich is the case at hand. However, since D for each beta energy is known (from the calcu lation of the semi-infinite source), D E(v) for each beta energy at a given volume may be determined. A ll contributions to the total dose at a given volume are then added together.

The volumes evaluated in this analysis were 10 3 , 10 4 , 10 5 , and 10 6 cm 3. Table J.E-1 summarizes the semi-infinite volum e for each beta energy group.

Table J.E-1 also indicates the beta dose reduction factor for each of the beta energy groups at the finite beta volumes of interest. A plot of the integrated 6-month dos es for these finite beta volumes is shown in Figure J.E-1. These results reflect the reduction in beta air dose from the semi-infinite C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.E-4 medium air dose to a finite volume medium air dos

e. The integrated beta infinite airborne dose for the reactor building as a function of time post-loss-of-coolant accident (LOCA) is shown in Figure J.E-2.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.E-5 Table J.E-1 Dose Rate Reduction Factors for the Post-Loss-of-Coolant Accident Beta Energy Groups at Finite Volumes

D(V) D oo for Volumes Energy Group (MeV) V E (cm 3) 10 3 cm 3 10 4 cm 3 10 5 c m 3 10 6 c m 3 0.02 - 0.10 120.0 1.0 1.0 1.0 1.0 0.10 - 0.20 4.08 x 10 5 0.486 0.763 0.960 1.0 0.20 - 0.40 8.58 x 10 6 0.260 0.478 0.755 0.955 0.40 - 0.70 1.36 x 10 8 0.127 0.254 0.468 0.744 0.70 - 1.0 1.04 x 10 9 0.0695 0.144 0.284 0.513 1.0 - 1.3 3.46 x 10 9 0.0467 0.0979 0.199 0.380 1.3 - 1.6 8.18 x 10 9 0.0348 0.0735 0.152 0.299 1.6 - 2.0 1.59 x 10 10 0.0276 0.0585 0.122 0.244 2.0 - 2.5 3.20 x 10 10 0.0215 0.0457 0.0960 0.195 2.5 - 3.0 6.47 x 10 10 0.0167 0.0356 0.0752 0.155

Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Total Integrated Beta Cloud Airborne Dose as a Function of Size 960222.64 J.E-1Finite Spherical Volume (cc)

Total Interated Dose (Rads) 3.00E+05 2.00E+5 1.00E+05 1.0E+031.0E+051.0E+06 1.0E+04 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Integrated Beta Infinite Airborne Dosefor the Reactor Building 960222.65 J.E-2Time (Hours)

Integrated Dose (Rads) 1.0E+05 1.0E+04 1.0E+03 1 10 100 1000 10,000 1.0E+06 Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-1 Attachment J.F PRIMARY CONTAINMENT ANALYSES

J.F.1 STATEMENT OF PROBLEM

It is required by NRC re gulations (NUREG-0737 a nd NUREG-0588, References J.7-5 and J.7-2) that safety-related equipm ent be qualified to withstand the radiation environment in which they are located for the 40 years of normal plant operation plus for the 6 months following a postulated design basis loss-of-coolant accident (LOCA). This attachment presents a summary of the evaluation of the radiation environment inside the primary containment of Columbia Generating Station (CGS) during normal plant operation and for the 6 months following the postulated LOCA. This attachment also calculates the maximum integrated dose due to those radiation sources.

J.F.2 BASIC APPROACH

NUREG-0737 offers two approaches for evaluating the qualification of equipment within primary containment; pressurized versus depressurized reactor coolant system, with the more conservative to be considered th e base case. Both cases assume the same source (100% noble gas, 50% halogens, and l% par ticulates of the core inventory). The difference between the two is that in the pressurized case, the source is assumed to remain in the reactor coolant system for the first 17 hr (Reference J.7-44) after the accident and then is assumed to be released into the primary containment. In the depressurized case, there is assumed to be an instantaneous release of 100% of the core noble gases and 50%

of the core halogens to the free volume of the primary containment (Reference J.7-45). It is also assumed that 50% of the core halogens and 1% of the core solids are released to the reactor coolant and the suppression pool. This causes some double counting of hal ogens and hence some c onservatism, since only 50% of the core halogens need ever be considered for release after a LOCA.

Both scenarios, the pressurized and depressuri zed were evaluated and it was determined that for CGS the depressurized case results in higher integrated doses (References J.7-46 , J.7-50 , and J.7-54). Therefore, it was considered to be the base case.

Due to the large number of C1E

• components inside primary containment, it was deemed impractical (from both scheduling and cost considerations) to calculate the integrated dose to each piece of equipment. Therefore, it was decided to calculate the worst point dose from each of the major sources in the drywell and wetwell, and then to sum these fo r a conservative estimate of the total in tegrated dose. This methodology fo r determining a wo rst-case dose for equipment in the drywell is not valid for the regi on inside the sacrificial shield wall or under

• Environmental qualification (EQ) of safety-related mechanical equipmen t has been eliminated from the overall CGS EQ program (SRM).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-2 the reactor pressure vessel. A point-specific radiation dose calculati on is required for all components present in either of these two regions.

J.F.3 DRYWELL

The integrated dose from each of the major sources to the drywell is tabulated in Table J.F-1. All values are the maximum dose for each source considered. Since the maximum dose does not occur at the same location or the same time from all sources, it is not appropriate to sum them to obtain the total integrated dose. Al l of the maximum doses calculated cannot be present for a particular acciden

t. The highest dose (7.4 x 10

7) is calculated for a depressurized reactor coolant system.

This dose is conservative since all of the source contributors summe d do not have the maximum dose at the same location. If it were determined that certain pieces of equipment could not withstand the maximum dose, a more detailed calculation would unquestionably result in an integrated dos e of lower than 7.4 x 10 7 rads. A lower bound for the more detailed calculation would be about 10 7 rads.

One major factor regarding the airborne contribution needs to be addr essed here to understand the results in Tables J.F-1 and J.F-2. Of the total airborne contribution (3.5 x 10 7 rads) slightly over 50% of it is due to photons which have an energy of less than 0.045 MeV. These photons are readily attenuated. As such, virtua lly any amount of shielding will result in a reduction by a factor of approxima tely two in the total airborne dose. Such an example is the smallest size conduit used in containm ent which has a wall thickness of 0.179 in.

This is not the only conservatism in the calculation; however, it is the most noteworthy. The following section addresses the individual contributors, assump tions, sources, models, etc., used to calculate the integrated dose.

J.F.3.1 Sources

There are six major radiation sour ces to the equipment in the dr ywell. Two of these sources are present during normal opera tion and four sources are present after a LOCA. They are Normal Ops Sources Rx Core Normal Ops Neutrons emana ting directly from the reactor core and the resultant capture gammas.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-3 Systems Normal Ops The follow i ng s y stems are the main sources of radiation during norm al plant operation:

a. Residual heat removal (RHR) system,

b. Reactor water cleanup (RWCU) system,

c. Main steam (MS) system, and

d. Reactor recirculation (RRC) system.

Systems Post-LOCA In addition to the syste m s considered under nor m al operation, (except for the MS) the following syste m s we r e a l so cons i d ered post-LOCA:

a. High-pressure core spray (HPCS),

b. Low-pressure core spray (LPCS), and

c. Reactor core isolation cooling (RCIC).

Airborne Post-LOCA Airborne radia tion from radionuclides (noble gases and halogens) which are pos tulated to be released into the primary containment atmosphere

following a LOCA.

Plateout Post-LOCA Plat eout on surfaces within containment. This consists of radioactive iodines which are initially airborne and subse quently plateout (Reference J.7-34). Wetwell Post-LOCA The radionuclides contained within the wetwell as a result of the blowdown after the accident.

J.F.3.1.1 Reactor (Normal Operation - Drywell)

There exists a general radiation field insi de primary containment due to normal plant operation. Part of this field is due to neutron leakage from the reactor core. A fraction of those neutrons penetrate the reactor vessel into the reactor cavity. Some will traverse vertically while others will penetr ate the sacrificial shie ld wall. In addition, secondary gammas will be generated from neutron interaction with materials along their path.

ANISN, a one-dimensional disc rete ordinates computer code was used to calculate the transport of these neutrons, and the gene ration of secondary ga mma rays (Reference J.7-55).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-4 The total neutron and gamma dose ra tes outside the sacrificial shield wall at core mid-plane are calculated to be

a. 5.7 rad/hr neutron, and

b. 50 rad/hr gamma

An estimate was made to determine the axial va riation of the dose rate based on geometric and material attenuation factors.

The approximate dose rate reduction factors are shown in Table J.F-3 as a function of distance from the core mid-plane.

J.F.3.1.2 Systems (Normal Operation - Drywell)

During normal operation, a radiati on field exists within containment due in part to radioactivity contained within the piping inside primary containment.

The single major source within the piping is 16 N [produced by the (n,p) 16 O 16 N reaction within the core]. The dose from other sources such as fission products, corros ion products, etc., are too small compared to 16 N to be considered.

Calculations were done to de termine the dose rate to which equipment was exposed. The results indicated that the dose rate ranged from a high of 35 ra d/hr to a low of 0.36 rad/hr.

These calculations were performed with KAP-V and QAD-BR.

They took into account the following systems: RHR, RWCU, MS, and RRC (References J.7-47 , J.7-48 , and J.7-49). The 16 N source used was 40 Ci/g (FSAR Table 11.1-4) maximum. This is the source strength of the 16N in the coolant exiting the reactor. Based on this initial source, the source strength for the pipes of the systems considered was evaluated, and the dose calculations were then performed.

J.F.3.1.3 System (Post-Loss-o f-Coolant Accident) - Drywell

The dose rate calculations for systems post-LO CA were performed using a method similar to that used for the systems under normal operation with two exceptions. Th e first was that in addition to the RHR, RWCU, and RRC systems, the HPCS, LPCS, and RCIC systems were also included. The second exception was that a different source was used (References J.7-48 and J.7-49). After a LOCA, the predominant source past the first minute or so is the assumed fission product release from the core. The 16N inventory, with a 7.1-sec half-life decays away in less than a mi nute once the (n,p) 16 O 16 N reaction stops occurring (after the reactor shuts down).

For the base case, i.e., the depressurized case, it was assumed th at 50% of the core halogens and 1% of the core solids were released a nd distributed within th e suppression pool and the reactor coolant systems (References J.7-51 , J.7-52 , and J.7-53). As noble gases were produced by the radioact ive decay of the halogens, they were discounted on the premise that C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-5 they would be released from the liquid to the gas rapidly. The released inventory is then decayed for 37 discrete time intervals out to 6 months (these are given in Table J.F-7

). An average source strength is then calculated for the 6-month period. The source strength is given in Table J.F-4. J.F.3.1.4 Airborne - Drywell

A nonmechanistic accident scenario was postulated in calculating th e airborne source. It was assumed that after 1000 days of operation at 3556 MWt (105% of core power), 100% of the noble gases and 50% of the haloge ns contained within the reac tor core are instantaneously released. After the release, no additional contribution of either noble gases or halogens is considered. Also, plateout of halo gens is considered (see Section J.F.3.1.5). The average airborne source strength is given in Table J.F-5.

The above source is calculated via the ORIGEN2 computer code. After the source strength was determined, the dose rate was calculated using the QAD-CG computer code. Details of the model and the calculation are discussed in Section J.F.5. The value for the airborne contribution presented in Table J.F-1 represents the dose rate at a point within the drywell which is predominantly surrounded by air. This point was chos en because of the absence of structural steel, piping, etc., surrounding the dose point. This would result in an upper limit dose rate which could be expected to occur in the drywell.

The effect of the shielding a fforded by the structural steel, piping, etc. (i.e., "shadow shielding"), within containment was considered. Advantage was taken of "shadow shielding" when considering the contribution of the more distant airborne sources (References J.7-48 and J.7-49). This significantly reduces the dose ra te compared to the case where "shadow shielding" is not employed. See Section J.F.5 for modeling of "s hadow shielding."

J.F.3.1.5 Plateout - Drywell

The basis for determining the plat eout source is 50% i odine inventory releas ed after 1000 days irradiation at a power level of 3556 MWt. However, the plateout source is only those iodines which are removed from the airborne source and assumed to plateout on the surfaces within containment. As such, plate out removes sources from the airborne source, and this was accounted for in the calculations.

It was assumed, however, that the noble gases generated by the decay of the plated out halogens (I Xe and BR KR) are instantaneously released and are mixed within the free volume of the drywell. In this manner, both the airborne and plateout sources are determined with no "double counting" of nuclides. The plateout source is given in Table J.F-6. When the halogens are initially re leased, not all of them are cons idered available to plateout. Of the halogens released, 2.0% are in the fo rm of organic compounds, and 2.5% are in the form of particulates (Reference J.7-2); and both of these forms are assumed not to plateout.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-6 The remaining 95.5% are considered to be in an elemental state of which one-two hundredth remain airborne and the rest plateout. Therefore, no more than 95% of the released halogens can ever plateout. The plateout was assumed to occur with an effectiv e deposition velocity of 0.05 cm/sec. This translated into an effective half-life of 1.01 hr-l (References J.7-34 and J.7-45). Given this half-life, the limit of a reducti on of a factor of 200 is attained in slightly over 5 hr. After that time, the percentage of plated out hal ogens remains c onstant at 95%.

The dose calculations were performed with the computer code QAD-CG, incorporating a model similar to that used for the airborne dose. See Section J.F.5 for discussion of model and calculations.

Initial calculations were perfor med with the total plateout be ing distributed over: (1) the drywell lateral surface, top, a nd bottom; (2) inner, outer and top surfaces of the sacrificial shield wall; and (3) heat reflector of pressure vessel surface.

Given this distribution area, the maximum dose rate ca lculated was 7.04 x l0 3 rad/hr. However, when the remaining surface areas within containment (i.e., equipment, piping, structural steel, etc.) were considered, the area over which the source would be plated out increased sevenfold. A counter-balancing effect to this reduction in plated out concentration was that the source would be more universally distributed ar ound any given receiver. It was estimated that the net effect would reduce the calculated maximum dose rate by a factor of approximately three.

It is noted that the energy spectrum for the plate out source is significantly harder than that of the airborne. As such, the comments in Section J.F.3 regarding low energy photons are not completely applicable.

J.F.3.1.6 Wetwell - Drywell

The wetwell was also considered as a source to the drywell.

However, due to distance, self-attenuation, and the availa ble shielding from the 2-ft-thick diaphragm floor, its contribution to the dr ywell was negligible.

It was assumed that 50% of th e halogens and 1% of the particulates from the core were entrained in the water in the suppression pool.

This is the same sour ce used for the systems post-LOCA. The air space volume above the suppr ession pool was assume d to have the same volumetric source strength as the drywell air space. These are conservative premises since only a total of 50% of the core halogens are assumed to be released after the accident.

J.F.4 WETWELL

The results for the wetwell are given in Table J.F-2. Doses were calculat ed for detector points both within the suppression pool as well as in the free volume above it using QAD-CG, applying the same modeli ng techniques as was used in the drywell.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-7 With regard to the airborne contribution, the volumetric source strength is the same as the drywell airborne source and the comments in Section J.F.3.1 regarding the low-energy photons applicability to the wetwell.

There does exist some double counting of nuclides in the wetw ell analysis. The airborne source is 100% noble gases and 50% halogens, released into the containment (wetwell and drywell) free volume. For the suppressi on pools the source is 50% halogens and 1% particulates. Since only 50% of the total core halogens are assumed to be released after an accident, they are double counted (the effect is small, however, becau se of the shielding offered by the suppression pool wa ter.) Another conservatism in the airborne source in the wetwell is that, since the path for the wetwell airborne sources is via the downcomers and then up through the suppression pool, so me halogens are expected to be entrained in the water during this transfer (this was not considered in the calculation.) The result would have been a smaller airborne source and in turn a smaller dose.

J.F.4.1 Sources

There are three sources of radia tion to the equipment in the wetwell, all of which are present only after an accident.

a. Airborne

The airborne source is pr esent as a result of the initial blowdown into the suppression pool via the downcomers,

b. Plateout

Plateout of halogens onto the surfaces in the wetwell (i.e., containment, downcomers, etc.), and

c. Suppression Pool

The radionuclides contained within th e suppression pool as a result of the blowdown after the accident.

J.F.4.1.1 Airborne - Wetwell

The airborne source, on a specific volume basis, is equal to the airborne source in the drywell (i.e., 100% noble gas and 50% ha logen released into the total primary containment immediately following a LOCA). However, the amount of "shadow shielding" within the wetwell is much less than in the drywell. Hence, the contribution from sources further away is greater. This factor accounts fo r the increased dose rate in th e wetwell with respect to the C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-8 drywell (due to airborne sour ces). Dose calculations in the wetwell were done in similar manner as for the drywell (i.e., using QAD-CG).

J.F.4.1.2 Plateout - Wetwell

As in the drywell, the source of the plateout in the wetwell is th e halogens. However, the area available for plateout is smaller in the wetwell than the drywell. This results in a dose rate in the wetwell slightly more than double that in the drywell.

J.F.4.1.3 Suppressi on Pool - Wetwell The source in the suppression pool was assumed to be 50% of the halogens and 1% of the particulates instantaneou sly released from the co re into the pool and the reactor coolant system.

It is further assumed that as noble gases are produced by the decay of the halogens (I Xe and B K), they "bubble out" of the pool , hence they are not consid ered a source term. Dose rates both in the suppression pool as well as in the wetwell free volume were calculated using QAD-CG.

J.F.5 QAD-CG MODEL The QAD-CG computer program was used to calculate dose rates for both the airborne source as well as the plateout. In both cases, i.e., airborne and plateout, similar modeling techniques were used. This section defines the modeling used in both calc ulations (with only the drywell used for illustrative purposes).

The QAD-CG computer code makes use of a geometry package, which allows the user to model a calculation with the use of predetermine d geometric bodies. The user defines a set of geometric "bodies" (boxes, truncated cones, spheres, cylinders, etc.) and using these "bodies," the user defines "zones" by intersection or forming unions of them to build the shapes desired in a manner analogous to "intersections" and "unions" when one deals with sets. The model is done three-dimensionally ther eby allowing the user cons iderable fl exibility.

These "zones" are then what constitute the computer model.

The parts of "bodies" that are not used have no effect on the model.

As an example, a dumbbell could be defined as the union of thr ee "bodies": two spheres and a long, thin cylinder between them (see Figure J.F-1). Likewise, a hemisphere could be formed by intersection of a sphere with a box (see Figure J.F-1

). In this manner, a complex model can be defined.

In our case, the basic model was defined as a truncated cone (approximating the containment shell) and two cylinders (approximating the sacrificial shield wall and the reactor vessel).

Figure J.F-2 illustrates this in a s ectional view. The free volume of the drywell was compartmentalized into cubes, 7 ft on a side. These c ubes were formed by intersection of a C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-9 series of tall rectangles , which are 7 ft on a side in cross-s ection, with cylinders at 7-ft high intervals. Each 7-ft high cyli nder constitutes the eleva tional boundaries of what is referred to as a "layer" below. Combining these "bodies" appropriately one winds up with a truncated cone (containment) with two cylinders (i.e., sacrificial wall and reactor vessel) and the remainder of the volume forced with cubes (except on the boundary of the cone or cylinders).

Figure J.F-3 illustrates this model, while Figure J.F-4 illustrates how the layer from el. 513 ft 6 in. to el. 520 ft 6 in. is modeled.

All major structures, pipes (6 in. and above), hangers, etc., w ithin the drywell were then located, and the mass of steel in each cubicle determined. These were tr anslated into average densities such that each cube ha d an average density assigned to it. These are illustrated on Figures J.F-5 to J.F-9 for the lower five layers; for the pur pose of clarity, the densities shown are much cruder than the 41 used in the code. In those cubicles which are noted to have zero density, the density of air was assumed.

In Figures J.F-7 to J.F-9 a large void (air only) exists in the southwest (fourth) quadrant in layers 3 and 4. It was in this region that the airborne dose rate was calculated. This region provides us with a volume which is large enoug h so that the "shadow shielding" (smearing discrete shielding within a cubi cle into an average density in the cubicl e) beyond its boundary is justified.

Several runs were made using this model with various sour ce volumes. Three runs were made placing the source terms within the elevational boundaries of layers 3, 4, a nd 5, respectively, and another run was ma de by placing the source from the lower elevational boundary of layer 1 up to the upper elevational boundary of layer 2.

It was noted that >95% of the total dose contribution from these five layers came equally from layers 3 and 4. In other words, the further away the source layer, the smaller the contribution. Also, shadowing shielding in layers 1 and 2 provided suffici ent attenuation as to make the contribution to the total dose negligible. The same is true al so for all layers above layer 5.

Plateout was calculated in a similar ma nner, increasing the source until successive contributions became negligible. For the plateout, the dose point was taken near the sacrificial shield wall. Other points were also considered, but the dose rate near the sacr ificial shield wall was found to be the maximum.

Again, the dose point was taken between layers 3 and 4 to maximize the dose rate.

J.F.6 CODES

J.F.6.1 FSPROD

FSPROD is a computer program which calculat es the inventory and activity of radioactive fission products, produced fr om the thermal fission of 235 U, as a function of fission rate and decay time after fission. The program is used in establishing the gross and specific gamma and C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-10 beta activity of those fission products. The calculation inco rporates 123 fission product nuclides and is based on Perkins and King data.

J.F.6.2 ORIGEN2

ORIGEN2 is a point depletion and decay computer code for use in simulating nuclear fuel cycles and calculating the nuc lide composition of materials c ontained therein. The code represents a revisi on and update of the original ORIGEN computer code. The general function of the ORIGEN2 computer code is to calculate the nuclides presen t in various nuclear materials by determining the buildup and de pletion of nuclides during irra diation and decay. The code can also account for reprocessing (i.e., chemical separation) and continuous feed, removal, and accumulation of nuclear materials.

J.F.6.3 QAD-BR

QAD-BR is a point kernal comput er code designed to evaluate gamma penetrati on of various shield configurations. It is a modification of QAD-P5A; i.e., it has no capability for neutron calculations. The program provides an estimat e of the uncollided and collided gamma flux, dose rate, energy deposition, and other quantities which re sult from a point-by-point representation of volume-distr ibution source of radiation.

J.F.6.4 QAD-CG

QAD-CG is also a modification of the QAD-P5A co mputer program. It is similar to QAD-BR in application with the major difference being in the geometry descri ption. QAD-CG makes use of a combinatorial geometry package originally developed for MORSE. It is one of the more versatile geometry packages to be available in the QAD family of computer codes.

J.F.6.5 KAP-V

KAP-V is a hybrid of the QAD com puter code. Analytically, it is identical to QAD as it is a point kernal code. The major differences are changes in input allowing more flexibility in running successive cases.

It also has internal libraries for attenuation and buildup data which can be used by default for convenience.

J.F.6.6 ANISN ANISN is a one-dimensional Sn transport code with anisotropic scattering. It allows for the solution of the transport equation for neutrons a nd photons using the disc rete ordinate method.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDCN-02-000 J.F-11 Table J.F-1 a Integrated Dose in Drywell

Source Maximum Average Dose Rate b (rad/hr)

Exposure Time Dose b (rad) Reactor 5.6 x 10 1 32 years c 1.6 x 1 0 7 Systems - normal 3.5 x 10 1 32 years c 9.9 x 1 0 6 Systems - LOCA

--- 6 months 3.2 x 1 0 6 Airborne --- 6 months 3.7 x 10 7 Plateout --- 6 months 1.0 x 10 7 Suppression pool

--- 6 months <4.5 x 1 0 4 a Not valid for regions inside the sacrificia l shield wall or under the reactor pressure vessel (a point specific radiation calcula tion is required for components in these two regions).

b Maximum dose rate from i ndividual contributors does not necessarily occur at the same location or for the same accident.

c (40-year plant life) x (0.8) av ailability to account for down time.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-12 Table J.F-2 Integrated Dose in Wetwell

Source Maximum Average Dose Rate a (rad/hr) Exposure Time Dose a (rad) Dose above suppression pool Airborne 1.8 x 1 0 4 6 months 8.2 x 10 7 Suppression pool 2.0 x 1 0 2 6 months 9.1 x 10 5 Plateout 2.7 x 1 0 3 6 months 1.2 x 10 7 Dose within suppression pool Airborne 2.9 x 1 0 2 6 months 1.4 x 10 6 Suppression pool 5.5 x 1 0 2 6 months 2.5 x 10 6 a Maximum dose rate from i ndividual contributors does not necessarily occur at the same location.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-13 Table J.F-3 Approximate Dose Ra te Reduction Factor Versus Distance from Core Mid-Plane for Reactor Integrated Dose Distance (ft) Reduction Factor 0 1.0 5 0.5 10 0.02 15 1 x 10-5 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-14 Table J.F-4 Suppression Pool and System (Loss-of-Coolant Accident) Liquid Source Terms 0-6 Month Average After Loss-of-Coolant Accident MeV MeV/sec MeV/c m 3-se c a 0.015 1.8E+14 4.4E+4 0.025 3.5E+14 8.5E+4 0.0375 4.8E+14 1.2E+5

0.0575 1.4E+14 3.5E+4

0.085 5.8E+14 1.4E+5 0.125 2.2E+15 5.3E+5 0.225 4.0E+15 9.6E+5

0.375 4.8E+16 1.2E+7

0.575 5.5E+16 1.3E+7

0.85 7.2E+16 1.7E+7 1.25 1.5E+16 3.7E+6 1.75 1.8E+16 4.3E+6

2.25 2.1E+15 5.1E+5

2.75 9.1E+14 2.2E+5

3.5 1.1E+14 2.6E+4

5.0 7.4E+13 1.8E+4 a Volume considered was that of the suppr e ssion pool plus that of the reactor coolant system.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-15 Table J.F-5 Airborne Source Terms 0-6 Month Average After Loss-Of-Coolant Accident MeV MeV/sec MeV/c m 3-se c a 0.015 2.5E+14 2.5E+4 0.025 2.1E+14 2.1E+4 0.0375 6.9E+15 7.1E+5 0.0575 3.0E+13 3.0E+3 0.085 1.4E+16 1.4E+6 0.125 3.7E+13 3.8E+3 0.225 4.4E+15 4.5E+5 0.375 3.0E+15 3.1E+5 0.575 4.0E+15 4.1E+5 0.85 3.2E+15 3.2E+5 1.25 3.4E+15 3.5E+5 1.75 2.6E+15 2.7E+5 2.25 3.9E+15 4.0E+5 2.75 6.7E+14 6.9E+4 3.5 2.1E+14 2.2E+4

5.0 9.5E+13 9.7E+3 a Volume considered was total; i.

e., drywell plus wetwell free volume.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-16 Table J.F-6 Drywell Plateout Source Terms 0-6 Month Average After Loss-Of-Coolant Accident MeV MeV/sec MeV/c m 3-se c a 0.015 3.6E+13 6.3E+5 0.025 1.8E+14 3.2E+6 0.0375 5.4E+13 9.5E+5 0.0575 2.0E+13 3.6E+5 0.085 3.2E+14 5.7E+6 0.125 1.9E+13 3.4E+5 0.225 2.8E+15 4.9E+7 0.375 4.4E+16 7.7E+8 0.575 2.4E+16 4.2E+8 0.85 7.6E+15 1.3E+8 1.25 9.6E+15 1.7E+8 1.75 3.4E+15 5.9E+7 2.25 5.2E+14 9.2E+6 2.75 7.3E+12 1.3E+5 3.5 1.3E+13 2.3E+5 5.0 2.9E+11 5.1E+3 a These values should be reduced by a factor of seven when all structural, component and equipment surfaces in containment are considered.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-17 Table J.F-7 Time Mesh Spacing Used in Source Calculations (Minutes) 0 640 28800 20 800 36000 40 960 43200

60 1120 57600

80 1280 72000 100 1440 86400

120 2160 108000

180 2880 129600

240 3600 151200

300 4320 172800

360 5040 216000

420 5740 259200

480 14400

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.F-18 Table J.F-8 Source Energy Group Structure Lower Boundary Upper Boundary Average Energy (MeV) (MeV)

(MeV) 0.00 0.02 0.015

0.02 0.03 0.025

0.03 0.045 0.0375

0.045 0.07 0.0575

0.07 0.10 0.085

0.10 0.15 0.125

0.15 0.30 0.225

0.30 0.45 0.375

0.45 0.70 0.575

0.70 1.0 0.85

1.0 1.5 1.25

1.5 2.0 1.75

2.0 2.5 2.25

2.5 3.0 2.75

3.0 4.0 3.5 4.0 6.0 5.0

Geometry Examples 970187.69 J.F-1 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Columbia Generating StationFinal Safety Analysis Report Basic QAD-CG Drywell Model 970187.70 J.F-2 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Columbia Generating StationFinal Safety Analysis Report Isometric of Drywell Model 970187.71 J.F-3 Figure Amendment 53 November 1998 Form No. 960690 Draw. No.Rev.Columbia Generating Station Final Safety Analysis Report Isometric of El. 513 ft 6 in. to 520 ft 6 in.

970187.72 J.F-4 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Columbia Generating StationFinal Safety Analysis Report Plan at El. 499 ft 6 in.

970187.73 J.F-5 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Illustrative Only - Code Used 41 Density Groups Air<0.10.1 - 0.50.5 - 1.0 1.0 gm cc Columbia Generating StationFinal Safety Analysis Report Plan at El. 506 ft 6 in.

970187.74 J.F-6 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Illustrative Only - Code Used 41 Density Groups Air<0.10.1 - 0.50.5 - 1.0 1.0 gm cc Columbia Generating StationFinal Safety Analysis Report Plan at El. 513 ft 6 in.

970187.75 J.F-7 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Illustrative Only - Code Used 41 Density Groups Air<0.10.1 - 0.50.5 - 1.0 1.0 gm cc Columbia Generating StationFinal Safety Analysis Report Plan at El. 520 ft 6 in.

970187.76 J.F-8 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Illustrative Only - Code Used 41 Density Groups Air<0.10.1 - 0.50.5 - 1.0 1.0 gm cc Columbia Generating StationFinal Safety Analysis Report Plan at El. 527 ft 6 in.

970187.77 J.F-9 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Illustrative Only - Code Used 41 Density Groups Air<0.10.1 - 0.50.5 - 1.0 1.0 gm cc Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.G-1 Attachment J.G BETA DOSE CONTRIBUTION IN PRIMARY CONTAINMENT

The source volume used for the beta dose analysis in primary containment is a sphere surrounded by a shell of sufficien t thickness to stop all outside be ta particles from entering the source volume. This spherical source volume is conservative for any generalized source volume shape (the dose at the cente r of the sphere is higher than the dose at any point of any generalized source of equal total volume).

The assumptions used in the analysis are as follows:

a. Atmosphere inside the equipment casi ng is identical to the atmosphere in primary containment. This is conservative because there will actually be some delay in transport of the gaseous fission products into the equipment;

b. The initial beta source term used was 100% of core nobl e gases and 50% of core halogens (References J.7-2 and J.7-34);

c. Daughter products of the airborne noble gases and halogens are included in the calculation of the airborne dose which is conservative and was required by the use of ORIGEN2 as a source code (Reference J.7-8);

d. Plateout of halogens inside primary containment was utilized as allowed per Reference J.7-34. The dose contribution of fission products plated out on equipment casings was neglected. The deletion of dose contributions from fission products plated out on equipment casings is acceptable, since equipment surface areas are small relative to the available containment surface area. In addition, the betas emitted from plated out fission pr oducts would be absorbed in the equipment casing and, hence, w ould not affect internal components;

e. No primary to secondary containment leakage is assumed si nce it maximizes the beta source concentration in primary containment;

f. Activity is assumed to be uniformly distributed thro ughout the containment free volume which is reasonable, considering the mixing effects of the loss-of-coolant accident (LOCA) blowdown a nd the operation of the drywell fan coolers; and

g. A spherical volume re presenting the equipment casing will be used.

The beta dose to equipment is de pendent on the internal volume size of the piece of equipment. The beta dose is determined th rough the use of an energy dependent geometry factor and a C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.G-2 ratio of the internal equipment volume to an infinite cloud. The be ta dose contribution is excluded from the worst case total integrated gamma doses of primary containment shown in Section J.6.1 and Tables J.F-1 and J.F-2. The beta dose contributi on is also excluded from the value, pump, and fan tables for C1E/S RM equipment in a primary containment environment.

The discussion and development of beta dose rate variation due to beta energy distribution in a one-dimensional absorbing medium is also valid for primary containment analysis.

Thus, the dose as a function of volume radius is given by the dual relation:

Dr r r rr E EE E ()[exp()][exp()] D 1 1 0 This relation may be transformed to a function of volume by noting that V = 4 r 3/3.

Since E and r E vary for each beta energy, this equation cannot be solved analytically from the case of a mixture of many beta energies, whic h is the case at hand.

However, since D for each beta energy is known (from the calculation of the semi-infinite source), D E(v) for each beta energy at a given volume may be determined. All contributions to the total dose at a given volume are then added together.

The volumes evaluated in this analysis were 10 3 , 10 4 , 10 5 , and 10 6 cm 3. Table J.G-1 summarizes the semi-infinite volum e for each beta energy group.

Table J.G-1 also indicates the beta dose reduction factor for each of the beta energy groups at the fi nite beta volumes of interest. A plot of the integrated post-LOCA doses for these finite beta volumes is shown in Figure J.G-1. These results reflect the reduction in beta air dose from the semi-infinite medium air dose to a finite volume air dose.

The integrated beta infinite airborne dose for the primary co ntainment as a function of time post-LOCA is shown in Figure J.G-2.

The absorbed beta dose within a physical target is not always equal to the beta dose at a mathematical point in air at the surface of that piece of equipment. The beta ionization energy (dose) deposited on the surface of a solid object is distributed in a thin surface layer to a depth equal to the beta range in the material. The relative material pe netration of the different beta energy groups is used to provide a total integrat ed LOCA dose as a functio n of material depth.

Finite volume beta dose reduction factors were determined for each of the 10 beta energy groups. These factors are used to provide total integrated LOCA dose as a function of material penetration to reduce volume exposure.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.G-3 Thus, the integrated dose values (Figure J.G-1) can be used as the abso rbed material dose with a standard order of magnitude for reduction for material beyond 0.030-in.

thickness or a dose reduction versus thickness based on the range of beta penetration within the material can be calculated.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.G-5 Table J.G-1 Dose Rate Reduction Factors for the Post-Loss-of-Coolant Accident

Beta Energy Groups at Finite Volumes D(V) D for Volumes Energy Group (MeV)

V E (cm 3) 10 3 cm 3 10 4 cm 3 10 5 c m 3 10 6 c m 3 0.02 - 0.10 120.0 1.0 1.0 1.0 1.0 0.10 - 0.20 4.08 x 10 5 0.486 0.763 0.960 1.0 0.20 - 0.40 8.58 x 10 6 0.260 0.478 0.755 0.955 0.40 - 0.70 1.36 x 10 8 0.127 0.254 0.468 0.744 0.70 - 1.0 1.04 x 1 0 9 0.0695 0.144 0.284 0.513 1.0 - 1.3 3.46 x 1 0 9 0.0467 0.0979 0.199 0.380 1.3 - 1.6 8.18 x 1 0 9 0.0348 0.0735 0.152 0.299 1.6 - 2.0 1.59 x 1 0 10 0.0276 0.0585 0.122 0.244 2.0 - 2.5 3.20 x 1 0 10 0.0215 0.0457 0.0960 0.195 2.5 - 3.0 6.47 x 1 0 10 0.0167 0.0356 0.0752 0.155 Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Total Integrated Beta Cloud Airborne Dose in Primary Containment as a Function of Size 960222.58 J.G-1Finite Spherical Volume (cc)

Total Integrated Dose (Rads) 1.0E+9 1.0E+8 1.0E+71E+01E+11E+21E+31E+41E+51E+61E+71E+8 Columbia Generating StationFinal Safety Analysis Report Figure Amendment 53November 1998 Form No. 960690Draw. No.Rev.Integrated Beta Infinite Airborne Dose for Primary Containment 960222.60 J.G-2Time (Hours)

Integrated Dose (Rads)1E+9 1E+8 1E+7 1 10 100 1000 10,000 Columbia Generating StationFinal Safety Analysis Report C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-1 Attachment J.H VITAL AREAS AND ACCESS ROUTES ANALYZED FOR POST-LOSS-OF-COOLANT-ACCIDENT OPERATIONS

This attachment represents the methodology and assumptions used to determine the integrated dose to equipment and personnel for vital areas and access routes outside the reactor building during post-loss-of-coolant accident (LOCA) operations. The source term is the reactor building elevated vent with gaseous effluent s being filtered by the standby gas treatment system (SGTS) prior to discharge to the atmosphere.

J.H.1 SOURCE OF RADIOACTIVITY TO THE REACTOR BUILDING ELEVATED VENT Two contributions were considered as the source of the radioactivity to the reactor building elevated vent:

a. Leakage from the drywe ll to the reactor building and discharged via the SGTS to the reactor building elevated vent was assumed at a rate of 0.5%/day = 2.1E-4/hr, and

b. Leakage from the assumed leaks on the main steam isolation valves (MSIVs) in the main steam tunnel was assumed at a rate of 0.17%/day =

7.1E-5/hr (Re f erence J.7-56)

Thus, the total leakage rate of activity from the primar y system is assumed to be

0.67%/day = 2.8E-4/hr.

J.H.1.1 Reactor Building Air Discharge Rate

All radioactivity considered outside the reactor building is assumed to di scharge via the reactor building elevated vent.

The removal rate of the reactor building ventilation can be determined as follows:

Removal rate = SGTS discharge rate Reactor bui l d ing volume

SGTS discharge flow = 2430 ft 3/minute C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-2 Reactor bui l d ing volume = 3.5E+6 f t 3 Thus, the removal rate is as follows representing one volume change per day:

Removalrate(/min)(min/)

.243060356 3 3 fthrEft Removal rate = 4.2E-2/hr

This removal rate was used in the determination of radiation levels outside the reactor building.

J.H.2 POSTACCIDENT DE SIGN DOSE (PADD)

A small computer program (PADD) was writte n to complete the calculations for the 18 nuclides over various time peri ods and sum the results. The equation used to determine the dose is as follows:

Dose() = DF(j) (Q TF Q + Q 3600)1 jrad j**1 2 (J.H-1) where Dose ji = Rads from jth nuclide for the ith time period.

DF j = Gamma dose factors for semi-infinite cloud RadmCihr**3 for jth nuclide.

/Q 1 = sec/m 3 for gaseous releases from the reactor building vent to the atmosphere for the ith time period.

RF = Removal fraction of activity via the standby gas treatment.

TF = 0.01 for particulates and i odines (99% efficiency or RF).

TF = 1.0 for noble gases (FSAR Section 6.5).

Q1 j = Integrated activity of jth nuclide over ith time period that was released via leaks in the MSIVs (curies/hour).

Q2 j = Integrated activity of jth nuclide over the ith time period that was released via leakage from the pr imary to secondary containment (curies/hour).

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-3 3600 = Conversion from hours to seconds.

J.H.2.1 Assumptions Used in /Q Calculation Methodology The following equation from "Meteor o logy and Atomic Energy" (Reference J.7-31) was used to determine the

/Q values shown in Table J.H-

1. D i lution = 2.22(M) (3.16 0.1 S (A e x)) V V/2 m e a n ex 2 (J.H-2)

= F B (building wake factor)

M = 1 if intake and e xhaust same elevation

M = 2 if intake and exha ust separated by one floor

M = 4 if intake is in building wake cavity

S = shortest intake exhaust arc length

Aex = exhaust area

Vmean = mean approach flow

V ex = mean exhaust flow

The intake was assumed to be for category F weather conditions with a Vmean = 1 meter/sec.

Then /Q = 1 F R BR F B = building wake factor R R = release rate from reactor building vent (m 3/sec)

Concentration in reactor vent

C V = Q/R R where Q = curies/sec released

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-4 Concentrat i on at intake C I = C V/F B C I also = Q (/Q) Therefore:

i VBBR C = C F = Q ( /Q) = (Q F R) (/Q) = 1 (F )(R ) = (D ).BR F tot a l dil u tion f a c t or An F class stability was assum e d for atmosphere conditions a nd 5% meteorology was then applied for time periods from 0 to 180 days.

The dilution factors decrease by the following ratios for the time peri o d s indicate

d.

Time (hr) 0-2 2-8 8-24 24-96 96-4320 Ratio 1.0 0.35 0.04 0.02 0.01 The dilution factors were multiplied by the 5% meteorology ratios to determine the actual

/Q values used in these co m putations as presented in Table J.H-

1.

J.H.2.2 Integrated Activity Equ a tions Used in this Analysis

The time dependent act i v ity of each nuclide being relea s ed f r om the MSIV was anal y zed as follows:

d A 1 0 dt = PA e (- + .0067 24)t o (J.H-3)

where

P = Fractional l eak from MSIV per hour (7.1 E-5/hr)

A o = Initial activity of jth nuclide in primary containment at t = 0 hr

Thus, the activity concentration over a time period of t 1 to t 2 is QPA Et o t t 1 1 2284 e (.) or C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-5 Q = PA(+2.8E-4) -(+2.8E-4) t1 e --( + 2.8E-4) t2 e 1 o (J.H-4) The integrated activity concentration from the primary to secondary containment leakage, Q2, was calculated as follows:

2 12 2 2 dA dt = KA - L C - A (J.H-5)

where K = Fractional leak rate from primary containment 0005 242141..hrEhr 0 A = Activity in primary containment

= o Ae -( + 0.0067 24)t xp A 1 = Initial activity (Ci) at t = 0

0.0067 = Leakage removal rate from primary containment per hour 24

= 2.8E-4 hr

-1 L2 = Discharge rate from reactor building vent via standby gas treatment = 2430 ft 3/min (60 min/hr)

= 1.46E+5 ft 3/hr C2 = Activity concentrati on in secondary containment

A2 = Curies in secondary containment

V2 = Volume in secondary containment

Rearranging dA dt = kAe -( + 2.8E-4)t - -

A2 2 2 2 2 o xp L V A C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-6 or(J.H-5A) dA = kA e -t - ( + ) A 2 o 2 2 2 1 F L V dA = o kA e -t - A2 dt 2 1 2 F F where 2 1 F = + 2 V F = ( + 2.8E-4) L 2 A2 = kA - F 12 A2 A2 + F = r(t)22 A r(t) = kA e o-F t 1 (J.H-6) solving A2 = e (kA F-F) e + C-F t o 21 (F-F)t 21 2 at t=o, A2 = o (J.H-7) c = .005 A o0 Thus, 2o-t-(.0042)t A(t) = .005 A e (1-e )0 Q2 = A(t)V2 2 L or 2 (J.H-8) Q2 = 1.45E+5 ft/hr3.5E+6 ft .005 A e (1-e )3 3 o-j t-C2t 0 where C2 = 0.042 (J.H-9)

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-7 thus, Q 2 = 2.1 1 E- 4 A e (1-e)o-j t -C 2 t To determine the integrated concent r ation:

Q 2 (t) = 2.1 E- A o 4 1 2 tt-t -( + C 2)t e -e dt (J.H-10)

Solving, Q2 = 2.1E-4 A e -e (e -e ) + C2 o-t-t-C2 t-Ct 12 1 2 2 (J.H-11)

The values of Q1 and Q2 are substituted in for each nuclide and each time period. Then using equation (J.H-1), the dose commitment for each nuclide and each time period may be calculated. These results are presented in Section J.6.3.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 J.H-9 Table J.H-1 Post-Loss-of-Coolant Accident /Q Value s a Used for Calculations of Integrated Doses O u tside the Reactor Building Time (hr) Area 0-2 2-8 8-24 24-96 96-4320 (180 days)

Security center 2.lE-4 b 7.35E-5 8.4E-6 4.2E-6 2.1E-6 Auxiliary security center 1.2E-4 4.2E-5 4.8E-6 2.4E-6 1.2E-6 Sample analysis area (end of cycle) 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Nitrogen supply to accumulators 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Standby service water pump valves 1.2E-4 4.2E-5 4.8E-6 2.4E-6 1.2E-6 Remote shutdown room 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Switchgear room 1 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Switchgear room 2 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Radwaste control room 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Battery racks Direct current battery charger

Motor control center

2.6E-4

9.1E-5

1.0E-5

5.0E-6

2.5E-6 Three motor control centers/

Three switchgears

Direct current battery charger and rack

2.6E-4

9.1E-5

1.0E-5

5.0E-6

2.5E-6 Diesel oil tanks 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Solid radwaste control panel 2.6E-4 9.1E-5 1.0E-5 5.0E-6 2.5E-6 Sample of elevated release duct 8.0E-4 2.8E-4 3.2E-5 1.6E-5 8.0E-6 The standby service water pump valves are approximat e ly 700 ft from the release point. This distance

is too great to calculate a dilution based solely on a building wake factor. However, the conservative

assumption will be made that the dilution at the valves is the same as at the auxiliary guard house which

is only 420 ft.

a These values are based on an MSIV leak rate of 0.22%/day not the 0.17%/day previously listed. The results are acceptable and conservative for a leak rate of 0.17%/day.

b Read as 2.1 x 1 0-4 etc.