ML15355A511

Final Safety Analysis Report, Amendment 63, Chapter 12 - Radiation Protection

ML15355A511
Person / Time
Site:	Columbia
Issue date:	12/21/2015
From:	Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15356A148	List: GO2-15-159, Final Safety Analysis Report, Amendment 63, Appendix a - Supplement Aging Management Programs and Activities Credited for Columbia License Renewal ML15355A480 ML15355A481 ML15355A486 ML15355A487 ML15355A488 ML15355A489 ML15355A490 ML15355A496 ML15355A497 ML15355A498 ML15355A499 ML15355A500 ML15355A507 ML15355A508 ML15355A509 ML15355A510 ML15355A511 ML15355A512 ML15355A515
References
GO2-15-159
Download: ML15355A511 (77)

Text

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS Section Page LDCN-13-039 12-i 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA).............. 12.1-1 12.1.1 POLICY CONSIDERATIONS...................................................... 12.1-1 12.1.2 DESIGN CONSIDERATIONS...................................................... 12.1-4 12.1.3 OPERATIONAL CONSIDERATIONS............................................ 12.1-8 12.1.3.1 Procedures and Methods of Operation........................................... 12.1-8 12.1.3.2 Design Changes for ALARA Exposures......................................... 12.1-9 12.1.3.3 Operational Information............................................................ 12.1-10 12.2 RADIATION SOURCES................................................................ 12.2-1 12.2.1 CONTAINED SOURCES............................................................ 12.2-1 12.2.1.1 General................................................................................ 12.2-1 12.2.1.2 Reactor and Turbine Building..................................................... 12.2-1 12.2.1.2.1 Reactor Core Radiation Sources................................................ 12.2-1 12.2.1.2.2 Process System Radiation Sources............................................. 12.2-2 12.2.1.2.2.1 Introduction...................................................................... 12.2-2 12.2.1.2.2.2 Recirculation System Sources................................................ 12.2-2 12.2.1.2.2.3 Reactor Water Cleanup System Sources.................................... 12.2-3 12.2.1.2.2.4 Reactor Core Isolation Cooling System Source........................... 12.2-3 12.2.1.2.2.5 Residual Heat Removal System Sources.................................... 12.2-3 12.2.1.2.2.6 Fuel Pool Cooling and Cleanup and System Sources..................... 12.2-4 12.2.1.2.2.7 Main Steam and Reactor Feedwater Systems Sources.................... 12.2-5 12.2.1.2.2.8 Offgas Sources in the Turbine Generator Building....................... 12.2-5 12.2.1.2.2.9 Traveling In-Core Probe System Sources.................................. 12.2-6 12.2.1.2.2.10 Sources Resulting From Crud Buildup.................................... 12.2-6 12.2.1.3 Radwaste Building................................................................... 12.2-6 12.2.1.4 Byproduct, Source, and Special Nuclear Materials............................ 12.2-6 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES........................ 12.2-6 12.2.2.1 General................................................................................ 12.2-6 12.2.2.2 Model for Computing the Airborne Radionuclide Concentration in a Plant Area.......................................................................... 12.2-7 12.2.2.3 Sources of Airborne Radioactivity................................................ 12.2-8 12.2.2.3.1 Effect of Leakage from Process Equipment in Radioactive Systems..... 12.2-8 12.2.2.3.2 Effect of Sumps, Drains, Tank and Filter Demineralizer Vents.......... 12.2-10 12.2.2.3.3 Effect of Relief Valve Exhaust.................................................. 12.2-11

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS (Continued)

Section Page LDCN-05-002 12-ii 12.2.2.3.4 Effect of Removing Reactor Pressure Vessel Head and Associated Internals............................................................................. 12.2-13 12.2.2.3.5 Effect of Sampling................................................................ 12.2-13 12.2.2.3.6 Effect of Spent Fuel Movement................................................. 12.2-13 12.2.2.3.7 Effects of Solid Radwaste Handling Areas................................... 12.2-14 12.2.2.3.8 Effects of Liquid Radwaste Handling Areas.................................. 12.2-14 12.

2.3 REFERENCES

......................................................................... 12.2-14 12.3 RADIATION PROTECTION DESIGN FEATURES.............................. 12.3-1 12.3.1 FACILITY DESIGN FEATURES.................................................. 12.3-1 12.3.1.1 Radiation Zone Designations...................................................... 12.3-1 12.3.1.2 Traffic Patterns....................................................................... 12.3-2 12.3.1.3 Radiation Protection Design Features............................................ 12.3-2 12.3.1.3.1 Facility Design Features......................................................... 12.3-2 12.3.1.3.2 Design Features That Reduce Crud Buildup.................................. 12.3-6 12.3.1.3.3 Field Routing of Piping.......................................................... 12.3-7 12.3.1.3.4 Design Features That Reduce Occupational Doses During Decommissioning................................................................. 12.3-7 12.3.1.4 Radioactive Material Safety........................................................ 12.3-8 12.3.1.4.1 Materials Safety Program........................................................ 12.3-8 12.3.1.4.2 Facilities and Equipment......................................................... 12.3-9 12.3.1.4.3 Personnel and Procedures........................................................ 12.3-9 12.3.1.4.4 Required Materials................................................................ 12.3-10 12.3.2 SHIELDING............................................................................ 12.3-10 12.3.2.1 General................................................................................ 12.3-10 12.3.2.2 Methods of Shielding Calculations................................................ 12.3-11 12.3.2.3 Shielding Description............................................................... 12.3-12 12.3.2.3.1 General.............................................................................. 12.3-12 12.3.2.3.2 Reactor Building................................................................... 12.3-12 12.3.2.3.3 Turbine Building.................................................................. 12.3-13 12.3.2.3.4 Radwaste Building................................................................ 12.3-13 12.3.3 VENTILATION........................................................................ 12.3-13 12.3.4 IN-PLANT AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION........................................... 12.3-16 12.3.4.1 Criteria for Necessity and Location.............................................. 12.3-16 12.3.4.2 Description and Location........................................................... 12.3-17

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS (Continued)

Section Page LDCN-05-056 12-iii 12.3.4.3 Specification for Area Radiation Monitors...................................... 12.3-20 12.3.4.4 Specification for Airborne Radiation Monitors................................. 12.3-21 12.3.4.5 Annuciators and Alarms............................................................ 12.3-21 12.3.4.6 Power Sources, Indicating and Recording Devices............................ 12.3-22 12.

3.5 REFERENCES

......................................................................... 12.3-22 12.4 DOSE ASSESSMENT................................................................... 12.4-1 12.4.1 DESIGN CRITERIA.................................................................. 12.4-1 12.4.2 PERSONNEL DOSE ASSESSMENT BASED ON BWR OPERATING DATA.................................................................. 12.4-1 12.4.2.1 General................................................................................ 12.4-1 12.4.2.2 Personnel Dose from Operating BWR Data..................................... 12.4-2 12.4.2.3 Occupancy Factors, Dose Rates, and Estimated Personnel Exposures..... 12.4-2 12.4.3 INHALATION EXPOSURES....................................................... 12.4-4 12.4.4 SITE BOUNDARY DOSE........................................................... 12.4-4 12.

4.5 REFERENCES

......................................................................... 12.4-5 12.5 RADIATION PROTECTION PROGRAM.......................................... 12.5-1 12.5.1 ORGANIZATION..................................................................... 12.5-1 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES.................. 12.5-2 12.5.2.1 Criteria for Selection................................................................ 12.5-4 12.5.2.2 Facilities............................................................................... 12.5-6 12.5.2.3 Equipment............................................................................. 12.5-8 12.5.2.4 Instrumentation....................................................................... 12.5-9 12.5.3 PROCEDURES......................................................................... 12.5-9 12.5.3.1 Personnel Control Procedures..................................................... 12.5-9 12.5.3.2 As Low As Is Reasonably Achievable Procedures............................. 12.5-10 12.5.3.3 Radiological Survey Procedures................................................... 12.5-12 12.5.3.4 Procedures for Radioactive Contamination Control........................... 12.5-13 12.5.3.5 Procedures for Control of Airborne Radioactivity............................. 12.5-14 12.5.3.6 Radioactive Material Control Including Special Nuclear Materials (SNM)..................................................................... 12.5-15 12.5.3.7 Personnel Dosimetry Procedures.................................................. 12.5-16 12.5.3.8 Radiation Protection Surveillance Program..................................... 12.5-18

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 Chapter 12 RADIATION PROTECTION LIST OF TABLES (Continued)

Number Title Page LDCN-14-005 12-iv 12.2-1 Basic Reactor Data for Source Computations.................................. 12.2-17 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary............................ 12.2-18 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation............ 12.2-19 12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown...................................................................... 12.2-20 12.2-5 Fission Product Source in RHR Piping and Heat Exchangers 4 Hours After Shutdown...................................................................... 12.2-21 12.2-6 Gamma Ray Energy Spectrum for Spent Fuel Sources....................... 12.2-22 12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater...... 12.2-23 12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell............................................................. 12.2-24 12.2-9 Nitrogen -16 Source Strength in Feedwater Heater 6......................... 12.2-25 12.2-10 Nitrogen -16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems................................... 12.2-26 12.2-11 Offgas System Sources in the Turbine Generator Building.................. 12.2-27 12.2-12a Special Sources With Strength Greater Than 100 Millicuries............... 12.2-28 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area................................................... 12.2-28a 12.2-13 List of Radioactive Piping and System Designations.......................... 12.2-29 12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building)................................................ 12.2-30

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 Chapter 12 RADIATION PROTECTION LIST OF TABLES (Continued)

Number Title Page LDCN-14-005 12-v 12.2-15 Airborne Radionuclide Concentration in Condensate Pump Area (el. 441 ft. 0 in. turbine generator building).................................... 12.2-31 12.2-16 Airborne Radionuclide Concentration in Secondary Containment from a Main Steam Relief Valve Blowdown................................... 12.2-32 12.2-17 Airborne Radionuclide Concentration in Liquid Radwaste Handling Area........................................................................ 12.2-33 12.3-1 Area Monitors........................................................................ 12.3-25 12.3-2 Maximum Design Basis Background Radiation Level for Area Monitors........................................................................ 12.3-27 12.4-1 Summary of Occupational Dose Estimates...................................... 12.4-7 12.4-2 Occupational Dose Estimates During Routine Operations and Surveillance........................................................................... 12.4-8 12.4-3 Occupational Dose Estimates During Nonroutine Operations and Surveillance........................................................................... 12.4-11 12.4-4 Occupational Dose Estimates During Routine Operations and Surveillance........................................................................... 12.4-12 12.4-5 Occupational Dose Estimates During Waste Processing...................... 12.4-13 12.4-6 Occupational Dose Estimates During Refueling............................... 12.4-14 12.4-7 Occupational Dose Estimates During Inservice Inspection................... 12.4-15 12.4-8 Occupational Dose Estimates During Special Maintenance.................. 12.4-16 12.4-9 Summary of Annual Information Reported by Commercial Boiling Water Reactors....................................................................... 12.4-17 12.5-1 Health Physics Instrumentation................................................... 12.5-21

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Chapter 12 RADIATION PROTECTION LIST OF FIGURES Number Title LDCN-12-037 12-vi 12.3-1 DELETED 12.3-2 DELETED 12.3-3 DELETED 12.3-4 DELETED 12.3-5 Radiation Zones - Turbine Generator Building 12.3-6 Radiation Zones - Ground Floor Plan - Turbine Generator Building 12.3-7 Radiation Zones - Mezzanine Floor Plan - Turbine Generator Building, East Side 12.3-8 Radiation Zones - Mezzanine Floor Plan - Turbine Generator Building, West Side 12.3-9 Radiation Zones - Operating Floor Plan - Turbine Generator Building, East Side 12.3-10 Radiation Zones - Operating Floor Plan - Turbine Generator Building, West Side 12.3-11 Radiation Zones - El. 437 ft 0 in. Radwaste Building 12.3-12 Radiation Zones - El. 467 ft 0 in. and Partial Plans Radwaste Building 12.3-13 Radiation Zones - El. 484 ft 0 in. and 487 ft 0 in., and Partial Plans Radwaste Building 12.3-14 Radiation Zones - El. 501 ft 0 in., 507 ft 0 in., and 525 ft 0 in. Radwaste Building 12.3-15 Radiation Zones - El. 422 ft 3 in., 441 ft 0 in., and 440 ft 0 in. Reactor Building 12.3-16 Radiation Zones - El. 471 ft 0 in. and 501 ft 0 in. Reactor Building 12.3-17 Radiation Zones - El. 522 ft 0 in. and 548 ft 0 in. Reactor Building 12.3-18 Radiation Zones - El. 572 ft 0 in. and 606 ft 10-1/2 in. Reactor Building

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 12 RADIATION PROTECTION LIST OF FIGURES (Continued)

Number Title LDCN-05-056 12-vii 12.3-19 Arrangement of Filtration and Demineralization Equipment (Typical) 12.3-20 Schematic Arrangement of the Cooler Condenser Loop Seal 12.3-21 Decontamination Concentrator Steam Supply Arrangement 12.3-22 Entombment Structure 12.3-23 Layout of the Standby Gas Treatment System Filter Units 12.3-24 Block Diagram - Area Radiation Monitoring System

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-042 12.1-1 Chapter 12 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS Energy Northwest is committed to maintaining occupational and public radiation exposures as far below regulatory dose limits as is practical while performing all activities related to the operation of Columbia Generating Station (CGS) and the Independent Spent Fuel Storage Installation (ISFSI). This commitment is reflected in the Radiation Protection Program (RPP) that meets the requirements of 10 CFR 20 and provides for effective control of radiation exposure through

a.

Management direction and support,

b.

Establishment of radiation control procedures,

c.

Consideration during design and modification of facilities and equipment, and

d.

Development of good radiation control practices, including preplanning and the proper use of appropriate equipment by qualified, well trained personnel.

The radiation protection practices are based, when practicable and feasible, on Regulatory Guides 8.8, Revision 3, and 8.10, Revision 1. The Radiation Protection Program provides for the majority of the recommended actions in both regulatory guides, including the following:

a.

Organization and position descriptions to ensure an adequate as low as is reasonably achievable (ALARA) program,

b.

Exposure reduction program,

c.

Cost-benefit analysis program, and

d.

Exposure tracking program employing the Radiation Work Permit.

Procedures for personnel radiation protection are prepared consistent with the requirements of 10 CFR Part 20 and are approved, maintained, and adhered to for all operations involving personnel radiation exposure.

Energy Northwest organization is structured to provide assurance that the ALARA policy is effective in the areas described above. The following is a description of the applicable activities conducted by individuals or groups having responsibility for radiation protection.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-13-061 12.1-2

a.

The Plant General Manager, has the overall responsibility for approving the RPP and ALARA policy consistent with Energy Northwest and regulatory requirements, and for the radiological safety of all on-site personnel. This includes the responsibility for implementation of the ALARA program by plant staff. The Plant General Manager has ensured consideration of exposure reduction methods by adoption of the ALARA review practices described in Section 12.1.3. The Plant General Manager is responsible for the management of plant operational activities and for providing the Radiological Services Manager the resources and support necessary to implement the RPP. This includes the responsibility for ensuring that the ALARA program is not adversely affected by production oriented goals;

b.

The Radiological Services Manager reports to the Chemistry and Radiological Safety Manager and is responsible for implementing the RPP. This position meets the requirements of the Radiation Protection Manager (RPM) as described in Reg Guide 1.8. This individual provides organizational leadership and direction to the Radiation Protection department;

c.

The Radiological Services Manager has direct access to the Plant General Manager in all matters relating to radiation safety, and has the responsibility and authority for ensuring that plant activities meet applicable radiation safety regulations and RPP requirements. Specific responsibilities are provided in Section 12.5.1;

d.

The Radiological Operations Supervisor reports to the Radiological Services Manager. This individual provides supervision, leadership, and technical direction for implementation of the RPP;

e.

The Health Physics (HP) Craft Supervisors report to the Radiological Operations Supervisor and implement the RPP through direct supervision of the plant HP Technicians. Areas of responsibility include characterization of plant radiological conditions, maintenance of radiological postings, detection and evaluation of radiological problems, and performance of facility and equipment decontamination, system flushes, and temporary shielding installation;

f.

The Radiological Support Supervisor reports to the Radiological Services Manager and is responsible for radiation exposure reduction and for providing technical support to the Radiation Protection organization. This is accomplished, in part, by providing input to the planning and coordination of plant activities. The Radiological Support Supervisor makes recommendations

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-15-028 12.1-3 for the control/elimination of radiological conditions that increase personnel exposure or the release of radioactivity; and provides additional assurance that regulatory requirements and industry guidance are incorporated into the program to ensure exposures are maintained ALARA.

In addition to the organization structure that has been provided for implementation of the ALARA policy, the following groups perform reviews to further ensure exposures are ALARA.

a.

The Plant Operations Committee (POC) has been established and is functional.

Its purpose is to serve as a review and advisory organization to the Plant General Manager in several areas including nuclear safety. A written administrative procedure describes the responsibilities of the POC. The RPM is a member of the POC and has direct input to this group on all radiological matters;

b.

The Quality organization performs audits, surveillances, and assessments of plant operations. Audits, surveillances, and assessments will include verification of compliance with plant procedures, with regulations involving nuclear safety, and with operating license provisions. Results of these audits and surveillances will be submitted to both plant and corporate management.

Since the system for ALARA review described in Section 12.1.3 provides for this consideration in all plant procedures, quality audits and surveillances will verify implementation of this principle;

c.

The Corporate Nuclear Safety Review Board (CNSRB) provides a continuing appraisal of the ALARA policy. The Operational Quality Assurance Program Description (OQAPD) provides a description of this groups responsibilities and authority. By its charter, the CNSRB reviews radiological safety policies and programs and determines that these policies and programs are in compliance with NRC requirements. The CNSRB has the capacity for review in the area of radiological safety and has established a direct line of communication with plant management; and

d.

The Senior Site ALARA Committee serves as a review and advisory organization to the Plant General Manager on radiological safety, including occupational exposure to personnel. Committee membership, responsibilities, authorities, and records are prescribed in plant procedures.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-4 The commitment to ALARA is implemented by employee training; audits, assessments, and reviews of the program; procedure development and reviews; enforcement of rules; and modifications to plant equipment or facilities where they will substantially reduce exposures at a reasonable cost. Managements commitment to the ALARA policy is further discussed in Section 12.5. This program meets all 10 CFR Part 20 regulations and considers the guidance of Regulatory Guides 8.8 and 8.10 in regard to policy considerations.

12.1.2 DESIGN CONSIDERATIONS To ensure that personnel occupational radiation exposures are ALARA, extensive consideration is given to equipment design and locations, accessibility requirements, and shielding requirements. Many of these design objectives and considerations were established prior to the issuance of Regulatory Guide 8.8. However, the design of the plant substantially incorporates the recommendations provided in the regulatory guide. Design considerations that ensure occupational radiation exposures to personnel during normal operation and anticipated operational occurrences are ALARA are the following:

a.

The facility is separated into controlled and uncontrolled areas based on anticipated radiation levels. The controlled areas of the facility are further defined by radiation zones established by personnel access requirements and are intended to limit radiation exposure to ALARA. The radiation zones provide guidance for the shielding design, contamination control, and ventilation flow pattern for the areas of anticipated personnel radiation exposure. See Section 12.3.1 for a description of the facility radiation zones.

b.

Equipment location

1.

Several radiation sources on the 437 ft 0 in. level of the radwaste building are located with two sources in each cubicle. This arrangement maintains occupational exposures ALARA by use of the following alternate ALARA methods.

The waste collector tank and floor drain collector tank are in the same cubicle. These tanks share redundant pumps and cross tie piping. If abnormal conditions occur and one or both of these tanks becomes a major source of radiation, either one of the pumps can be used to empty the tanks prior to any maintenance.

The chemical waste tank and distillate tank share the same cubicle.

These tanks are not expected to be major sources of radiation. Based on the source terms described in Table 11.2-1, the dose rate at 3 ft from the surface of these tanks normally does not exceed 0.1 mrem/hr. In

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-5 addition, redundant pumps and cross tie piping permit the transfer of tank contents should abnormally high radioactivity levels occur.

Gas coolers and charcoal adsorbers share the same cubicle. These items have no moving parts and are highly reliable with no routine maintenance requirements. In addition, system redundancy and remote isolation capabilities eliminate the need for prompt entry into the cubicle.

This permits the noble gases and radioiodines to significantly decay prior to entry.

Placing the preceding sources in shared cubicles does not result in increased occupational exposures.

2.

Radioactive pipes are routed so that radiation exposure to plant personnel is minimized. The extent to which radioactive pipes are routed through normally accessible areas is minimized. Shielded pipe chases are utilized in normally accessible areas. Whenever possible, radioactive and nonradioactive pipes are kept separate for maintenance purposes.

3.

Shielded valve stations are used where practical. To further minimize personnel exposure, remotely operated valves are used where practical.

Normally operated manual valves in high radiation areas are provided with extension stems through a shield wall to a low radiation area.

4.

Where practical, pumps are located in shielded areas outside cubicles containing radioactive components.

5.

Where practical, local instrumentation readouts are routed to points outside shielding walls.

6.

To minimize maintenance time and hence exposure, sufficient space is available in shield cubicles housing radioactive equipment (e.g., heat exchangers, demineralizers, etc.) to perform required tasks. Access platforms at intermediate levels are provided to enhance access to portions of equipment inaccessible from the floor.

7.

Where possible, equipment and components which require frequent servicing are designed so that they can be removed, with minimum exposure, to appropriate low radiation areas.

8.

Access to corridor C-125 on the 437 ft 0 in. level of the radwaste building will only be required for routine surveillance of equipment and nonroutine maintenance. Area radiation monitors 28 and 29 are located

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-6 in the corridor to detect abnormal radiological conditions and warn personnel if radiation levels are excessive. During reactor and radwaste operations, entry to this area will be under the direction of a Radiation Work Permit (RWP).

c.

Shielding design is based on satisfying the radiation zone requirements utilizing design basis radiation sources for the shielding calculations. Shielding design is conservative since the design basis radiation sources are not expected to occur frequently.

Penetrations through radiation shields are generally designed to prevent direct radiation streaming from a high radiation area to a low radiation area. Shielding discontinuities, such as shield plugs, are provided with offsets to reduce radiation streaming. Shield doors and labyrinths are used to eliminate radiation streaming through access openings in the cubicles.

d.

Auxiliary systems that may become contaminated are designed with provisions for flushing or remote chemical cleaning prior to maintenance. This is accomplished by the following:

1.

Providing connections for the purpose of backflushing,

2.

Providing water connections to tanks containing spargers to allow for water injection to uncake contaminants, and

3.

Providing cross ties between redundant equipment and/or related equipment capable of redundant operation to allow removal of contaminated equipment from service.

e.

The ventilation systems are designed to ensure control of airborne contaminants by providing air flow from areas of low radioactivity potential to areas of high radioactivity potential. Access to the systems is facilitated by the following:

1.

Filter access doors, which are sized to enhance the ease of performing maintenance, and

2.

Providing for periodic inservice testing of the equipment and filters.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-7

f.

Spread of contamination is minimized in the event spillage occurs by the following:

1.

Drains are provided in areas where equipment with large volumes of radioactive fluid is located. Drains are sized to conduct spillage to the appropriate liquid waste processing system;

2.

Floors and walls are protected with the appropriate coating to facilitate decontamination; and

3.

An equipment decontamination facility is provided to decontaminate tools and radioactive components.

g.

While pipe runs are not sloped, those that carry radioactive fluids can be chemically decontaminated. Tank bottoms in the radwaste system are either sloped or dished. The exception is the waste surge tank located on el. 437 ft 0 in. level in the radwaste building.

h.

Drain tap-offs are provided at low points in the piping systems.

i.

Connections are placed above the centerline (top) of pipes when consistent with overall design requirements. Such connections are not always practical or desirable. For example, air pockets may form in an infrequently used line which is connected above the centerline (top) of another pipe.

j.

Where practicable, the use of dead legs and low points in pipe systems are avoided, especially in pipes carrying demineralizer resins or concentrated radwaste.

k.

T-connections in piping are minimized with the exception of

1.

Multiple flow paths, such as in the condensate filter demineralizer system, and

2.

Interconnected systems acting as mutual back ups, such as the floor drain and waste filter demineralizer systems.

l.

Large pipe bend radii and piping elbows are used.

m.

Butt welding by the open root method is used as described in Section 12.3.1.3.2.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-022 12.1-8

n.

Seal glands of pumps carrying concentrated radwaste or spent filter and demineralizer resins are flushed with condensate water during operation. After operation, the pumps are also flushed with condensate. Canned pumps are not used.

o.

Where practicable, equipment used in the same process are located close together, resulting in short runs of interconnecting piping.

p.

Pipes carrying slurries and resins are sized for turbulent flow to prevent any settling out of solids.

q.

All equipment cubicles housing filters are equipped with removable ceiling plugs through which filter elements may be serviced or changed with the aid of tools to allow remote handling.

r.

Operating experience from other BWR plants is periodically reviewed.

Problems are reviewed and the plant design is checked to ensure that similar problems will not occur.

s.

Design changes are reviewed by Radiation Protection.

12.1.3 OPERATIONAL CONSIDERATIONS 12.1.3.1 Procedures and Methods of Operation A positive means of ensuring that occupational and public radiation exposures are ALARA has been incorporated into the Plant Procedures Manual (PPM) and Procedure Program.

Procedures are formally reviewed for ALARA considerations as part of the approval process.

The guidance provided by Regulatory Guide 8.8 is considered during this review. Additional information regarding ALARA considerations in procedures is provided in Section 12.5.3.2.

In addition to the above process, the Radiation Work Permit (RWP) process is used for individual tasks to identify radiological conditions and controls. This provides a means of prescribing precautionary measures, protective equipment, and other exposure reduction methods in each situation. Individual exposures, as determined by dosimeters, are recorded and provide a means of determining exposures for various tasks and work groups. This information is used for preplanning work, identifying sources, determining radiation levels and otherwise evaluating exposure problems.

Administrative controls ensure that occupational and public radiation exposures are as low as reasonably achievable. These are supplemented by the Radiation Protection staff through a

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-042 12.1-9 program of surveillance, guidance, and investigation. A description of the program is outlined in Section 12.5 and includes the following aspects:

a.

The Energy Northwest RPP includes procedures that provide for routine and special survey to determine sources and trends of exposure and for investigation to determine causes of normal and unusual exposure;

b.

Plant procedures are formally reviewed by Radiation Protection for ALARA considerations when required;

c.

Plant modifications that have radiological implications are reviewed by Radiation Protection to ensure that the changes incorporate ALARA considerations;

d.

All maintenance tasks that involve radiological systems or controls are initially reviewed for ALARA and radiological concerns. For these tasks, a RWP is completed which describes the proposed actions. The RWP describes the conditions necessary such as survey requirements, surveillance, and protective apparel;

e.

Prior to each scheduled maintenance and refueling outage, HP reviews the outage tasks and schedules and participates in outage planning for ALARA considerations; and

f.

Adequate surveillance and supervision is provided to ensure that procedures are followed, prescribed precautions are taken, and radiation sources are identified.

12.1.3.2 Design Changes for ALARA Exposures Operational requirements were considered in the original design of CGS for maintaining occupational exposures ALARA, and several design changes have been made since issuance of the PSAR and Regulatory Guide 8.8. These changes or additions were implemented as a result of review by both the architect-engineer and Energy Northwest personnel and include the following:

a.

Revised offgas system valve design to prevent release of radioactive gases to building atmosphere,

b.

Relocation of the counting room for lower background levels and adequate shielding,

c.

Revised effluent monitoring capabilities to provide for more efficient monitoring,

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-10

d.

Increased capability for in-plant continuous airborne radioactivity monitoring with remote readout and recording features,

e.

Increased capability for the area radiation monitoring system to include remote readout/alarm and local alarm functions for the individual monitors,

f.

Inclusion of supplied air stations throughout the plant for efficient respiratory protection,

g.

Space and services provisions made for a decontamination facility and hot shop to reduce contact maintenance exposures and airborne radioactivity,

h.

Revised penetration access design at sacrificial shield wall to reduce time required in this area,

i.

Installation of overload protection on valve motor operators to minimize motor replacement in high dose rate areas,

j.

Generated additional specification for replacement valve packing for selected valves to reduce time consumed in repacking,

k.

Replaced hydraulic snubbers with mechanical snubbers to reduce maintenance requirements,

l.

Provided method of venting the reactor vessel head through main steam line A to be ultimately filtered through the offgas system via the condenser, and

m.

Made provisions for future connections to increase reactor water cleanup capacity during shutdown conditions to reduce radioiodine and particulate activity prior to vessel head removal and during outage periods.

New designs or design revisions are considered for exposure reduction as plant operation identifies problem areas.

12.1.3.3 Operational Information Information from operating BWRs has been incorporated into the Radiation Protection procedures as discussed below:

a.

Procedures have been written using guidance obtained from BWR operational experience reviews, from technical information obtained at power reactor conferences, and from participational experience in outages at operating BWRs;

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-11

b.

Respiratory protection procedures incorporate proven practices from other nuclear facilities;

c.

Typical procedures on survey methods, personnel monitoring, personnel dosimetry, and process/effluent radiological monitoring have been observed in the implementation stage at several operating reactor facilities. In addition, the procedures used at operating BWR plants have been reviewed. All of these were evaluated and applied in the procedure generating process;

d.

Specific HP procedures or instructions have been written to furnish guidance on the following:

1.

The issuance, requirements, conditions, and controls of RWPs,

2.

The review process of plant procedures for ALARA considerations, and

3.

Methods for minimizing personnel exposures during RPV head removal, drywell entry, and conduct during emergencies.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-1 12.2 RADIATION SOURCES 12.2.1 CONTAINED SOURCES 12.2.1.1 General The design basis radiation sources considered are the following:

a.

The reactor core,

b.

Activation of structures and components in the vicinity of the reactor core,

c.

Radioactive materials (fission and corrosion products) contained in system components,

d.

Spent fuel, and

e.

Radioactive wastes for offsite shipment.

The design basis process radiation sources used for shielding are based on the source terms given in Tables 11.1-1 through 11.1-5.

12.2.1.2 Reactor and Turbine Building The reactor building sources include the following:

a.

The reactor core,

b.

Activated structures and components,

c.

Components and equipment containing activation, fission, and corrosion products, and

d.

Spent fuel.

12.2.1.2.1 Reactor Core Radiation Sources During normal power operation, the reactor core radiation sources considered are the prompt fission neutrons and gamma rays, capture gamma rays, and fission product gamma rays.

During shutdown, the reactor core radiation sources are the fission product gamma rays. The core is treated as a cylindrical volume source for use in the NRN+(1) and QAD+(2) codes.

See Section 12.3.2 for details.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-98-117 12.2-2 Table 12.2-1 lists the basic data required for reactor core source computations including volume fractions, power peaking factors, and core power density.

Table 12.2-2 presents the neutron multigroup flux at the reactor core-reflector boundary, 238 cm from the core centerline. The NRN computer code (Reference 12.2-1) is used in computing the flux. This code is based on a multigroup slowing down and diffusion system with the deep penetration source corrected by a multigroup removal source.

Table 12.2-3 lists the gamma ray energy spectrum for the reactor core during normal operation. The operating core energy spectrum represents the sum of the spectra for prompt fission, fission product, and to capture gamma rays. The postoperation fission product gamma ray energy spectrum in the core immediately after shutdown is listed in Table 12.2-4.

12.2.1.2.2 Process System Radiation Sources 12.2.1.2.2.1 Introduction. The following process systems govern the shielding requirements within the reactor and turbine buildings:

a.

Recirculation (RRC),

b.

Reactor water cleanup (RWCU),

c.

Reactor core isolation cooling (RCIC),

d.

Residual heat removal (RHR),

e.

Fuel pool cooling and cleanup (FPC),

f.

Main steam (MS) and the reactor feedwater system (RFW),

g.

Traveling in-core probe (TIP), and

h.

Offgas system (OG).

The locations of the equipment, which are part of these systems, are shown on the radiation zone drawings, Figures 12.3-5 through 12.3-18.

12.2.1.2.2.2 Recirculation System Sources. The coolant activation products, principally 16N, are the dominant sources of radiation in the RRC system during normal operation. The 16N introduced into this system decays significantly by the time it returns to the core. Therefore, decay is considered in the portions of this system where elapsed transit time is significant.

For shielding purposes, the pipes in the RRC system are treated as equivalent line sources.

The 16N source strength in these lines varies from 0.10 Ci/cm at the line from the reactor vessel to 0.02 Ci/cm at the header which distributes coolant flow back to the reactor. These sources are located within the drywell of the primary containment of the reactor building, from approximately el. 501 ft to el. 540 ft.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-3 During shutdown, fission products become the dominant source. This shutdown source is considerably less than that of 16N during normal operation. Shielding design is based on the 16N source, which is more than adequate to shield against the fission product shutdown source.

12.2.1.2.2.3 Reactor Water Cleanup System Sources. The major sources of radiation in the portions of the RWCU system in the reactor building during normal operation are the coolant activation products, principally 16N. The 16N source strength (given in activity per unit length of line) in the RWCU system ranges from 1.00 x 10-3 Ci/cm in the lines flowing into the cleanup recirculation pumps to 4.68 x 10-8 Ci/cm at the exhaust of the tube side of the nonregenerative heat exchanger. Returning from the radwaste building, the 16N source strength ranges from 3.08 x 10-10 Ci/cm to negligible (less than 10-14 Ci/cm).

The 16N source strengths in the regenerative and nonregenerative heat exchangers are

a.

Tube side of the regenerative heat exchanger: 2.69 x 10-6 Ci/cm3,

b.

Tube side of nonregenerative heat exchanger: 6.24 x 10-8 Ci/cm3, and

c.

Shell side of the regenerative heat exchanger: 1.70 x 10-14 Ci/cm3.

These heat exchangers are treated as cylindrical sources for input into the QAD computer program (see Section 12.3.2). The portions of this system located in the radwaste building are discussed in Chapter 11. The RWCU pumps are located at el. 522 ft 0 in. and the RWCU regenerative and nonregenerative heat exchangers are located at el. 548 ft 0 in.

During shutdown, the fission products are the dominant radiation source. Since the shielding is designed to shield against the 16N source during normal operation, it is more than adequate to shield against the shutdown fission product source.

12.2.1.2.2.4 Reactor Core Isolation Cooling System Source. The dominant source of radioactivity in the RCIC system is the 16N present in the main steam used to drive the RCIC turbine. This 16N source occurs when the RCIC turbine is tested on a monthly basis. The shielding design is based on this source. The transit time is assumed to be negligible in this system.

The resulting 16N activity (given in activity/unit length of line) in the inlet line is 2.96 x 10-4 Ci/cm and in the outlet line, it is 6.57 x 10-5 Ci/cm. These sources are treated as equivalent line sources for shielding purposes.

The RCIC turbine source strength is 8.44 x 10-2 Ci of 16N. It is treated as a point source since the distance to the dose points is much greater than the turbine dimensions. The RCIC pump and turbine is located at el. 422 ft 3 in. of the reactor building.

12.2.1.2.2.5 Residual Heat Removal System Sources. The RHR system radiation sources consist of the fission and corrosion products. Table 12.2-5 lists the gamma ray energy

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-4 spectrum of the radionuclides in the RHR pumps, pipes, and heat exchangers 4 hr after shutdown. These sources are based on the RHR system operating in the normal reactor shutdown cooling mode. The fission and corrosion product isotope concentrations used are listed in Tables 11.1-2 through 11.1-4.

The RHR heat exchangers are located approximately from el. 559 ft 0 in. to el. 589 ft 0 in. on the west side of the reactor building. The RHR pumps are located at el. 422 ft 3 in. on the west side of the reactor building.

The pipes in this system are treated as equivalent line sources. The heat exchangers are treated as cylindrical sources.

12.2.1.2.2.6 Fuel Pool Cooling and Cleanup and System Sources. The primary sources of radioactivity in the spent fuel assemblies, which are stored in the fuel pool, are the fission products. Table 12.2-6 lists the gamma ray energy spectrum for the spent fuel sources for shutdown time of 2 days. The refueling area is located at el. 606 ft 10.5 in. in the reactor building.

These source terms are calculated using the Perkins and King data (Reference 12.2-2). The shielding calculations are done using the QAD point kernel code (Reference 12.2-3). The following assumptions are used in determining the shielding requirements:

a.

After radioactivity has reached equilibrium in the fuel assemblies, it is assumed that the reactor is shut down and the whole core is moved, within 2 days, into the spent fuel pool;

b.

The whole core and another one-fourth of a core from the last refueling are located by the north wall of the spent fuel pool to give the most conservative dose rate on the outside of the wall. Less water exists between the assembly racks and the north wall than between the assembly racks and any other side of the pool. The assemblies from past refuelings do not add to the shielding requirements because they have decayed for more than 1 year, they are shielded by pool water, and they provide self shielding; and

c.

The water, racks, spent fuel, and other constituents that are located within the array of spent fuel assemblies are homogenized for the purpose of determining the required values of the linear attenuation coefficients.

The minimum depth of water needed to adequately shield the refueling area from the spent fuel assemblies is calculated. It is found that the elevated fuel assembly during the fuel transfer operation controls the dose rate to the refueling area and, hence, the water depth in the spent fuel pool. The fuel assembly is treated as a cylindrical source geometry for the purpose of computing the water depth.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-5 The source strength used to determine the shielding requirements for the dryer-separator pool is based on a contact dose rate for the separator of 10 R/hr. The average gamma ray energy is approximately equal to 1 MeV.

12.2.1.2.2.7 Main Steam and Reactor Feedwater Systems Sources. In these systems, the dominant sources of radioactivity are the coolant activation products, principally 16N. The following equipment is considered:

a.

Moisture separators and reheaters (MSR),

b.

Main condenser and hotwell,

c.

Feedwater heaters, and

d.

The piping associated with these systems.

The moisture separator and reheaters govern the shielding requirements on the operating floor (el. 501 ft 0 in.) of the turbine building. The 16N source strengths for the first stage reheater tubes, the second stage reheater tubes, and the MSR plena are listed on Table 12.2-7. The first and second stage reheater tubes are approximated by rectangular parallelepipeds. The plena are divided into an array of rectangular parallelepipeds and cylinders, depending on their physical arrangement.

The 16N source strength in the main condenser is 6.0 x 10-8 Ci/cm3. This is based on the incoming flows from the exhaust of the reactor feed pump turbine, and the exhaust steam from the low pressure turbine. The main condenser is treated as either a truncated cone or infinite slab depending on the view angle and distance from the condenser to the dose point.

Since most of the 16N exists as a noncondensable gas, it comes out of solution in the main condenser. Thus, the radiation sources in the hotwell are the fission or corrosion product radionuclides. Table 12.2-8 lists the gamma ray volumetric source strengths calculated from the hotwell. The hotwell is treated as an infinite slab.

The 16N source strength of feedwater heater 6 listed in Table 12.2-9, governs the shielding requirements on the mezzanine floor of the turbine building (el. 471 ft 0 in.). The feedwater heater is treated as an array of rectangular parallel pipes and cylinders for input into QAD.

Table 12.2-10 lists the 16N source strengths in selected steam piping in the MS and RFW systems.

12.2.1.2.2.8 Offgas Sources in the Turbine Generator Building. Table 12.2-11 lists the source strength from the offgas system equipment in the turbine building. Nitrogen-16 is the dominant radionuclide present in this system. The offgas equipment is located at el. 441 ft 0 in. of the turbine building.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-6 12.2.1.2.2.9 Traveling In-Core Probe System Sources. The primary source of radiation in the TIP system is the 56Mn in the traversing in-core probe (TIP) cable. The average source strength per unit length of cable is 3.27 x 104 Ci/cm. This is calculated using an exposure time of 864 sec. The average radioactivity emitted per unit length is calculated using a cosine distribution for the neutron flux in the axial direction of the core. The cable is treated as a line source for shielding purposes. The TIP components are located at el. 501 ft 0 in. of the reactor building.

12.2.1.2.2.10 Sources Resulting From Crud Buildup. Fission and corrosion product deposition is complex to describe analytically. In estimating the effects of this source, operating experience from other BWR plants is used (Reference 12.2-4); Section 12.4 discusses this in greater detail.

12.2.1.3 Radwaste Building The radiation sources present in the radwaste building are discussed in Chapter 11.

12.2.1.4 Byproduct, Source and Special Nuclear Materials A list of all byproduct, source and special nuclear material in sealed sources with an activity greater than 100 mCi is given in Table 12.2-12a. Several areas outside the Reactor, Turbine, and Radwaste Buildings and within the site area have been established for use and storage of radioactive material in the form of activated components, sealed sources, and contaminated tools, equipment, and protective clothing. Those areas where the total quantity of activity could at times be greater than 100 mCi under normal conditions are listed in Table 12.2-12b.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 12.2.2.1 General Design features that limit the airborne radioactivity in normally occupied areas are incorporated into the plant. Areas that are designated as radiation Zones I and II are considered to be normally occupied areas.

The plant was designed so that the airborne radionuclide concentrations in normally occupied areas are well below the limits specified in 10 CFR Part 20. The criteria for Zones I and II are specified in 10 CFR 20, Appendix B to 20.1001-20.2401, Table 1, Column 3.

No radiation Zone I areas exist in the reactor or turbine generator building. The only Zone I areas found in the radwaste and control building are the main control room and the counting room shown on Figures 12.3-14 and 12.3-13, respectively. The main control room is located at el. 501 ft 0 in. It has a separate heating, ventilating, and air conditioning (HVAC) system which brings in fresh air from the outside. See Section 9.4.1 for further discussion. The

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 12.2-7 counting room is located at el. 487 ft 0 in. As seen in Figure 9.4-3, the incoming ventilation flow into this room consists of fresh air. The outgoing air flow is to the area surrounding the counting room. It is concluded that the airborne concentration in the counting room is small.

See Section 12.2.2.3.5 for discussion on the contribution of sampling and radiochemical analysis on airborne radioactivity levels within this area.

12.2.2.2 Model for Computing the Airborne Radionuclide Concentration in a Plant Area The model used for computing the airborne radionuclide concentration is based on the continuous leakage of a radioactive fluid into a plant area. The removal of radionuclides from this area is through ventilation exhaust and decay. The equation which yields the airborne radionuclide concentration in a plant area is:

C A q PF i

q q

V i

i s

i v

a i

a

(

)

exp (

/

)

1 t

(12.2-1) where:

Ci

= concentration of radionuclide i in a given plant area (ci/cm3)

Ai

= concentration of radionuclide i in the fluid (mCi/g) qs

= rate of radionuclide leakage into an area (g/minute)

(PF)i = partition factor for radionuclide i (dimensionless) i

= decay constant for isotope i (1/minute)

V

= volume of area (cm3) qa

= HVAC air flow rate out of area (cm3/minute) t

= time interval between start of leak and calculation of concentration (minute)

The equilibrium value of Ci is given by C

A q PF V

q i

i s i

i a

(

)

(12.2-2)

Note that in all analyses where these two equations are used, the partition factor is set equal to 1.0. This yields conservative results.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 LDCN-01-069 12.2-8 12.2.2.3 Sources of Airborne Radioactivity The potential sources of airborne radioactivity found in the plant are as follows:

a.

Leakage from process equipment in radioactive systems, such as valves, flanges, and pumps,

b.

Sumps, drains, tanks, and filter/demineralizer vessels which contain radioactive

fluid,

c.

Exhaust from relief valves,

d.

Removal of reactor pressure vessel (RPV) head and associated internals,

e.

Radioactivity released from sampling, and

f.

Airborne radioactivity released from the spent fuel pool water and spent fuel movement.

Sections 12.2.2.3.1 through 12.2.2.3.6 discuss each of these sources and their effect on the airborne radionuclide concentration in normally occupied areas of the plant. All design features that serve to reduce the airborne radionuclide concentration are also discussed.

12.2.2.3.1 Effect of Leakage from Process Equipment in Radioactive Systems Leakage into normally occupied plant areas from radioactive process systems is described by three parameters.

The first parameter is the location of the source of radioactive leakage. If this source is located in a Zone III or Zone IV radiation area, it does not contribute to the airborne radionuclide concentration in a normally occupied area because the air flow is from areas of potentially low radioactivity contamination to areas of potentially high radioactivity contamination. Any leakage of radioactive airborne contamination into a Zone III or Zone IV radiation area in the reactor, radwaste and control or turbine generator building is exhausted from that area through the HVAC system and is not transported into a normally occupied area. See the HVAC flow diagrams for the reactor, radwaste and control or turbine generator buildings, Figures 9.4-2, 9.4-3, and 9.4-6, and the radiation zone drawings, Figures 12.3-5 through 12.3-18.

Areas with multiple zone designation are regarded as having a high radioactivity contamination potential. Air from these areas is either exhausted to another Zone IV area or to the HVAC exhaust equipment. Thus, any radioactive leakage that occurs within an area with a multiple zone designation should not affect the airborne radionuclide concentration of a normally occupied area.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 LDCN-01-069 12.2-9 The second parameter is the rate at which the radioactive leakage is introduced into an area.

Any system that operates continuously is potentially a greater source of airborne radioactivity than a system that operates periodically when radionuclide concentration and leakage rate are the same for each system. The operating pressure of the system is another consideration which affects the leakage rate. A system, such as the main steam system, is expected to have a higher leakage rate than a system which operates at atmospheric pressure. However, radioactive systems which operate at high pressure are generally located in Zone IV areas.

Thus, these systems do not significantly contribute to the airborne radioactivity level in normally occupied areas. This is due to the HVAC air path which was discussed earlier.

The third parameter is the radionuclide concentration in the leaking fluid. A system that is leaking highly radioactive fluid such as reactor coolant or main steam is of greater concern, initially, than a system which is leaking condensate storage tank water. Those systems which can leak highly radioactive fluid are located in either a Zone III or Zone IV radiation area and the HVAC air flow paths do not transport this radioactivity to a low radiation area, as discussed earlier. The effect of systems which leak fluid with a low radionuclide concentration is discussed below.

A list of all radioactive systems found in the plant is provided in Table 12.2-13. Each radioactive system listed has been checked to determine its contribution to the airborne radioactivity levels in normally occupied areas. It is found that most of these systems are located in Zone III or Zone IV areas and do not contribute to the airborne radioactivity levels in normally occupied areas, due to HVAC design features as explained earlier. Some of the radioactive systems are located within areas which have a multiple radiation zone designation; e.g., Zone IV during system operation and Zone III during system shutdown. The systems in these areas do not contribute to the airborne radioactivity levels due to HVAC system design features, as explained earlier. Any system which is not entirely located in a Zone IV, Zone III, or multiple zone radiation area and which may contribute to airborne radionuclide levels in normally occupied areas is discussed in the following paragraphs. Those systems which are used only during loss-of-coolant accident (LOCA) conditions are not discussed.

These include the high-pressure core spray (HPCS), low-pressure core spray (LPCS), and main steam leakage control (MSLC) systems. These systems are not regarded as being significant sources of airborne radioactivity since they are used only in a test mode or in an accident situation.

The major source of control rod drive (CRD) leakage which can contribute to the airborne radionuclide levels in a normally occupied area are the CRD pumps located in the CRD pump room. The CRD pump room is located between column lines 3.4/6 and N/N.8 at el. 422 ft 3 in. of the reactor building. Figure 4.6-5 shows suction flow for the CRD pumps is from the outlet of the condensate filter demineralizers or the condensate storage tank. Only one pump is in operation at any time and the leakage rate from this pump is assumed to be 50 gal/day. The incoming HVAC air flow to this area is approximately 3500 ft3/minute. The

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-10 mathematical model described in Section 12.2.2.2 is used to evaluate the airborne activity.

The results are listed in Table 12.2-14. It is seen that the expected airborne radionuclide level is well below the derived air concentrations (DAC) values.

The major source of main condensate leakage which can contribute to the airborne radionuclide levels in a normally occupied area are the condensate pumps. These pumps are located between column lines D/F and 13/15 at el. 441 ft 0 in. of the turbine generator building. The suction flow from these pumps is taken from the main condenser hotwell. Since the suction side of the pumps is at a negative pressure to the surrounding area, no leakage from the pump seals should occur. However, leakage may occur from the valves and other equipment which are located on the discharge side of the pump. It is conservatively assumed that the leakage is 1 gpm. The airborne radionuclide concentration in the area where the condensate booster pumps and condensate pumps are located is listed in Table 12.2-15.

The offgas equipment which is radioactive is located in Zone IV areas. The exception is the afterfilter, which is located between column lines K.1/L.9 and 14.7/15.4 at el. 452 ft 0 in. of the radwaste building. This filter is not a source of airborne radioactivity. The reasons are that the filter operates at approximately atmospheric pressure and the radioactivity contained within the filter is small. See Section 11.3 for details on offgas system operation.

12.2.2.3.2 Effect of Sumps, Drains, Tank, and Filter Demineralizer Vents The equipment drain (EDR), floor drain (FDR), and miscellaneous radwaste (MWR) systems are designed to collect and process various types of liquid radioactive waste generated in the reactor, radwaste, or turbine generator buildings, as explained in Section 11.2. In each of the aforementioned buildings, there is a network of EDR, FDR, and MWR drains which conduct radioactive liquid waste to an EDR, FDR, and MWR sump, respectively. These drains and sumps are not significant sources of airborne radionuclides for the following reasons:

a.

Each of the EDR, FDR, and MWR sumps present in the reactor, radwaste, and turbine generator buildings is covered with a steel plate. The free volume in the sump is maintained at a negative pressure with respect to the surrounding area by use of a riser vent, which is connected to the HVAC system in the building of concern. The steel plate covering the sump does not provide an airtight seal. However, air is drawn into the sump, then through the riser vent and is exhausted to the HVAC system. Any radionuclides that escape into the free volume of the sump are discharged to the HVAC system and do not escape into the area surrounding the sump; and

b.

The EDR, FDR, and MWR drains are connected to their respective sumps through the same riser vents discussed in the preceding paragraph. Some of the drains employ loop seals, which prevent radioactive gases from escaping into the areas around the location of the drains. Other drains do not employ loop

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-11 seals, but since the riser vent is connected to the HVAC system, air will be drawn into the drain through the riser vent and out to the HVAC system.

The tanks and filter demineralizer vessels that contain significant inventories of radionuclides are vented to the HVAC system. These tanks and filter demineralizer vessels are located in Zone III or Zone IV radiation areas. Even if any airborne radionuclides were released from these tanks or filter demineralizers, there would be no effect on normally occupied areas due to the HVAC system design features, which are explained in Section 12.2.2.3.1.

12.2.2.3.3 Effect of Relief Valve Exhaust The relief valves found in the various plant systems which can exhaust radioactive fluids are not considered to be a significant source of airborne radioactivity in normally occupied areas.

The reasons are as follows:

a.

All relief valves (except the main steam safety relief valves), which relieve pressure in the turbine main steam or bleed systems, exhaust directly to the condenser, and

b.

All relief valves, which relieve pressure in a system which contains radioactive fluid, either exhaust directly into the suppression pool, an equipment or floor drain, or into a line which is part of the system in question.

With reference to the equipment or floor drain systems, the receiving point for relief valve discharges is in a radiation zone equivalent or higher than the equipment being relieved. For discharge back to the system, the same is true.

The main steam safety relief valves, which are located within primary containment, provide pressure relief to the main steam lines coming from the reactor. These valves exhaust directly to the suppression pool. The effect of main steam relief valve blowdown in normally occupied areas in secondary containment is analyzed by assuming that all radionuclides that are present in the main steam blowdown are released to the primary containment air. The radionuclide distribution within the free volume of the primary and secondary containment is assumed to be uniform. The radionuclide concentration in secondary containment becomes, in mCi/cm3:

C R q t A q

R V R

q V

t sc b

i v

i v

sc

,i (exp

(

)

/

) )

=

+

+

b sc i

t - exp - (

(12.2-3) where:

R

= primary containment leakage constant (1/minute) qb

= main steam blowdown flow (g/minute)

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-12 tb

= duration of blowdown flow (minute) qv

= ventilation flow rate out of secondary containment (cm3/minute)

Vsc

= volume of secondary containment (cm3) i

= decay constant for isotope i (1/minute) t

= time after blowdown event Csc,i

= airborne radionuclide concentration of radionuclide i in the secondary containment (µCi/cm3)

Ai

= radionuclide concentration in blowdown fluid (µCi/g)

The value of t which yields the maximum value of Csc,i is t

R q

V n

R q

V v

sc i

i v

sc

=

+

+

1 1

/

/

(12.2-4)

The calculated results are based on the occurrence of a main steam isolation valve closure.

This results in all 18 relief valves being actuated for a maximum duration of 40 sec. This event results in the maximum release of radionuclides to primary containment (see Table 5.2-3 and Sections 15.1 and 15.2). The values of the various parameters used in equations 12.2-3 and 12.2-4 are given as follows:

R = 0.5 vol. %/day (Section 3.8.2.3-1) qb = 1.6 x 107 lb/hr = 1.2 x 108 g/minute (Table 5.2-3) tb = 40 sec = 0.67 minute (Table 5.2-3) qv = 9.5 x 104 cfm (Table 11.3-6)

Vsc = 3.5 X 106 ft3 (Table 11.3-6)

The values of Ai are based on the information found in Section 11.1.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-13 The maximum airborne concentration of the most important radionuclides in secondary containment from a main steam relief valve blowdown is listed in Table 12.2-16. The concentrations are far below the DAC criteria given in 10 CFR Part 20. It is concluded that the main steam relief valve blowdown has a negligible effect on the airborne radionuclide level in secondary containment.

12.2.2.3.4 Effect of Removing Reactor Pressure Vessel Head and Associated Internals Experience at BWR plants has shown that an inventory of radioactive gases will accumulate inside the reactor vessel head between the time of shutdown and head removal. These gases consist primarily of the longer lived radioiodines and noble gases. To prevent these gases from being released to the refueling area, provisions are made so that the gases are vented to the HVAC system prior to reactor pressure vessel (RPV) head removal. These provisions include the ability to vent the RPV head through a 4-in. flexible hose connection to the reactor building sump vent filter units which contain particulate and charcoal filter components. The connection to the sump vent filter unit is shown in Figure 9.4-2.

Provisions are also made for clean water services to the RPV cavity area. This permits wetting of the RPV cavity and components to minimize airborne contamination. This is done prior to flooding the RPV cavity.

It is anticipated that RPV head and reactor internals removal will have a minimal effect on the airborne radionuclide level in the spend fuel area.

12.2.2.3.5 Effect of Sampling The possibility of releasing radionuclides which could become airborne during sampling operations is recognized. Design features are incorporated into the sample system to limit the radionuclide release. Radioactive liquids that require frequent grab sampling are piped via sample tube lines to fume hoods located in sample rooms. Grab sampling will normally be accomplished in the fume hoods. During sampling, an inflow face air velocity of approximately 100 ft/minute will be maintained to sweep any airborne radioactive particles to the exhaust duct. Administrative control is further exercised in the form of procedures that are followed when process fluids are sampled. This minimizes the release of radioactive fluids and, hence, exposure to personnel during the sampling process.

12.2.2.3.6 Effect of Spent Fuel Movement Experience at four operating BWR plants has shown that fuel movement does not present any unusual radiological problems. The expected level of radioactivity in the spent fuel pool water is listed in Table 3.5-7 of the Environmental Report. This activity is not expected to have any significant radiological effects on the airborne radionuclide levels above the spent fuel pool floor.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-14 12.2.2.3.7 Effects of Solid Radwaste Handling Areas The solid radwaste handling equipment contained Zone IV area between columns N.1 and S (Figure 12.3-11) are designed for remote operation. Entry for maintenance activities will normally entail shut down and flushing of systems and equipment. The solid radwaste handling equipment is operated primary from low background areas of Zone III between columns Q.1 and R.2 near the waste compactor.

The ventilation supply to this Zone III area is clean outside air with air flow into surrounding normally unoccupied areas. The only source of airborne radioactivity, the waste compactor, is vented directly to the filtered exhaust ventilation system.

Airborne radioactivity levels in the normally occupied solid radwaste handling areas will be ambient outside air concentrations.

12.2.2.3.8 Effects of Liquid Radwaste Handling Areas Normally occupied liquid radwaste handling areas include the valve corridor (a Zone III area),

the precoat room and the radwaste control room (Zone II areas) shown in Figure 12.3-12.

This valve corridor is supplied directly with outside air. Components that are operated from this corridor contain radioactive materials which are located in normally unoccupied valve and pump rooms which are served by separate ventilated supply and exhaust. The radwaste control room and the precoat rooms do not house components containing radioactive material.

Although not normally occupied, the possibility exists that entry into pump corridor (a Zone IV area between columns 11.2 and 12.2) (Figure 12.3-11) and valve and pump (East and West side) rooms (Zone IV areas of Figure 12.3-12) could be necessary while systems are operating.

The pump corridor of Figure 12.3-11 contains the highest concentrations of radioactive material and has the lowest air flow to volume ratio, so is taken as a worst case. Equilibrium airborne radioactivity concentration is calculated as described in Section 12.2.2.2 assuming a leak rate of 50 gal/day from the reactor water cleanup phase separator decant pump. The results are shown in Table 12.2-17.

12.

2.3 REFERENCES

12.2-1 Aalto, E. et al., A Users Manual for the NRN Shield Design, AE-145, June 1964.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-15 12.2-2 Perkins, J. F. and King, R. W., Energy Release from Decay of Fission Products, Nuclear Science and Engineering, Vol. 3, 1958 and Perkins, J. F.,

U.S. Army Missile Command Redstone Arsenal, Report No. RR-TR-63-11, July 1963.

12.2-3 Malenfant, Richard E., QAD: A Series of Point-Kernel General Purpose Shielding Program, National Technical Information Service, LA-3573, October 1966.

12.2-4 Butrovich, R. et al., Millstone Nuclear Power Station, Refueling/Maintenance Outage, GE NEDO-20761, Class 1, Fall 1974.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-17 Table 12.2-1 Basic Reactor Data for Source Computations (During Plant Operation)

Reactor thermal power 3486 MW Overall average core power density 51.6 w/cm3 Core power peaking factors At core center:

Pmax Pave Z (axial) 1.5 Pmax Pave R (radial) 1.4 At core boundary:

Pmax Pave Z (axial) 0.5 Pmax Pave R (radial) 0.7 Core volume fractions:

Material Density (g/cm3)

Volume Fraction UO2 10.4 0.254 Zr 6.4 0.140 H2O 1.0 0.274 Void 0

0.332 Average water density between core and vessel below the core 0.74 g/cm3 Average water-steam density above core In the plenum region 0.23 g/cm3 Above the plenum (homogenized) 0.6 g/cm3 Average steam density 0.036 g/cm3

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-18 Table 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary Energy Range (MeV)

Neutron Flux (Neutrons/cm2 sec) 17.9-14.0 7.13E9 14.0-12.0 2.37E9 12.0-10.0 1.79E10 10.0-9.0 2.37E10 9.0-8.0 4.69E10 8.0-7.0 1.17E11 7.0-6.0 3.45E11 6.0-5.0 6.57E11 5.0-4.0 1.23E12 4.0-3.0 2.34E12 3.0-2.5 2.04E12 2.5-2.0 1.27E12 2.0-1.5 2.97E12 1.5-1.0 5.63E12 1.0-0.7 3.18E12 0.7-0.5 3.92E12 0.5-0.3 4.15E12 0.3-0.1 5.62E12 0.1-0.03 3.50E12 0.03-0.01 2.31E12 1.0(-2)-1.0(-3) 3.76E12 1.0(-3)-1.0(-4) 3.07E12 1.0(-4)-1.0(-5) 2.40E12 1.0(-5)-1.0(-6) 1.94E12 1.0(-6)-1.0(-7) 1.50E12 1.05(-7)-thermal 2.58E12

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-19 Table 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation Energy Range (MeV)

Mid-Range Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 10.0-7.0 8.5 5.97E11 7.0-5.0 6.0 4.00E11 5.0-3.0 4.5 4.77E12 3.0-2.0 2.5 6.83E12 2.0-1.0 1.5 1.02E13 1.0-0.0 0.5 1.25E13

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-20 Table 12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown Energy Range (MeV)

Average Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec)

>2.60 3.00 8.70E11 2.60-2.20 2.40 4.24E11 2.20-1.80 2.00 2.63E11 1.80-1.35 1.58 1.15E12 1.35-0.90 1.13 1.40E12 0.90-0.40 0.65 1.46E12 0.40-0.00 0.20 2.83E11

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-21 Table 12.2-5 Fission Product Source in RHR Piping and Heat Exchangers 4 Hours After Shutdown Energy Range (MeV)

Average Energy (MeV)

Energy Release (MeV/cm3 sec) 4.00-3.00 3.50 2.30E2 3.00-2.60 2.75 3.70E2 2.60-2.20 2.40 8.20E2 2.20-1.80 2.00 1.30E3 1.80-1.35 1.58 4.80E3 1.35-0.90 1.13 5.00E3 0.90-0.40 0.65 9.20E3 0.40-0.10 0.25 1.80E3

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-22 Table 12.2-6 Gamma Ray Energy Spectrum For Spent Fuel Sources (One Core)

Energy Range (MeV)

Average Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 2 Days After Shutdown

>2.60 3.00 2.54E8 2.60-2.20 2.40 9.89E9 2.20-1.80 2.00 4.93E9 1.80-1.35 1.58 1.34E11 1.35-0.90 1.13 2.59E10 0.90-0.40 0.65 2.97E11 0.40-0.00 0.20 5.17E10

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-23 Table 12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater Component Radioactivity Concentration (Ci/cm3)

Moisture separators and reheaters (MSR)

Plena 3.48E-7 First stage reheater tube bundle (east end of MSR) 7.23E-7 First stage reheater tube bundle (west end of MSR) 5.91E-7 Second stage reheater tube bundle (east end of MSR) 1.43E-6 Second stage reheater tube bundle (west end of MSR) 1.14E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-24 Table 12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell Group Average Group Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 1 3.50 3.82E1 2

2.80 7.92E1 3

2.40 1.43E2 4

2.00 1.24E2 5

1.57 3.94E2 6

1.12 3.00E2 7

0.65 6.71E2 8

0.20 8.26E1

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-25 Table 12.2-9 Nitrogen-16 Source Strength in Feedwater Heater 6 Radionuclide Concentration (Ci/cm3)

Feedwater Heater Steam Water 6

4.93E-7 8.40E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-26 Table 12.2-10 Nitrogen-16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems Point of Interest Line Source (Ci/cm)

Input lines to high pressure turbine 7.60E-3 Extraction steam line from high pressure turbine to FWH 6A 7.50E-4 Crossunder piping from high pressure turbine to main steam relief 5.50E-4 Crossover piping from main steam relief to low pressure turbine 3.80E-4 Extraction steam line from low pressure turbine to FWH 1A 7.50E-5 Extraction steam line from low pressure turbine to FWH 2A 1.20E-4 Extraction steam line from low pressure turbine to FWH 3A 1.40E-4 Extraction steam line from low pressure turbine to FWH 4A 1.50E-4 Heater drain line from FWH 6A to FWH 5A 2.30E-5 Heater drain line from FWH 5A to FWH 4A 1.01E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-27 Table 12.2-11 Offgas System Sources in the Turbine Generator Building Component 16N Source Strength

(µCi/cm3)

Steam jet air ejector (first stage) 2.3E0 Intercondenser 4.2E1 Steam jet air ejector (second stage) 3.6E0 Preheater 2.8E0 Recombiner 2.3E0 Offgas condenser 3.7E1 Water separatora 2.7E1 a The preheater, recombiner, offgas condenser, and water separator are located in the same room.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-28 Table 12.2-12a Special Sources With Strength Greater Than 100 Millicuries Isotope Identification Form Quantity (mCi)

Use/Location 241AmBe 2-81-020 Solid 14,770 Calibration (EOF) 241AmBe 2-82-050 Solid 2,845 Neutron source (plant) 241AmBe 2-86-073 Solid 988 WNP-1 source (plant) 137Cs 2-79-033 Solid 524 Panoramic shepard cal (EOF) 137Cs 2-83-097 Solid 4,673 ARM calibration (plant) 137Cs 2-84-058 Solid 1,474 Shepard series 28 cal (EOF) 137Cs 2-88-002 Solid 133 Victoreen model 878-W cal (plant) 137Cs 2-93-026 Solid 909 MG calibrator (EOF) 137Cs 2-99-028 Solid 2,078 Mini-89 calibrator (plant) 137Cs08-132 Solid 358,600 Hopewell calibrator (EOF) 137Cs08-133 Solid 422 Hopewell calibrator (EOF) 137Cs13-230 Solid 12,940 ARM calibration (plant) 238PuBe 2-84-047 Solid 11,150 WNP-3 startup source (plant) 238PuBe 2-84-048 Solid 11,260 WNP-3 startup source (plant)

Table as of 9/9/2015.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-28a Table 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area Area Location Approximate Size (sq. ft.)

Normal Contents Normal Activity (mCi)

LSA Storage Pad Outside SW side of Radwaste Bldg 1800 Low-activity dry and compacted solid wast containers 930 Building 13 West portion of Bldg 13 2700 Contaminated protective clothing 300 Warehouse 5 NE portion of Bldg 80 at Snake River Warehouse Complex 4000 Radioactive &

contaminated equipment 590 Building 167

~0.5 miles E of Plant 6332 Radioactive &

contaminated equipment 1370 Building 167 Storage Yard

~0.5 miles E of Plant 68000 C-van storage 1620 Outage Storage Yard Fenced area W of Bldg 105 2400 Outage radioactive material & C-van storage 114 Kootenai HP Calibration Lab Kootenai (Bldg

34) Rms 102 &

102A 600 Calibrators/irradiators, calibration sources, radioactive HP instruments 377030

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-29 Table 12.2-13 List of Radioactive Piping and System Designations Air removal (AR)

Bleed steam (BS)

Condensate filter/demineralizer (CPR)

Condenser vents and drains (CND)

Control rod drive (CRD)

Equipment drains radioactive (EDR)

Exhaust steam (ES)

Floor drains radioactive (FDR)

Fuel pool cooling (FPC)

Heater drains (HD)

Heater vents (HV)

High pressure core spray (HPCS)

Low pressure core spray (LPCS)

Main condensate before condensate demineralizers (COND)

Main steam (MS)

Main steam isolation valve leakage control system (MSLC)

Miscellaneous waste radioactive (MWR)

Offgas (OG)

Process sample radioactive (PSR)

Process vents (PVR)

Process waste radioactive (PWR)

Reactor core isolation cooling (RCIC)

Reactor recirculation (RRC)

Reactor water cleanup (RWCU)

Relief valve vents radioactive (VR)

Residual heat removal (RHR)

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-03-040 12.2-30 Table 12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building)

Radionuclide Airborne Concentration Ci (Ci/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 1.9E-11 2E-8 1E-3 132I 2.5E-12 3E-6 1E-6 133I 2.2E-11 1E-7 2E-4 134I 1.9E-12 2E-5 2E-7 135I 8.7E-12 7E-7 1E-5 83Br 3.3E-13 3E-5 1E-8 84Br 6.3E-14 2E-5 3E-9 85Br 1.3E-16 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-31 Table 12.2-15 Airborne Radionuclide Concentration in Condensate Pump Area (el. 441 ft. 0 in. turbine generator building)

Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 4.2E-10 2E-8 2E-2 132I 3.8E-9 2E-6 3E-3 133I 2.9E-9 1E-7 2E-2 134I 7.4E-9 2E-5 4E-4 135I 4.2E-9 7E-7 6E-3 83Br 4.8E-10 3E-5 2E-5 84Br 8.2E-10 2E-5 4E-5 85Br 3.2E-10 1E-7 3E-3 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-32 Table 12.2-16 Airborne Radionuclide Concentration in Secondary Containment from a Main Steam Relief Valve Blowdown Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 3.0E-11 2E-8 2E-3 133Xe 5.0E-10 1E-4 5E-6 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-33 Table 12.2-17 Airborne Radionuclide Concentration in Liquid Radwaste Handling Area Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 140Ba 5.8E-10 6E-7 1E-3 140La 6.5E-10 6E-7 1E-3 239Np 2.2E-10 9E-7 2E-3 58Co 9.8E-10 3E-7 3E-3 89Sr 4.8E-10 6E-8 1E-2 99Mo 2.6E-10 6E-7 4E-4 99MTc 1.7E-10 6E-5 3E-6 132Te 1.5E-10 9E-8 2E-3 131I 9.2E-10 2E-8 4E-2 132I 2.4E-10 3E-6 1E-4 133I 4.1E-10 1E-7 4E-3 135I 1.8E-10 7E-7 2E-4 a 10 CFR 20.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-037 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES Columbia Generating Station plant incorporates the design objectives and the design features guidance given in Regulatory Guide 8.8 to the extent discussed in Section 12.1.1. Examples of these features are discussed in Section 12.3.1.3.

Figures 12.3-5 through 12.3-18 show the general arrangement for each of the plant buildings.

In addition, these figures show the shielding arrangement, radiation zone designations for both normal operation and shutdown conditions, controlled access areas, personnel and equipment decontamination areas, location of the health physics facilities, location of area radiation monitors, location of radwaste control panels, location of the counting room, and location of the onsite laboratory for analysis of chemical and radiochemical samples. The counting room is located at el. 487 ft 0 in. of the radwaste and control building (see Figure 12.3-13). The design basis radiation level within the counting room is 0.1 mrem/hr during normal operation.

Plant areas, as identified in Section 12.3 figures, are categorized as design radiation zones according to expected maximum radiation levels and anticipated personnel occupancy with consideration given toward maintaining personnel exposures ALARA and within the standards of 10 CFR 20.

12.3.1.1 Radiation Zone Designations The design basis criteria used for each zone are given below, and the plant layout including major equipment, locations, and radiation zone designations are shown in Figures 12.3-5 through 12.3-18.

For purposes of radiation exposure control, a restricted area is defined in 10 CFR 20, paragraph 20.1003, and plant procedures.

Maximum Dose Rate Zone (mrem/hr)

Design Bases Criteria I

1.0 Unlimited occupancy.

II 2.5 Unlimited occupancy for plant personnel during the normal work week.

III 100.0 Design base occupancy less than 1 hr per week.

Posted zones and controlled entries.

IV Unlimited Positive access control. Controlled entry and occupancy.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-037 12.3-2 Each access point to every Zone IV area may be secured by locked door or other positive control method while it is a high radiation area. Occupancy of such areas is limited in frequency and duration and entrance must be authorized in advance by the Radiological Operations Supervisor or his representative.

An area survey of radiation levels will be conducted prior to first entry of Zone IV areas to determine the maximum habitation time.

12.3.1.2 Traffic Patterns Access control and traffic patterns in the plant have been evaluated to maintain personnel radiation exposures ALARA and to minimize the spread of contamination.

Normal entry into the plant is as follows:

a.

Personnel normally access the plant Protected Area through the security Protected Area Access Point (PAAP).

b.

The main Radiologically Controlled Area (RCA) normally includes the reactor building, turbine generator building, radwaste building, and diesel generator building. Normal access to these areas is through one of two Health Physics control points located at each end of the main plant corridor.

12.3.1.3 Radiation Protection Design Features Section 12.1.2 discusses the design objectives for the plant. Examples showing the incorporation of these objectives are provided in the following sections.

12.3.1.3.1 Facility Design Features Filters and Demineralizers Liquid radioactive waste and other process streams containing radioactive contaminants are processed through filters and demineralizers. The pressure-precoat type of filter is used in the major fluid processing systems. Cartridge type filters are used in a few select instances such as in the control rod drive (CRD) system. In the case of demineralizers, either the pressure-precoat or deep-bed type of demineralizer is employed.

Each filter and demineralizer is located in a shielded cubicle at el. 487 ft 0 in. of the radwaste and control building. These filters and demineralizers are accessible through the ceiling of the cubicle by removing the shielding plug at el. 507 ft 0 in. This minimizes the radiation

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 12.3-3 exposure to plant personnel from adjacent sources. After removal of the shielding plug, the filter or demineralizer can be serviced remotely by use of special tools designed for this purpose. If it is necessary, the filter or demineralizer can be removed from the cubicle to a work area for servicing. There are overhead cranes provided for the purpose of shielding plug and filter or demineralizer vessel removal.

Each pressure precoat type filter or demineralizer has its own support equipment such as a holding pump, process control valves, and precoating equipment. The holding pump and process control valves associated with each filter or demineralizer is located in the valve gallery. This valve gallery is located on the east side of the radwaste-control building, el. 467 ft 0 in., and is a Zone III radiation area (Figure 12.3-12).

The holding pump and motor-operated valves can be operated from control panels located in Zone III radiation areas. Manually operated process control valves are operated by use of reach rods that pass through the shielding walls into the corridor. This corridor is a Zone III radiation area. With the exception of instrument root valves, all pumps and valves can be remotely operated. Instrument root valves are normally open and are not used during normal plant operation. The filter or demineralizer precoat equipment and associated controls, which includes metering equipment, are located in a Zone III radiation area to provide for maintenance access. Each filter or demineralizer may be backwashed with condensate water, chemically cleaned, or purged with air from a remote control panel. Each deep-bed demineralizer has its own support equipment. A gravity feed subsystem transfers the fresh resins from the resin addition tank to the demineralizers. The spent resins are transferred hydropneumatically to the spent resin tank.

All piping routed to and from filter or demineralizer cubicles is located in the valve gallery area or in shielded pipe tunnels. The pipes are routed to avoid unnecessary bends, thus minimizing possible crud buildup.

Specific examples of filters or demineralizers that incorporate the aforementioned design features are the waste collector filter and waste collector demineralizer. A typical layout is shown in Figure 12.3-19.

Tanks All tanks that contain radioactive liquids and solids are categorized into different groups depending on their level of radioactivity. Tanks that contain significant levels of radioactivity are located at el. 437 ft-0 in. of the radwaste and control building.

The tanks that contain radioactive backwash and resins are located in shielded cubicles. These include the condensate phase separator tanks-condensate backwash receiving tank, spent resin tanks waste sludge phase separator tanks, and the reactor water clean up (RWCU) phase

COLUMBIA GENERATING STATION Amendment 55 FINAL SAFETY ANALYSIS REPORT May 2001 12.3-4 separator tanks. These tanks are constructed of either stainless steel or epoxy-lined carbon steel.

The tanks that contain unprocessed liquid radioactive waste, such as the waste collector, floor drain collector, and chemical waste tanks are not located in individually shielded cubicles.

However, as described in Section 12.1.1, administrative controls will be used to keep doses to plant personnel ALARA. The waste collector and the floor drain collector tanks are constructed of carbon steel. The chemical waste tanks are stainless steel.

To measure the liquid level in the aforementioned tanks, a bubbler-type level indicator, coupled with a pressure transducer, provides for remote monitoring of the liquid level.

All tanks described above are vented to the radwaste building heating, ventilating, and air conditioning (HVAC) exhaust system as described in Section 12.2.2. This limits the release of airborne radionuclides to the area surrounding the tank.

Pumps Pumps handling spent demineralizer resins are shielded from the phase separator tank. An example of this is the waste sludge discharge pump which is shielded from the waste sludge phase separator by concrete and block wall. A shutoff valve in the suction line keeps the pump isolated when it is not in use. Following the discharge of a sludge batch from the phase separator tanks, the pump and its associated piping is automatically flushed with condensate water. Thus, when it is not in use, the pump is free of sludge. A condensate supply, controlled by a valve electrically linked to the pump, maintains an external sealing barrier, preventing sludge leakage past the shaft seal during pump operation.

Heat Exchangers Heat exchangers handling radioactive fluids are designed to limit occupational exposures. An example is the cooler condensers whose function is to condense moisture from the offgas process stream. The cooler condensers are located in a separate cell in the radwaste building, as shown in Figure 12.3-11. No personnel access is required during plant operation. The associated valves, other than those used during maintenance operations, are remotely operated.

The glycol coolant which flows through the tube side is kept at a higher pressure than the offgas flowing through the shell side of the condenser. Thus, a tube failure will not result in radioactivity contaminating the glycol cooling loop. Condensate water that is separated from the offgas is returned, via a water loop seal, to the radioactive sump for processing. The 16-ft deep loop seal prevents escape of radioactive gas through the drain connection. An enlarged discharge section in the loop seal protects it against siphoning. The enlarged discharge section also provides for automatic loop seal restoration should its contents be displaced by a temporary pressure surge. Figure 12.3-20 shows schematically the cooler condenser loop seal arrangement.

COLUMBIA GENERATING STATION Amendment 55 FINAL SAFETY ANALYSIS REPORT May 2001 12.3-5 Recirculation Pumps The decontamination concentrator bottoms recirculation pump serves the decontamination solution concentrator. The pump is made of 316 stainless steel to minimize the corrosive effects of miscellaneous chemical waste fluid. The pump is located in a separate cell from the concentrating equipment it serves. A steam/air sparger at the bottom of the concentrator can be used to flush the pump prior to any maintenance operations. The internals of the pump may be removed without disturbing the connecting piping. A double mechanical seal with clean water circulating through it prevents leakage of process liquid past the shaft seal.

The decontamination concentrator bottoms recirculation pump is not used. There are no plans to use the pump.

Evaporators The decontamination solution concentrators use steam to boil off water from the incoming chemical waste, as described in Section 11.2. The heating steam is supplied from an evaporator. As shown in Figure 12.3-21, steam generated from demineralized water flows in a closed loop through the shell side of the evaporator and the shell side of the concentrator heating element. The steam is then circulated to the condensate return tank to be pumped back to the evaporator. The steam in the concentrator heating element is at a higher pressure than the decontamination solution flowing through the tube side. Furthermore, the auxiliary steam system, which provides heating steam to the tube side of the evaporator, is at a higher pressure than the shell side. This arrangement provides an effective barrier against contamination of the house auxiliary steam.

The decontamination solution evaporator system is deactivated. There are no plans to use the system.

Valve Gallery and Valve Operating Stations Valves handling radioactive fluids and requiring manual operation are located in a valve gallery at el. 467 ft 0 in. of the radwaste and control building. These valves are operated from behind a shielding wall with the aid of reach rods. Reach rod penetrations are not in the line-of-sight with major radiation sources, such as resin traps. In addition, the reach rod wall penetrations are grouted about the reach rod assembly, and steel plates are added on both sides of the penetration to minimize radiation exposure. A typical application of reach rods and details of a shielding wall penetration is shown in Figure 12.3-19.

The operating stations for motor-operated valves are located in Zone III radiation areas.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-05-002,05-007 12.3-6 Sampling Areas The location of the sampling areas within the plant is discussed in Section 9.3. Design features of sample areas that reduce occupational exposure are discussed in Section 12.2.2.3.5.

Ventilation Filters and Filter Trains Filters that are installed as part of the HVAC units in the Columbia Generating Station plant are located in an accessible area. Selected filter units are designed so that it is possible to introduce a plastic bag over the filter through a service access opening in the HVAC unit. For other filter units, the filter can be pulled into a plastic bag as the filter is removed from the HVAC unit.

Hydrogen Recombiners The hydrogen recombiners for the offgas system are located in the turbine-generator building.

These recombiners are single-pass devices which do not require process control valves. They are located in a shielded cell and do not require personnel access during operation.

Temperature and pressure in the recombiners are remotely monitored. The recombiners and associated piping are designed to withstand an internal explosion.

12.3.1.3.2 Design Features That Reduce Crud Buildup Design features and considerations are included to reduce radioactive nickel and cobalt production and buildup. For example, the primary coolant system consists mainly of austenitic stainless steel, carbon steel, and low alloy steel components. Nickel content of these materials is low. Nickel and cobalt contents are controlled in accordance with applicable ASME material specifications. A small amount of nickel base material (Inconel 600) is employed in the reactor vessel internal components. Inconel 600 is required where components are attached to the reactor vessel shell, and coefficient of expansion must match the thermal expansion characteristics of the low-alloy vessel steel. Inconel 600 was selected because it provides the proper thermal expansion characteristics, adequate corrosion resistance and can be readily fabricated and welded. Alternate low nickel materials which meet the above requirements and are suitable for long term reactor service are not available. Hardfacing and wear materials having a high percentage of cobalt are restricted to applications where no alternate materials of equally good characteristics are available.

To minimize crud buildup, lines carrying slurries in the reactor water cleanup system and in the radwaste system use extensively self-flushing valves. Furthermore, all of the pipe joints of 2.5 in. and above in size in the reactor water cleanup (RWCU) and radwaste systems are butt welded. Butt welding is also used in the equipment drain processing (EDR) and floor drain processing (FDR) systems for lines of 1 in. and above, except for two 1-in. flow control valves, one 1.5-in. flow control valve, two 2-in. socket welded ball valves, one 1-in. socket

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-05-007 12.3-7 welded ball valve, and four 3-in. flanged butterfly valves. Lines of 2.5-in. and larger in the offgas system are also butt welded.

The recirculation system is equipped with decontamination flanges for decontamination of the recirculation pump and associated hardware. Also the cleanup, emergency core cooling system (ECCS), and radwaste systems are provided with decontamination connections to enable decontamination of their hardware. In addition, the RWCU and condensate process filter demineralizers reduce the amount of activation products in these systems. Boiling water reactors (BWRs) do not use high temperature filtration.

Depleted zinc is injected into the feedwater system to displace cobalt from the corrosion product layer in the recirculation system. Iron is injected into the feedwater system to cause the displaced cobalt to adhere to the surface of the fuel rods. This has caused a reduction of exposure rates from the recirculation system.

12.3.1.3.3 Field Routing of Piping All code Group A piping is dimensioned in detail. Other piping, 2 in. in diameter and smaller, carrying radioactive fluids, are not dimensioned in detail but are required to be in specified space envelops for shielding and in-plant exposure considerations. Such piping runs have their inception points and terminal points dimensionally located. These inception or terminal points may be an outlet or inlet of a pump, a junction with a larger pipe, or a tank nozzle. Also located dimensionally are all pipe penetrations through a wall, ceiling, or floor. Radioactive piping routed through lower radiation zones is enclosed within a shielded tunnel when warranted by high expected radiation levels. Radioactive piping 2.5-in. in diameter and larger is detailed dimensionally.

12.3.1.3.4 Design Features That Reduce Occupational Doses During Decommissioning Many of the design facilities which presently exist in the plant can be used to minimize occupational exposure during decommissioning, whether decommissioning is accomplished through mothballing, entombment, removal/dismantling, or any combination of the above alternatives. Such facilities include those used for handling and for offsite shipment of fresh fuel, spent fuel, more sources, contaminated filter elements, resins, and other radioactive wastes. The radioactively contaminated spent fuel pool water can be removed by a portable pump placed into the pool after all spent fuel has been removed from the site and any decommissioning use of the pool is finished. The portable pump will discharge into the skimmer surge tank from where the existing decontamination procedure can be implemented.

The number of man rems due to the airborne radioactivity, that may be introduced by the handling of radioactively contaminated systems, as well as the number of man rems due to contact with the same systems, can be reduced by first decontaminating them. Means exist in the present design where radioactively contaminated systems can be decontaminated chemically

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-8 by remote control and flushed. The plant has a hot machine shop and a hot instrument shop located in the radwaste building where contaminated equipment can be decontaminated under controlled conditions. Provisions have been made in the same building for a future decontamination facility with expanded features.

If decommissioning is accomplished by mothballing, the above provisions will reduce to low levels the occupational radiation exposure to personnel. According to Regulatory Guide 1.86, this method involves putting the facility in a state of protective storage. In general, the facility may be left intact except that all fuel assemblies and the radioactive fluids and waste should be removed from the site.

If entombment is chosen as the method of decommissioning, the previously described plant design facilities are adequate to accomplish the tasks with low occupational radiation exposure to personnel. The additional requirements described in Regulatory Guide 1.86 for sealing all the remaining highly radioactive or contaminated components (e.g., the pressure vessel and reactor internals) within a structure integral with the biological shield after having all fuel assemblies, radioactive fluids and wastes, and certain selected components shipped offsite can be met. The necessary modifications to the Columbia Generating Station primary containment structure are shown in Figure 12.3-22.

Low occupational radiation exposure to personnel can be achieved if the decommissioning method adopted is that of immediate removal/dismantling of the plant. For example, the laydown area, el. 606 ft 10.5 in., in the reactor building is provided with a drainage system.

There the drywell head and the reactor pressure vessel head can be decontaminated if it is required. Furthermore, there is ample space to erect a contamination control envelope around them during their segmentation for transportation and burial. The plant design accommodates the cutting of the reactor internal under water and within contamination control envelopes.

The turbidity of the water can be reduced by using the RWCU system. The only modification that might be required to the existing system could be the installation of a strainer for the removal of large filings or other large size contaminants. The highly radioactive pieces can be transferred under water to the cask loading area in the spent fuel pool by methods similar to loading spent fuel. The airborne radioactivity generated by the dismantling process can be removed by the filtration systems that the contaminated control envelopes can have and/or by the standby gas treatment system (SGTS).

12.3.1.4 Radioactive Material Safety 12.3.1.4.1 Materials Safety Program Columbia Generating Station has a program to ensure the safe storage, handling, and use of sealed and unsealed special nuclear source and byproduct materials. Included in the program are procedures for the following:

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-13-039 12.3-9

a.

Receiving and opening shipments as required by 10 CFR 20.1906,

b.

Storage of licensed materials as required by 10 CFR 20.1801 and 20.1802,

c.

Inventory and control of radioactive materials,

d.

Posting of radioactive material storage areas and tagging of source,

e.

Leak tests - sources are checked for leakage or loss of material at least semiannually, and

f.

Disposal - all licensed material disposals are in accordance with 10 CFR Part 20 requirements or by transfer to an authorized recipient as provided in 10 CFR Parts 30, 40, or 70.

12.3.1.4.2 Facilities and Equipment Facilities are provided for handling unsealed sources, such as the liquid standard solutions used for calibration of plant instrumentation. The radiochemical laboratory is equipped with a negative pressure fume hood with filtered exhaust. The hood work surface is designed to withstand heavy weights so that shielding can be provided in the form of lead brick. Drains from the fume hood are routed to the liquid radwaste system.

Remote handling tools are used as needed for movement of the high level sealed sources from their normal storage containers. Shielding to reduce personnel exposure is provided for these sources when they are not in use and to the extent practicable while they are in use.

Portable radiation and contamination monitoring instrumentation is provided, as described in Section 12.5.2, for surveillance to maintain control of the sources.

12.3.1.4.3 Personnel and Procedures The Columbia Generating Station Radiological Services Manager/Radiation Protection Manager (RPM) is responsible for the control and monitoring of sealed and unsealed source and byproduct materials. The Nuclear Material Manager appointed by the Engineering Manager is accountable for special nuclear materials (SNM). The Chemistry Technical Supervisor is responsible for the minimization of radioactive waste and the preparation, offsite shipment, and disposal of radioactive materials and radwaste. Monitoring during handling of these materials is provided by Radiation Protection. Experience and qualifications of Radiation Protection personnel are described in Section 13.1.

Health Physics requirements and instructions to personnel involved in handling byproduct materials are included in implementing procedures.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-10 12.3.1.4.4 Required Materials Any byproduct, source, or special nuclear materials in the form of reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrument and radiation monitoring equipment calibration, or as fission detectors, will be limited to the amounts required for reactor operation or specific calibration purposes except as noted in the facility operating license.

12.3.2 SHIELDING 12.3.2.1 General The radiation shielding design is in compliance with all NRC regulations concerning permissible radiation doses to individuals in restricted and nonrestricted areas. The guidance provided in Regulatory Guide 1.69 on concrete radiation shields is followed to the extent discussed in Section 1.8. The guidance provided in Regulatory Guide 8.8 on radiation protection is followed to the extent discussed in Section 12.1.2. The plant design and layout optimizes personnel and/or equipment protection. Shielding is supplemented with whatever administrative control procedures are necessary to ensure regulatory compliance. These control procedures include area access restrictions, occupancy limitations, personnel monitoring requirements, and radiation survey practices. Other criteria and considerations are listed in Section 12.1.2.

The shielding design is evaluated under the following conditions of plant operation:

a.

Operation at design power, including anticipated operational occurrences,

b.

Shutdown conditions, with radiation from the subcritical reactor core, spent fuel assemblies, and other sources discussed in Section 12.2, and

c.

Postaccident conditions, including those accident occurrences analyzed in Chapter 15. Emphasis is placed on control room habitability.

The majority of the shielding calculations performed are of the bulk shielding type.

Ordinary concrete, having a density of about 150 lb/ft3, is used for shielding except for special applications. In special applications, water, steel, high density concrete, lead, and permali JN P/3% boron are used.

The effects of mechanical or electrical penetrations in shield walls on radiation exposure to personnel is minimized by locating penetrations to preclude direct view of radiation sources through the penetration. The effect of penetrations in shield walls is also minimized by keeping penetration openings to the smallest practicable size. Penetrations are located away

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-11 from immediate areas with personnel access. When these criteria cannot be implemented, penetrations are offset.

Access into shielded areas is, in general, by labyrinths. Labyrinths are located to preclude direct personnel radiation exposure. Where labyrinths are not practicable, shield doors are used. Knock-out walls for equipment removal are constructed of brick arranged in staggered rows to preclude direct streaming.

Portable and removable shielding devices are used when practical and feasible. Portable shielding devices are easily moved from one location to another. Removable shielding devices are normally used at specific locations and can be removed when necessary. The reactor vessel to fuel pool transfer passage is a location where removable shielding is employed primarily for the protection of personnel working in the drywell. Personnel evaluation of the affected drywell area may be employed instead of, or in conjunction with, the above mentioned shielding.

12.3.2.2 Methods of Shielding Calculations Standard methods are used in computing the required shielding thickness for a given source.

These methods are described in References 12.3-1 through 12.3-4. Specific methods of calculation and the computer codes used in the shielding design are discussed below.

The NRN computer code (Reference 12.3-5) is used to determine the shielding requirements for the core generated neutrons and to calculate the thermal neutron flux used to determine captured gamma sources outside the core. This code is based on a multigroup slowing down and diffusion system corrected by a multigroup first flight or removal neutron source. The neutron cross sections used with this code are from Oak Ridge National Laboratory with modifications which have resulted from comparisons with data in BNL-325 (Reference 12.3-6) and the ENDF/B data libraries.

The QAD-BR computer code, which is based on the QAD-P5 code (Reference 12.3-7), is the basic code used to determine shielding requirements for gamma ray sources. This code provides gamma flux, dose rate, energy deposition, and other quantities which result from a point by point representation of a volume distributed source of radiation. Attenuation coefficients for water, iron, and lead, used in this program are taken from the Engineering Compendium for Radiation Shielding (Reference 12.3-1). Concrete attenuation coefficients are taken from ANL-6443 (Reference 12.3-8).

Any shielding requirements which are not determined with the QAD code were determined by using the methods discussed in the Reactor Shielding Design Manual (Reference 12.3-2). The various sources are reduced to their basic geometric configuration (line, disc, cylinder, sphere, etc.) and the corresponding equations are solved to find the dose. The Taylor exponential form

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-12 of the buildup factor is used in these equations. All required data is taken from Reference 12.3-1.

The criteria for penetration acceptability is based on the radiation zone levels in the areas separated by the wall where the penetration is located. The penetration is analyzed if the penetration passes from Radiation Zone IV to Zones III, II, or I; or, if the penetration passes from Radiation Zone III to Zones II or I. This analysis considers both the scattered as well as the direct radiation. The direct component is calculated using the point source attenuation equation, as described in the Reactor Shielding Design Manual (Reference 12.3-2). The scattered component is calculated using the Chilton-Huddleston equations (Reference 12.3-9).

Compensatory shielding (e.g., labyrinths, steel plate, lead wool) is used as needed, to reduce radiation streaming through penetrations and to protect against localized hot spots.

The general dose rate in each plant area, including contributions from radiation streaming, satisfies the design dose rate specified for that area.

Entrances to shielding cubicles are designed to prevent source radiation from passing directly through the opening. Whenever possible, a labyrinthine entrance is designed to reduce the emerging radiation to a level compatible with the access requirements outside the cubicle. The scattered and direct components of the emerging radiation are calculated by the above methods.

12.3.2.3 Shielding Description 12.3.2.3.1 General The description of the shielding throughout the entire plant is summarized within the following sections. These descriptions are to be used in conjunction with the radiation zone maps, Figures 12.3-5 through 12.3-18, to locate the process equipment which is shielded and to determine the design dose rate.

12.3.2.3.2 Reactor Building The sacrificial shield is an ordinary concrete structure 2 ft thick, lined on the inside and outside by steel plates of a minimum thickness of 0.5 in. each. The sacrificial wall extends between el. 519 ft 2.25 in. and 567 ft 4.5 in.

The biological shield wall protects station personnel in the reactor building from radiation emanating from the reactor vessel. The dose rate at the outer face of the biological shield as well as above the shield plug (above the reactor vessel) is, except at penetrations, less than 2.5 mrem/hr during normal reactor operation. The reactor core is the primary source of radiation, and it is used in computing the above dose rate. The wall is in the shape of a shell of the frustum of a cone, and its composition is ordinary concrete at least 5 ft thick. Inside the biological wall exists the primary containment vessel which has the same shape as the wall.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-13 The primary containment vessel is made of 0.75 in. minimum steel plate. The 16N contained in the recirculation system, the main steam lines, and the water in the vessel below the core along with the fission process in the core constitute the major sources of radiation used to determine the radial dose rate. The shielding arrangement for the other major sources in the reactor building is shown on Figures 12.3-15 through 12.3-18.

Personnel evacuation of the affected drywell area(s) and/or employing removable shielding at the fuel pool passage are two methods used for personnel protection in the drywell during fuel handling operations. The shielding is designed such that radiation levels are no greater than 100 R/hr at contact. Portable locally alarming radiation monitors and/or direct Health Physics monitoring are employed for additional personnel protection.

12.3.2.3.3 Turbine Building In the turbine building, 16N constitutes the major source of radiation and basis for shielding design. It is contained in the turbines, moisture separator reheaters, and the feedwater heater system that are located in the turbine access areas at el. 501 ft 0 in. and 471 ft 0 in. These areas are surrounded by ordinary concrete walls at least 3.5 ft thick. The dose rate to the areas outside these walls is less than 2.5 mrem/hr.

The walls which surround the turbine-generator access area at el. 501 ft 0 in. extend up to 524 ft 0 in. This minimizes the effects of direct radiation streaming at the site boundary.

The shielding arrangement for the major sources of radiation in the turbine building, including those discussed above, are shown in Figures 12.3-5 through 12.3-10.

12.3.2.3.4 Radwaste Building The shielding arrangement for the major sources in the radwaste building are shown in Figures 12.3-11 through 12.3-14.

12.3.3 VENTILATION The plant ventilation systems for the different areas of the plant are designed to meet the requirements of 10 CFR Parts 20 and 50. Gaseous wastes will be released in a controlled manner to environs such that during plant operation, onsite and offsite radioactivity levels are ALARA. The design features which limit and reduce the airborne radioactivity are as follows:

a.

In the reactor, radwaste, and turbine generator buildings the air flow is from areas of low airborne radioactivity potential to areas of higher airborne radioactivity potential. This serves to isolate and segregate airborne radioactivity which may be released due to equipment failure or malfunction and leakage from fluid systems;

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-14

b.

To prevent radioactivity buildup, all ventilation air is supplied to the reactor, turbine, and radwaste buildings on a once through basis;

c.

All cubicles housing equipment which handles radioactivity contaminated material are ventilated at a minimum rate of three volume air changes per hour;

d.

All sinks and chemical laboratory work areas where radioactive samples or materials are handled are provided with exhaust hoods to protect operating personnel from airborne contaminants;

e.

All liquid equipment leaks which are potential sources of airborne radioactivity in the reactor building are collected in the reactor building equipment drain system. The drain system is maintained at a negative pressure with respect to the balance of the building to protect operating personnel from airborne leakage generated in the system sumps. All exhaust air drawn from the reactor building equipment drain system is filtered by absolute particulate and charcoal filters.

The particulate and charcoal filters minimize the release of contaminated particulates and iodine; and

f.

The primary containment purge system reduces airborne radioactivity within the drywell to acceptable levels prior to entry of working personnel. The level of radioactivity released to the environment during normal primary containment purge will not exceed the requirements of 10 CFR Part 20 Appendix B. When airborne radiation levels in the primary containment are too high to allow direct purging to the atmosphere through the reactor building exhaust, purge air at a reduced flow rate is passed through the SGTS prior to exhaust. In this latter mode, airborne iodine and particulates are removed from the purge exhaust air prior to release; The air cleaning systems which utilize special filtration equipment to limit airborne radioactive contaminants are

a.

Standby gas treatment system (see Section 6.5),

b.

Control room emergency filtration system (see Sections 9.4 and 6.4),

c.

Reactor building sump vent exhaust filter system (see Section 9.4), and

d.

Radwaste building exhaust filtration system (see Section 9.4).

In addition, small local absolute particulate filters are used to locally filter the effluent from sample sink hoods and chemical hoods. These small filter units are all described in Section 9.4.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-15 The SGTS filter units discussed in Section 6.5.1 and the control room emergency filter units discussed in Sections 6.4 and 9.4.1 are the only engineered safety feature (ESF) filtration systems. A detailed evaluation of these units with respect to Regulatory Guide 1.52 is given in Section 1.8. The balance of the filtration systems meet the intent of Regulatory Guide 1.52 with the following deviations:

a.

Reactor building sump vent filter units These units are composed of demisters, an electric heater, a medium efficiency prefilter, an absolute particulate (HEPA) filter and tray type adsorber filters composed of 2-in. deep charcoal beds in a sheet metal housing. These units do not have absolute particulate filters downstream of the adsorber section as described in Regulatory Guide 1.52.

The 5-ft spacing between filter frames, as required in Regulatory Guide 1.52, is not provided in these units. These units are of low capacity (1000 cfm), and personnel access into the units for servicing and testing is not possible. Access panels are provided for servicing and removal of all unit components. Sufficient space is provided between elements to permit removal of any element without disturbing any other element.

b.

Radwaste building exhaust filter units These three units are composed of medium efficiency prefilters, absolute particulate (HEPA) filters and centrifugal fans in a sheet metal housing.

Unit capacity is 42,000 cfm, which is in excess of the 30,000 cfm maximum capacity recommended in Regulatory Guide 1.52. These units are composed of a 5 filter high by 8 filter wide array. Permanent service platforms are provided on both sides of the HEPA filters at an intermediate height for operating personnel during filter testing and service.

Absolute particulate (HEPA) filter and charcoal adsorber filters will be tested periodically to ensure continued filter efficiency as discussed in Sections 6.5.1.4, 9.4.2.4, and 9.4.3.4.

Figure 12.3-23 (el. 572 ft 0 in.) shows the layout of the SGTS filter units. The concrete wall between the units serves as both a fire wall and missile barrier between the two filter trains.

Access doors, 20 in. x 50 in., are provided into each plenum section between unit elements.

Ample aisle space is provided outside the units for ease of access for personnel to perform tests and maintenance. Charcoal test canisters are provided as shown in Figure 12.3-23. There are

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-05-033 12.3-16 12 test canisters per 4-in. deep filter bed. Dioctylphthalate (DOP) and freon injection and detection ports are provided as shown.

12.3.4 IN-PLANT AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION 12.3.4.1 Criteria for Necessity and Location The objectives of the in-plant area radiation and airborne radioactivity monitoring systems are to

a.

Warn of excessive gamma radiation levels in areas where nuclear fuel is stored or handled,

b.

Provide operating personnel with a record and indication in the main control room of gamma radiation levels at selected locations within the various plant buildings,

c.

Provide information to the main control room so that decision may be made with respect to deployment of personnel in the event of a radiation accident or equipment failure,

d.

Assist in the detection of unauthorized or inadvertent movement of radioactive material within the various plant buildings,

e.

Provide local alarms at selected locations where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area,

f.

Monitor areas having a high potential for increase gamma radiation levels where personnel may be required to work,

g.

Supplement other systems including process radiation leak detection or building release detection in detecting abnormal migrations of radioactive materials from process streams,

h.

Monitor the general conditions in the reactor building following an accident, and

i.

Furnish information for making radiation surveys.

No credit is taken for the operability of the in-plant area radiation and airborne radioactivity monitors in the event of an accident. However, the probability is high that many or all of the 13 area monitors in the reactor building will be operable. These monitors have local sensors in separate physical locations within the reactor building. The wiring from the local sensors is

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-031 12.3-17 run to the main control room in cable runs that have Seismic Category I qualified supports.

The electrical supply for the area monitor indicator and control room alarms in the main control room are powered from a critical 120-V ac supply. The local audible alarms are powered from a local ac supply and its loss would not impact control room readings. The sensors are designed to maintain operability and accuracy in atmosphere up to 95% RH, and temperature to 50°C.

12.3.4.2 Description and Location

a.

Area radiation monitors Area radiation monitors consist of local detector alarm units and main control room mounted indicator trip units, alarms, and recorders. Redundant criticality monitors are located in the reactor building new fuel storage pit as recommended by Regulatory Guide 8.12. When practicable and feasible the guidance in Regulatory Guide 8.12 has been followed. Major items in Regulatory Guide 8.12 have been addressed and include

1.

Employing two detectors in the new fuel vault,

2.

Emergency planning that includes both the Emergency Plan and the Emergency Plan Implementing Procedures, evacuation routes, assembly areas, and yearly drills, and

3.

Surveillance testing of criticality alarm systems, including procedures that address methods and frequency of testing.

10 CFR 50.68(b) requires radiation monitors in storage and associated handling areas when fuel is present to detect excessive radiation levels and to alert personnel to initiate appropriate safety actions.

Other detector locations have been selected in accordance with good operating practice and from past operating experience with similar plants. Detector locations are shown in Figures 12.3-5 through 12.3-18. Annunciations are given in the main control room and locally at the sensors when radiation levels exceed a predetermined level. Point indication and recording are provided for in the main control room. Local detectors are wall-mounted approximately 7 ft off the floor. The detectors have sufficient cable length to be taken from their normal positions to floor level for insertion into calibrating chambers to verify instrument accuracy. Direct-reading dosimeters, worn by individuals in

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-18 radiologically controlled areas, provide protection in areas where area radiation monitors have not been installed.

An additional area radiation monitor is installed on the refueling bridge during bridge operation to provide personnel protection. This monitoring system provides local indication, visual, and audible alarm.

There is no area radiation monitor in the area of the radwaste building where the waste containers are filled and stored. Waste containers will normally be processed either in cask or in the shielded waste storage bay.

The location and ranges of the 31 area radiation monitors are given in Table 12.3-1. Table 12.3-2 lists the maximum background radiation levels for the area radiation monitors in the reactor building based on design basis calculation.

b.

Airborne radiation monitors Airborne radioactivity monitoring for plant personnel protection and surveillance uses fixed location, continuous particulate monitors which include continuous iodine samplers; portable continuous particulate monitors with continuous iodine samplers positioned at specific work sites; and particulate and iodine grab samples taken before and during specific jobs.

Movable local alarming continuous air monitors are placed at predetermined plant locations for personnel protection and to substantiate the quality of the plant breathing atmosphere. The monitors have local readouts (charts) and have radioiodine sampling capabilities.

The installed continuous particulate monitoring system was designed for responsive personnel protection and plant surveillance. The three installed particulate monitors measure the airborne particulate activity levels in the radwaste and reactor building ventilation exhaust and furnish recording signals to the main control room. These units draw approximately 3 cfm air sample through the particulate filter which is monitored by a shielded beta detector with an efficiency of approximately 30%. The resultant response of the system is an increase of about 350 cpm for 1 hr of sampling at a 1 x 10-10 Ci/cm3 concentration. External gamma radiation will increase the background by 70 cpm/mrem/hr.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-050 12.3-19 The actual ability of a ventilation exhaust monitoring system to detect the airborne particulate concentration in a specific space is dependent on the following factors:

1.

Flow rate ratio (flow of air from a specific confined space/flow rate of bulk ventilation system exhaust),

2.

Particulate activity and its half-life of the bulk ventilation system exhaust

air,

3.

Radionuclide composition in the specific confined space, and

4.

The energy of the beta radiation from the radionuclide composition.

Normal plant conditions are expected to yield a bulk ventilation exhaust air concentration (primarily short-lived fission product daughters and natural activity half-life about 20 minutes) of 1-3 x 10-10 Ci/cm3. This will reach an equilibrium on the sample filter of about 500 cpm.

The MPCa for normal plant airborne contamination is expected to be greater than 6 x 10-8 Ci/cm3. At this MPCa concentration a 1-hr accumulation (one MPCa-hr) will equal 2.0 x 105 cpm. Applying a dilution factor of 270:1, the 1-hr accumulation will equal 750 cpm.

This is a worst case dilution that considers the reactor building TIP Drive room at 400 cfm in the 105,000 cfm building bulk exhaust flow. Therefore, the ventilation monitoring system will easily detect 10 MPCa-hr on all locations.

Local particulate constant air monitoring instruments and a comprehensive (particulate, noble gases, and iodine) grab air sampling program complement the plant air sampling program.

Under these conditions, corrective actions will be taken and an assessment by portable sampling system results and portable monitoring activities will establish activity levels in all occupied areas which have potential for abnormal airborne activity.

In the radwaste building, the potentially contaminated areas normally entered by people would be those corridors adjacent to radioactive liquid and gaseous waste processing systems equipment such as demineralizers, concentrators, waste storage tanks, recombiners, dryers, moisture separators, and charcoal holdup vessels. Assuming that exfiltration from any one of the process systems to a normally entered corridor was sufficient to attain MPCa levels for 137Cs in that corridor, the dilution ratio would approach a factor of 10 to 100. For the worst case (100 to 1 ratio of bulk ventilation flow rate to corridor flow rate), 137Cs at MPCa (6 x 10-8 Ci/cm3) would be detected within 1 hr on the continuous particulate monitor. If exfiltration from a process vessel cubicle was sufficient to produce MPCa levels in an adjoining corridor, it is more probable that the normal cubicle flow rate input to the bulk ventilation flow would produce a prior distinguishable countrate ramp.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-20 In the turbine building, airborne contamination is most likely to arise from nuclear steam leaks or offgas processing systems piping. Heater bay areas, steam jet air ejector rooms, turbine areas, and steam-driven feedwater pump areas have the potential for airborne contamination.

Regular radiological surveys in the turbine building are not augmented by continuous particulate monitoring of the turbine building ventilation exhaust.

Each of the continuous particulate monitors has an associated iodine sampling cartridge which is counted regularly for baseline and surveillance information. This cartridge and iodine sampled/collected with portable sampling devices will be analyzed in the plant laboratory counting facility when abnormal airborne activity levels are signaled by a continuous particulate monitoring system. At a 5 cfm air flow rate through an iodine sampling cartridge, iodine present at an occupational MPCa concentration of 9 x 10-9 Ci/cm3 would be quantitatively observable within a 1-minute sample interval. Bases for this assessment are a 10-minute count time on a 12-15% Ge(Li) detector system having an overall efficiency of about 1% when source and geometry considerations are included. The information presented for detecting one MPCa concentration for 137Cs in areas having a low ventilation flow rate can also be applied to the iodine case. One MPCa of iodine can be ascertained within a 1-hr sampling period with a dilution factor of greater than 100. The routine weekly analysis of integrated iodine samples of radwaste, reactor, and turbine building ventilation air will permit observation of small iodine inputs. When these inputs are significant, a particulate and iodine sampling program is initiated to establish the source point.

Continuous particulate monitoring is reasoned to be the most responsive personnel protection and internal plant surveillance mechanism available. In addition, all tasks with potential for generating airborne contamination will be performed only when authorized by a radiation work permit (RWP).

The RWP assesses the radiological hazards, establishes additional monitoring and sampling requirements and, if necessary, specifies required engineering control and/or respiratory protection.

During outages, the above airborne monitoring system will be augmented by additional iodine sampling (continuous and grab) on the refueling floor since airborne iodine concentrations are known to become significant at this time.

12.3.4.3 Specification for Area Radiation Monitors The area radiation monitoring system is shown as a function block diagram in Figure 12.3-24.

Each channel consists of a sensor and a converter unit, a combined indicator and trip unit, a shared power supply, and a shared multipoint recorder. All channels also have a local meter and visual alarm auxiliary unit mounted near the sensor.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-21 Each monitor has an upscale trip that indicates high radiation and a downscale trip that indicates instrument failure. These trips actuate alarms but cause no control function. The trip circuits are set so that a loss of power annunciates in the control room.

The type of detector used is a Geiger-Mueller tube responsive to gamma radiation over an energy range of 80 KeV to 7 MeV. Detector ranges are given in Table 12.3-1.

The calibrating frequency is once every 18 months using standard sources with National Institute of Standards and Technology (NIST) traceability. This ensures accuracies of (+) or

(-) 20% over the detection interval.

An internal trip test circuit, which is adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip-unit input so that a meter reading is provided in addition to a real trip. High-range radiation alarm trip circuits for high level and criticality monitors are of the latching type and must be manually reset at the front of the control room panel. The trip circuits for all other area radiation monitors are of the nonlatching type.

12.3.4.4 Specification for Airborne Radiation Monitors The airborne particulate monitors contain scintillation detectors with count ratemeters. Means for remote recording are provided for in the main control room. Sample collectors consist of shielded, fixed particulate, filter-type air collectors. The calibration frequency will occur at least annually and after major maintenance. Instrument response checks will be made at least monthly. Monitors will be calibrated using standard radioactive sources in the same geometry as the location of the particulate filter or by collection and analysis of mixtures present at known air flow rates. Particulate monitors are provided in the reactor and radwaste buildings.

The monitors are located so as to monitor the exhaust air from that building prior to any filtration. In addition, charcoal sampling cartridges are installed in each monitor for laboratory analysis of iodine.

Each of the three channels of the airborne radioactivity monitors has an independent local visual and audible alarm. High radioactivity or equipment failure will generate an alarm signal. No automatic system functions are performed by the alarm signals.

12.3.4.5 Annunciators and Alarms The annunciators are of the window type with pushbutton switches for sound acknowledgment and light reset functions. There are no automatic system actions performed by the area radiation monitors. The annunciator alarm windows are located in the main control room.

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-06-000 12.3-22 Area monitors have local/remote alarms that sound on exceeding their high radiation alarm settings and remote alarms to indicate instrument failure (see Figure 12.3-24). Monitors located in the reactor building near the fuel pool and in the new fuel areas have individual high radiation alarm windows. The remainder of the area monitor units in the reactor building have a common annunciator window for high radioactivity. Area monitors in the turbine building and the radwaste building each have a common building high radioactivity alarm window. All the area monitors have one common alarm window for instrument failure.

The two area monitors that are used as criticality detectors are located in the new fuel vault.

These monitors have a range of 10+2-10+6 mrem/hr and alarm on either of a fully adjustable upscale or downscale trip point. The alarm setpoint and bases are given in the Licensee Controlled Specifications.

12.3.4.6 Power Sources, Indicating and Recording Devices The area radiation monitor power supply units, indicating devices (except local alarms), and the recorders are all located on panels in the control room and receive power from the 120-V ac instrument power supply bus. The local audio alarms are supplied from a local 120-V ac instrument distribution panel. The recorder is a multipoint, strip chart type, which compiles a permanent record of inputs from the area radiation monitors.

12.

3.5 REFERENCES

12.3-1 Jaeger, R. G. et al., Engineering Compendium on Radiation Shielding, Volume 1, Shielding Fundamentals and Methods.

12.3-2 Rockwell, T., Reactor Shielding Design Manuals, 1st Edition, D. Van Nostrand Co., Inc., New York, 1956.

12.3-3 Goldstein, H., Fundamental Aspects of Reactor Shielding, Addison-Wesley Publishing Co., Inc., Reading, 1959.

12.3-4 Blizard, E. P., Reactor Handbook, Vol. III, Part B, Shielding.

12.3-5 Aalto, E. et al., A Users Manual for the NRN Shield Design, AE-145, June 1964.

12.3-6 Hughes, D. J. and Schwartz, R. B., Neutron Cross Sections, BNL 325, 2nd Edition, 1958.

Hughes, D. J., Magurno, B. A. and Brussel, M. K., Neutron Cross Sections, BNL 325, wnd. ed., Supplement 1, 1960.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-23 Stehn, John R. et al., Neutron Cross Sections, BNL 325, 2nd. Edition, Supplement 2, 1964.

12.3-7 Malenfant, Richard, QAD-A Series of Point Kernel General Purpose Shielding Program, U.S. Department of Commerce, LA-3573, Springfield, VA, 1966.

12.3-8 Walker, R. L., and Grotenhuis, M., A Summary of Shielding Constants for Concrete, ANL-6443, November 1961.

12.3-9 Chilton, A. B. and Huddleston, C. M., A Semiemperical Formula for Differential Dose Albedo for Gamma Rays on Concrete, Nuclear Science and Engineering, 17, 419-424, 1963.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 Table 12.3-1 Area Monitors Station Location Building Level (ft)

Range (mrem/hr)

LDCN-98-117 12.3-25 1

Reactor building fuel pool area 606 102-106 2

Reactor building fuel pool area 606 1-104 3

Reactor building new fuel area 606 102-106 3A Reactor building new fuel area 2 606 102-106 4

Reactor building control rod hyd equipment area E 522 1-104 5

Reactor building control rod hyd equipment area W 522 1-104 6

Reactor building equipment access area S 572 1-104 7

Reactor building neutron monitor system drive mechanical area 501 1-104 8

Reactor building SGTS filters area 572 1-104 9

Reactor building northwest RHR pump room 422 1-104 10 Reactor building southwest RHR pump room 422 1-104 11 Reactor building northeast RHR pump room 422 1-104 12 Reactor building RCIC pump room 422 1-104 13 Reactor building HPCS pump room 422 1-104 14 Turbine building turbine front standard 501 1-104 15 Turbine building entrance 441 1-104 16 Turbine building reactor feed pump area 1A 441 1-104 17 Turbine building reactor feed pump area 1B 441 1-104 18 Turbine building condensate pump area 441 1-104

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 Table 12.3-1 Area Monitors (Continued)

Station Location Building Level (ft)

Range (mrem/hr)

LDCN-98-117 12.3-26 19 Main control room 501 1-104 20 Radwaste building valve room E 467 1-104 21 Radwaste building valve room W 467 1-104 22 Radwaste building sample room 487 1-104 23 Reactor building CRD pump room 10 422 1-104 24 Reactor building equipment access area (W) 471 1-104 25 Radwaste building hot machine shop 487 1-104 26 Radwaste building contaminated tool room 467 1-104 27 Radwaste building waste surge tank area 437 1-104 28 Radwaste building tank corridor area north 437 1-104 29 Radwaste building tank corridor area south 437 1-104 30 Radwaste building radwaste control room 467 1-104 32 Reactor building NE entrance 471 10-1-104 33 Reactor building NW entrance 501 10-1-104 34 Reactor building eastside 606 10-1-104 35a Reactor building refueling bridge 606 0.1-2000 a Item 35 is installed at its dedicated location on the refueling bridge prior to bridge operation.

Alarm settings for all of the above monitors will be selected to provide indication of any abnormal increase in radiation levels while minimizing false alarms.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-98-117 12.3-27 Table 12.3-2 Maximum Design Basis Background Radiation Level for Area Monitors ARM Building Level (ft)

Maximum Design Basis Background Level (mrem/hr)

ARM-RE-1 606 100 ARM-RE-2 606 100 ARM-RE-3 606 100 ARM-RE-3A 606 100 ARM-RE-4 522 3000 ARM-RE-5 522 500 ARM-RE-6 572 100 ARM-RE-7 501 100 ARM-RE-8 572 100 ARM-RE-9 422 4000 ARM-RE-10 422 4000 ARM-RE-11 422 100 ARM-RE-12 422 3000 ARM-RE-13 422 100 ARM-RE-23 422 3000 ARM-RE-24 471 100 ARM-RE-32 471 100 ARM-RE-33 501 100 ARM-RE-34 606 100

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS Section Page LDCN-13-039 12-i 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA).............. 12.1-1 12.1.1 POLICY CONSIDERATIONS...................................................... 12.1-1 12.1.2 DESIGN CONSIDERATIONS...................................................... 12.1-4 12.1.3 OPERATIONAL CONSIDERATIONS............................................ 12.1-8 12.1.3.1 Procedures and Methods of Operation........................................... 12.1-8 12.1.3.2 Design Changes for ALARA Exposures......................................... 12.1-9 12.1.3.3 Operational Information............................................................ 12.1-10 12.2 RADIATION SOURCES................................................................ 12.2-1 12.2.1 CONTAINED SOURCES............................................................ 12.2-1 12.2.1.1 General................................................................................ 12.2-1 12.2.1.2 Reactor and Turbine Building..................................................... 12.2-1 12.2.1.2.1 Reactor Core Radiation Sources................................................ 12.2-1 12.2.1.2.2 Process System Radiation Sources............................................. 12.2-2 12.2.1.2.2.1 Introduction...................................................................... 12.2-2 12.2.1.2.2.2 Recirculation System Sources................................................ 12.2-2 12.2.1.2.2.3 Reactor Water Cleanup System Sources.................................... 12.2-3 12.2.1.2.2.4 Reactor Core Isolation Cooling System Source........................... 12.2-3 12.2.1.2.2.5 Residual Heat Removal System Sources.................................... 12.2-3 12.2.1.2.2.6 Fuel Pool Cooling and Cleanup and System Sources..................... 12.2-4 12.2.1.2.2.7 Main Steam and Reactor Feedwater Systems Sources.................... 12.2-5 12.2.1.2.2.8 Offgas Sources in the Turbine Generator Building....................... 12.2-5 12.2.1.2.2.9 Traveling In-Core Probe System Sources.................................. 12.2-6 12.2.1.2.2.10 Sources Resulting From Crud Buildup.................................... 12.2-6 12.2.1.3 Radwaste Building................................................................... 12.2-6 12.2.1.4 Byproduct, Source, and Special Nuclear Materials............................ 12.2-6 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES........................ 12.2-6 12.2.2.1 General................................................................................ 12.2-6 12.2.2.2 Model for Computing the Airborne Radionuclide Concentration in a Plant Area.......................................................................... 12.2-7 12.2.2.3 Sources of Airborne Radioactivity................................................ 12.2-8 12.2.2.3.1 Effect of Leakage from Process Equipment in Radioactive Systems..... 12.2-8 12.2.2.3.2 Effect of Sumps, Drains, Tank and Filter Demineralizer Vents.......... 12.2-10 12.2.2.3.3 Effect of Relief Valve Exhaust.................................................. 12.2-11

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS (Continued)

Section Page LDCN-05-002 12-ii 12.2.2.3.4 Effect of Removing Reactor Pressure Vessel Head and Associated Internals............................................................................. 12.2-13 12.2.2.3.5 Effect of Sampling................................................................ 12.2-13 12.2.2.3.6 Effect of Spent Fuel Movement................................................. 12.2-13 12.2.2.3.7 Effects of Solid Radwaste Handling Areas................................... 12.2-14 12.2.2.3.8 Effects of Liquid Radwaste Handling Areas.................................. 12.2-14 12.

2.3 REFERENCES

......................................................................... 12.2-14 12.3 RADIATION PROTECTION DESIGN FEATURES.............................. 12.3-1 12.3.1 FACILITY DESIGN FEATURES.................................................. 12.3-1 12.3.1.1 Radiation Zone Designations...................................................... 12.3-1 12.3.1.2 Traffic Patterns....................................................................... 12.3-2 12.3.1.3 Radiation Protection Design Features............................................ 12.3-2 12.3.1.3.1 Facility Design Features......................................................... 12.3-2 12.3.1.3.2 Design Features That Reduce Crud Buildup.................................. 12.3-6 12.3.1.3.3 Field Routing of Piping.......................................................... 12.3-7 12.3.1.3.4 Design Features That Reduce Occupational Doses During Decommissioning................................................................. 12.3-7 12.3.1.4 Radioactive Material Safety........................................................ 12.3-8 12.3.1.4.1 Materials Safety Program........................................................ 12.3-8 12.3.1.4.2 Facilities and Equipment......................................................... 12.3-9 12.3.1.4.3 Personnel and Procedures........................................................ 12.3-9 12.3.1.4.4 Required Materials................................................................ 12.3-10 12.3.2 SHIELDING............................................................................ 12.3-10 12.3.2.1 General................................................................................ 12.3-10 12.3.2.2 Methods of Shielding Calculations................................................ 12.3-11 12.3.2.3 Shielding Description............................................................... 12.3-12 12.3.2.3.1 General.............................................................................. 12.3-12 12.3.2.3.2 Reactor Building................................................................... 12.3-12 12.3.2.3.3 Turbine Building.................................................................. 12.3-13 12.3.2.3.4 Radwaste Building................................................................ 12.3-13 12.3.3 VENTILATION........................................................................ 12.3-13 12.3.4 IN-PLANT AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION........................................... 12.3-16 12.3.4.1 Criteria for Necessity and Location.............................................. 12.3-16 12.3.4.2 Description and Location........................................................... 12.3-17

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 12 RADIATION PROTECTION TABLE OF CONTENTS (Continued)

Section Page LDCN-05-056 12-iii 12.3.4.3 Specification for Area Radiation Monitors...................................... 12.3-20 12.3.4.4 Specification for Airborne Radiation Monitors................................. 12.3-21 12.3.4.5 Annuciators and Alarms............................................................ 12.3-21 12.3.4.6 Power Sources, Indicating and Recording Devices............................ 12.3-22 12.

3.5 REFERENCES

......................................................................... 12.3-22 12.4 DOSE ASSESSMENT................................................................... 12.4-1 12.4.1 DESIGN CRITERIA.................................................................. 12.4-1 12.4.2 PERSONNEL DOSE ASSESSMENT BASED ON BWR OPERATING DATA.................................................................. 12.4-1 12.4.2.1 General................................................................................ 12.4-1 12.4.2.2 Personnel Dose from Operating BWR Data..................................... 12.4-2 12.4.2.3 Occupancy Factors, Dose Rates, and Estimated Personnel Exposures..... 12.4-2 12.4.3 INHALATION EXPOSURES....................................................... 12.4-4 12.4.4 SITE BOUNDARY DOSE........................................................... 12.4-4 12.

4.5 REFERENCES

......................................................................... 12.4-5 12.5 RADIATION PROTECTION PROGRAM.......................................... 12.5-1 12.5.1 ORGANIZATION..................................................................... 12.5-1 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES.................. 12.5-2 12.5.2.1 Criteria for Selection................................................................ 12.5-4 12.5.2.2 Facilities............................................................................... 12.5-6 12.5.2.3 Equipment............................................................................. 12.5-8 12.5.2.4 Instrumentation....................................................................... 12.5-9 12.5.3 PROCEDURES......................................................................... 12.5-9 12.5.3.1 Personnel Control Procedures..................................................... 12.5-9 12.5.3.2 As Low As Is Reasonably Achievable Procedures............................. 12.5-10 12.5.3.3 Radiological Survey Procedures................................................... 12.5-12 12.5.3.4 Procedures for Radioactive Contamination Control........................... 12.5-13 12.5.3.5 Procedures for Control of Airborne Radioactivity............................. 12.5-14 12.5.3.6 Radioactive Material Control Including Special Nuclear Materials (SNM)..................................................................... 12.5-15 12.5.3.7 Personnel Dosimetry Procedures.................................................. 12.5-16 12.5.3.8 Radiation Protection Surveillance Program..................................... 12.5-18

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 Chapter 12 RADIATION PROTECTION LIST OF TABLES (Continued)

Number Title Page LDCN-14-005 12-iv 12.2-1 Basic Reactor Data for Source Computations.................................. 12.2-17 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary............................ 12.2-18 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation............ 12.2-19 12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown...................................................................... 12.2-20 12.2-5 Fission Product Source in RHR Piping and Heat Exchangers 4 Hours After Shutdown...................................................................... 12.2-21 12.2-6 Gamma Ray Energy Spectrum for Spent Fuel Sources....................... 12.2-22 12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater...... 12.2-23 12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell............................................................. 12.2-24 12.2-9 Nitrogen -16 Source Strength in Feedwater Heater 6......................... 12.2-25 12.2-10 Nitrogen -16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems................................... 12.2-26 12.2-11 Offgas System Sources in the Turbine Generator Building.................. 12.2-27 12.2-12a Special Sources With Strength Greater Than 100 Millicuries............... 12.2-28 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area................................................... 12.2-28a 12.2-13 List of Radioactive Piping and System Designations.......................... 12.2-29 12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building)................................................ 12.2-30

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 Chapter 12 RADIATION PROTECTION LIST OF TABLES (Continued)

Number Title Page LDCN-14-005 12-v 12.2-15 Airborne Radionuclide Concentration in Condensate Pump Area (el. 441 ft. 0 in. turbine generator building).................................... 12.2-31 12.2-16 Airborne Radionuclide Concentration in Secondary Containment from a Main Steam Relief Valve Blowdown................................... 12.2-32 12.2-17 Airborne Radionuclide Concentration in Liquid Radwaste Handling Area........................................................................ 12.2-33 12.3-1 Area Monitors........................................................................ 12.3-25 12.3-2 Maximum Design Basis Background Radiation Level for Area Monitors........................................................................ 12.3-27 12.4-1 Summary of Occupational Dose Estimates...................................... 12.4-7 12.4-2 Occupational Dose Estimates During Routine Operations and Surveillance........................................................................... 12.4-8 12.4-3 Occupational Dose Estimates During Nonroutine Operations and Surveillance........................................................................... 12.4-11 12.4-4 Occupational Dose Estimates During Routine Operations and Surveillance........................................................................... 12.4-12 12.4-5 Occupational Dose Estimates During Waste Processing...................... 12.4-13 12.4-6 Occupational Dose Estimates During Refueling............................... 12.4-14 12.4-7 Occupational Dose Estimates During Inservice Inspection................... 12.4-15 12.4-8 Occupational Dose Estimates During Special Maintenance.................. 12.4-16 12.4-9 Summary of Annual Information Reported by Commercial Boiling Water Reactors....................................................................... 12.4-17 12.5-1 Health Physics Instrumentation................................................... 12.5-21

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 Chapter 12 RADIATION PROTECTION LIST OF FIGURES Number Title LDCN-12-037 12-vi 12.3-1 DELETED 12.3-2 DELETED 12.3-3 DELETED 12.3-4 DELETED 12.3-5 Radiation Zones - Turbine Generator Building 12.3-6 Radiation Zones - Ground Floor Plan - Turbine Generator Building 12.3-7 Radiation Zones - Mezzanine Floor Plan - Turbine Generator Building, East Side 12.3-8 Radiation Zones - Mezzanine Floor Plan - Turbine Generator Building, West Side 12.3-9 Radiation Zones - Operating Floor Plan - Turbine Generator Building, East Side 12.3-10 Radiation Zones - Operating Floor Plan - Turbine Generator Building, West Side 12.3-11 Radiation Zones - El. 437 ft 0 in. Radwaste Building 12.3-12 Radiation Zones - El. 467 ft 0 in. and Partial Plans Radwaste Building 12.3-13 Radiation Zones - El. 484 ft 0 in. and 487 ft 0 in., and Partial Plans Radwaste Building 12.3-14 Radiation Zones - El. 501 ft 0 in., 507 ft 0 in., and 525 ft 0 in. Radwaste Building 12.3-15 Radiation Zones - El. 422 ft 3 in., 441 ft 0 in., and 440 ft 0 in. Reactor Building 12.3-16 Radiation Zones - El. 471 ft 0 in. and 501 ft 0 in. Reactor Building 12.3-17 Radiation Zones - El. 522 ft 0 in. and 548 ft 0 in. Reactor Building 12.3-18 Radiation Zones - El. 572 ft 0 in. and 606 ft 10-1/2 in. Reactor Building

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 Chapter 12 RADIATION PROTECTION LIST OF FIGURES (Continued)

Number Title LDCN-05-056 12-vii 12.3-19 Arrangement of Filtration and Demineralization Equipment (Typical) 12.3-20 Schematic Arrangement of the Cooler Condenser Loop Seal 12.3-21 Decontamination Concentrator Steam Supply Arrangement 12.3-22 Entombment Structure 12.3-23 Layout of the Standby Gas Treatment System Filter Units 12.3-24 Block Diagram - Area Radiation Monitoring System

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-042 12.1-1 Chapter 12 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS Energy Northwest is committed to maintaining occupational and public radiation exposures as far below regulatory dose limits as is practical while performing all activities related to the operation of Columbia Generating Station (CGS) and the Independent Spent Fuel Storage Installation (ISFSI). This commitment is reflected in the Radiation Protection Program (RPP) that meets the requirements of 10 CFR 20 and provides for effective control of radiation exposure through

a.

Management direction and support,

b.

Establishment of radiation control procedures,

c.

Consideration during design and modification of facilities and equipment, and

d.

Development of good radiation control practices, including preplanning and the proper use of appropriate equipment by qualified, well trained personnel.

The radiation protection practices are based, when practicable and feasible, on Regulatory Guides 8.8, Revision 3, and 8.10, Revision 1. The Radiation Protection Program provides for the majority of the recommended actions in both regulatory guides, including the following:

a.

Organization and position descriptions to ensure an adequate as low as is reasonably achievable (ALARA) program,

b.

Exposure reduction program,

c.

Cost-benefit analysis program, and

d.

Exposure tracking program employing the Radiation Work Permit.

Procedures for personnel radiation protection are prepared consistent with the requirements of 10 CFR Part 20 and are approved, maintained, and adhered to for all operations involving personnel radiation exposure.

Energy Northwest organization is structured to provide assurance that the ALARA policy is effective in the areas described above. The following is a description of the applicable activities conducted by individuals or groups having responsibility for radiation protection.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-13-061 12.1-2

a.

The Plant General Manager, has the overall responsibility for approving the RPP and ALARA policy consistent with Energy Northwest and regulatory requirements, and for the radiological safety of all on-site personnel. This includes the responsibility for implementation of the ALARA program by plant staff. The Plant General Manager has ensured consideration of exposure reduction methods by adoption of the ALARA review practices described in Section 12.1.3. The Plant General Manager is responsible for the management of plant operational activities and for providing the Radiological Services Manager the resources and support necessary to implement the RPP. This includes the responsibility for ensuring that the ALARA program is not adversely affected by production oriented goals;

b.

The Radiological Services Manager reports to the Chemistry and Radiological Safety Manager and is responsible for implementing the RPP. This position meets the requirements of the Radiation Protection Manager (RPM) as described in Reg Guide 1.8. This individual provides organizational leadership and direction to the Radiation Protection department;

c.

The Radiological Services Manager has direct access to the Plant General Manager in all matters relating to radiation safety, and has the responsibility and authority for ensuring that plant activities meet applicable radiation safety regulations and RPP requirements. Specific responsibilities are provided in Section 12.5.1;

d.

The Radiological Operations Supervisor reports to the Radiological Services Manager. This individual provides supervision, leadership, and technical direction for implementation of the RPP;

e.

The Health Physics (HP) Craft Supervisors report to the Radiological Operations Supervisor and implement the RPP through direct supervision of the plant HP Technicians. Areas of responsibility include characterization of plant radiological conditions, maintenance of radiological postings, detection and evaluation of radiological problems, and performance of facility and equipment decontamination, system flushes, and temporary shielding installation;

f.

The Radiological Support Supervisor reports to the Radiological Services Manager and is responsible for radiation exposure reduction and for providing technical support to the Radiation Protection organization. This is accomplished, in part, by providing input to the planning and coordination of plant activities. The Radiological Support Supervisor makes recommendations

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-15-028 12.1-3 for the control/elimination of radiological conditions that increase personnel exposure or the release of radioactivity; and provides additional assurance that regulatory requirements and industry guidance are incorporated into the program to ensure exposures are maintained ALARA.

In addition to the organization structure that has been provided for implementation of the ALARA policy, the following groups perform reviews to further ensure exposures are ALARA.

a.

The Plant Operations Committee (POC) has been established and is functional.

Its purpose is to serve as a review and advisory organization to the Plant General Manager in several areas including nuclear safety. A written administrative procedure describes the responsibilities of the POC. The RPM is a member of the POC and has direct input to this group on all radiological matters;

b.

The Quality organization performs audits, surveillances, and assessments of plant operations. Audits, surveillances, and assessments will include verification of compliance with plant procedures, with regulations involving nuclear safety, and with operating license provisions. Results of these audits and surveillances will be submitted to both plant and corporate management.

Since the system for ALARA review described in Section 12.1.3 provides for this consideration in all plant procedures, quality audits and surveillances will verify implementation of this principle;

c.

The Corporate Nuclear Safety Review Board (CNSRB) provides a continuing appraisal of the ALARA policy. The Operational Quality Assurance Program Description (OQAPD) provides a description of this groups responsibilities and authority. By its charter, the CNSRB reviews radiological safety policies and programs and determines that these policies and programs are in compliance with NRC requirements. The CNSRB has the capacity for review in the area of radiological safety and has established a direct line of communication with plant management; and

d.

The Senior Site ALARA Committee serves as a review and advisory organization to the Plant General Manager on radiological safety, including occupational exposure to personnel. Committee membership, responsibilities, authorities, and records are prescribed in plant procedures.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-4 The commitment to ALARA is implemented by employee training; audits, assessments, and reviews of the program; procedure development and reviews; enforcement of rules; and modifications to plant equipment or facilities where they will substantially reduce exposures at a reasonable cost. Managements commitment to the ALARA policy is further discussed in Section 12.5. This program meets all 10 CFR Part 20 regulations and considers the guidance of Regulatory Guides 8.8 and 8.10 in regard to policy considerations.

12.1.2 DESIGN CONSIDERATIONS To ensure that personnel occupational radiation exposures are ALARA, extensive consideration is given to equipment design and locations, accessibility requirements, and shielding requirements. Many of these design objectives and considerations were established prior to the issuance of Regulatory Guide 8.8. However, the design of the plant substantially incorporates the recommendations provided in the regulatory guide. Design considerations that ensure occupational radiation exposures to personnel during normal operation and anticipated operational occurrences are ALARA are the following:

a.

The facility is separated into controlled and uncontrolled areas based on anticipated radiation levels. The controlled areas of the facility are further defined by radiation zones established by personnel access requirements and are intended to limit radiation exposure to ALARA. The radiation zones provide guidance for the shielding design, contamination control, and ventilation flow pattern for the areas of anticipated personnel radiation exposure. See Section 12.3.1 for a description of the facility radiation zones.

b.

Equipment location

1.

Several radiation sources on the 437 ft 0 in. level of the radwaste building are located with two sources in each cubicle. This arrangement maintains occupational exposures ALARA by use of the following alternate ALARA methods.

The waste collector tank and floor drain collector tank are in the same cubicle. These tanks share redundant pumps and cross tie piping. If abnormal conditions occur and one or both of these tanks becomes a major source of radiation, either one of the pumps can be used to empty the tanks prior to any maintenance.

The chemical waste tank and distillate tank share the same cubicle.

These tanks are not expected to be major sources of radiation. Based on the source terms described in Table 11.2-1, the dose rate at 3 ft from the surface of these tanks normally does not exceed 0.1 mrem/hr. In

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-5 addition, redundant pumps and cross tie piping permit the transfer of tank contents should abnormally high radioactivity levels occur.

Gas coolers and charcoal adsorbers share the same cubicle. These items have no moving parts and are highly reliable with no routine maintenance requirements. In addition, system redundancy and remote isolation capabilities eliminate the need for prompt entry into the cubicle.

This permits the noble gases and radioiodines to significantly decay prior to entry.

Placing the preceding sources in shared cubicles does not result in increased occupational exposures.

2.

Radioactive pipes are routed so that radiation exposure to plant personnel is minimized. The extent to which radioactive pipes are routed through normally accessible areas is minimized. Shielded pipe chases are utilized in normally accessible areas. Whenever possible, radioactive and nonradioactive pipes are kept separate for maintenance purposes.

3.

Shielded valve stations are used where practical. To further minimize personnel exposure, remotely operated valves are used where practical.

Normally operated manual valves in high radiation areas are provided with extension stems through a shield wall to a low radiation area.

4.

Where practical, pumps are located in shielded areas outside cubicles containing radioactive components.

5.

Where practical, local instrumentation readouts are routed to points outside shielding walls.

6.

To minimize maintenance time and hence exposure, sufficient space is available in shield cubicles housing radioactive equipment (e.g., heat exchangers, demineralizers, etc.) to perform required tasks. Access platforms at intermediate levels are provided to enhance access to portions of equipment inaccessible from the floor.

7.

Where possible, equipment and components which require frequent servicing are designed so that they can be removed, with minimum exposure, to appropriate low radiation areas.

8.

Access to corridor C-125 on the 437 ft 0 in. level of the radwaste building will only be required for routine surveillance of equipment and nonroutine maintenance. Area radiation monitors 28 and 29 are located

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-6 in the corridor to detect abnormal radiological conditions and warn personnel if radiation levels are excessive. During reactor and radwaste operations, entry to this area will be under the direction of a Radiation Work Permit (RWP).

c.

Shielding design is based on satisfying the radiation zone requirements utilizing design basis radiation sources for the shielding calculations. Shielding design is conservative since the design basis radiation sources are not expected to occur frequently.

Penetrations through radiation shields are generally designed to prevent direct radiation streaming from a high radiation area to a low radiation area. Shielding discontinuities, such as shield plugs, are provided with offsets to reduce radiation streaming. Shield doors and labyrinths are used to eliminate radiation streaming through access openings in the cubicles.

d.

Auxiliary systems that may become contaminated are designed with provisions for flushing or remote chemical cleaning prior to maintenance. This is accomplished by the following:

1.

Providing connections for the purpose of backflushing,

2.

Providing water connections to tanks containing spargers to allow for water injection to uncake contaminants, and

3.

Providing cross ties between redundant equipment and/or related equipment capable of redundant operation to allow removal of contaminated equipment from service.

e.

The ventilation systems are designed to ensure control of airborne contaminants by providing air flow from areas of low radioactivity potential to areas of high radioactivity potential. Access to the systems is facilitated by the following:

1.

Filter access doors, which are sized to enhance the ease of performing maintenance, and

2.

Providing for periodic inservice testing of the equipment and filters.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-7

f.

Spread of contamination is minimized in the event spillage occurs by the following:

1.

Drains are provided in areas where equipment with large volumes of radioactive fluid is located. Drains are sized to conduct spillage to the appropriate liquid waste processing system;

2.

Floors and walls are protected with the appropriate coating to facilitate decontamination; and

3.

An equipment decontamination facility is provided to decontaminate tools and radioactive components.

g.

While pipe runs are not sloped, those that carry radioactive fluids can be chemically decontaminated. Tank bottoms in the radwaste system are either sloped or dished. The exception is the waste surge tank located on el. 437 ft 0 in. level in the radwaste building.

h.

Drain tap-offs are provided at low points in the piping systems.

i.

Connections are placed above the centerline (top) of pipes when consistent with overall design requirements. Such connections are not always practical or desirable. For example, air pockets may form in an infrequently used line which is connected above the centerline (top) of another pipe.

j.

Where practicable, the use of dead legs and low points in pipe systems are avoided, especially in pipes carrying demineralizer resins or concentrated radwaste.

k.

T-connections in piping are minimized with the exception of

1.

Multiple flow paths, such as in the condensate filter demineralizer system, and

2.

Interconnected systems acting as mutual back ups, such as the floor drain and waste filter demineralizer systems.

l.

Large pipe bend radii and piping elbows are used.

m.

Butt welding by the open root method is used as described in Section 12.3.1.3.2.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-022 12.1-8

n.

Seal glands of pumps carrying concentrated radwaste or spent filter and demineralizer resins are flushed with condensate water during operation. After operation, the pumps are also flushed with condensate. Canned pumps are not used.

o.

Where practicable, equipment used in the same process are located close together, resulting in short runs of interconnecting piping.

p.

Pipes carrying slurries and resins are sized for turbulent flow to prevent any settling out of solids.

q.

All equipment cubicles housing filters are equipped with removable ceiling plugs through which filter elements may be serviced or changed with the aid of tools to allow remote handling.

r.

Operating experience from other BWR plants is periodically reviewed.

Problems are reviewed and the plant design is checked to ensure that similar problems will not occur.

s.

Design changes are reviewed by Radiation Protection.

12.1.3 OPERATIONAL CONSIDERATIONS 12.1.3.1 Procedures and Methods of Operation A positive means of ensuring that occupational and public radiation exposures are ALARA has been incorporated into the Plant Procedures Manual (PPM) and Procedure Program.

Procedures are formally reviewed for ALARA considerations as part of the approval process.

The guidance provided by Regulatory Guide 8.8 is considered during this review. Additional information regarding ALARA considerations in procedures is provided in Section 12.5.3.2.

In addition to the above process, the Radiation Work Permit (RWP) process is used for individual tasks to identify radiological conditions and controls. This provides a means of prescribing precautionary measures, protective equipment, and other exposure reduction methods in each situation. Individual exposures, as determined by dosimeters, are recorded and provide a means of determining exposures for various tasks and work groups. This information is used for preplanning work, identifying sources, determining radiation levels and otherwise evaluating exposure problems.

Administrative controls ensure that occupational and public radiation exposures are as low as reasonably achievable. These are supplemented by the Radiation Protection staff through a

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-042 12.1-9 program of surveillance, guidance, and investigation. A description of the program is outlined in Section 12.5 and includes the following aspects:

a.

The Energy Northwest RPP includes procedures that provide for routine and special survey to determine sources and trends of exposure and for investigation to determine causes of normal and unusual exposure;

b.

Plant procedures are formally reviewed by Radiation Protection for ALARA considerations when required;

c.

Plant modifications that have radiological implications are reviewed by Radiation Protection to ensure that the changes incorporate ALARA considerations;

d.

All maintenance tasks that involve radiological systems or controls are initially reviewed for ALARA and radiological concerns. For these tasks, a RWP is completed which describes the proposed actions. The RWP describes the conditions necessary such as survey requirements, surveillance, and protective apparel;

e.

Prior to each scheduled maintenance and refueling outage, HP reviews the outage tasks and schedules and participates in outage planning for ALARA considerations; and

f.

Adequate surveillance and supervision is provided to ensure that procedures are followed, prescribed precautions are taken, and radiation sources are identified.

12.1.3.2 Design Changes for ALARA Exposures Operational requirements were considered in the original design of CGS for maintaining occupational exposures ALARA, and several design changes have been made since issuance of the PSAR and Regulatory Guide 8.8. These changes or additions were implemented as a result of review by both the architect-engineer and Energy Northwest personnel and include the following:

a.

Revised offgas system valve design to prevent release of radioactive gases to building atmosphere,

b.

Relocation of the counting room for lower background levels and adequate shielding,

c.

Revised effluent monitoring capabilities to provide for more efficient monitoring,

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-10

d.

Increased capability for in-plant continuous airborne radioactivity monitoring with remote readout and recording features,

e.

Increased capability for the area radiation monitoring system to include remote readout/alarm and local alarm functions for the individual monitors,

f.

Inclusion of supplied air stations throughout the plant for efficient respiratory protection,

g.

Space and services provisions made for a decontamination facility and hot shop to reduce contact maintenance exposures and airborne radioactivity,

h.

Revised penetration access design at sacrificial shield wall to reduce time required in this area,

i.

Installation of overload protection on valve motor operators to minimize motor replacement in high dose rate areas,

j.

Generated additional specification for replacement valve packing for selected valves to reduce time consumed in repacking,

k.

Replaced hydraulic snubbers with mechanical snubbers to reduce maintenance requirements,

l.

Provided method of venting the reactor vessel head through main steam line A to be ultimately filtered through the offgas system via the condenser, and

m.

Made provisions for future connections to increase reactor water cleanup capacity during shutdown conditions to reduce radioiodine and particulate activity prior to vessel head removal and during outage periods.

New designs or design revisions are considered for exposure reduction as plant operation identifies problem areas.

12.1.3.3 Operational Information Information from operating BWRs has been incorporated into the Radiation Protection procedures as discussed below:

a.

Procedures have been written using guidance obtained from BWR operational experience reviews, from technical information obtained at power reactor conferences, and from participational experience in outages at operating BWRs;

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 12.1-11

b.

Respiratory protection procedures incorporate proven practices from other nuclear facilities;

c.

Typical procedures on survey methods, personnel monitoring, personnel dosimetry, and process/effluent radiological monitoring have been observed in the implementation stage at several operating reactor facilities. In addition, the procedures used at operating BWR plants have been reviewed. All of these were evaluated and applied in the procedure generating process;

d.

Specific HP procedures or instructions have been written to furnish guidance on the following:

1.

The issuance, requirements, conditions, and controls of RWPs,

2.

The review process of plant procedures for ALARA considerations, and

3.

Methods for minimizing personnel exposures during RPV head removal, drywell entry, and conduct during emergencies.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-1 12.2 RADIATION SOURCES 12.2.1 CONTAINED SOURCES 12.2.1.1 General The design basis radiation sources considered are the following:

a.

The reactor core,

b.

Activation of structures and components in the vicinity of the reactor core,

c.

Radioactive materials (fission and corrosion products) contained in system components,

d.

Spent fuel, and

e.

Radioactive wastes for offsite shipment.

The design basis process radiation sources used for shielding are based on the source terms given in Tables 11.1-1 through 11.1-5.

12.2.1.2 Reactor and Turbine Building The reactor building sources include the following:

a.

The reactor core,

b.

Activated structures and components,

c.

Components and equipment containing activation, fission, and corrosion products, and

d.

Spent fuel.

12.2.1.2.1 Reactor Core Radiation Sources During normal power operation, the reactor core radiation sources considered are the prompt fission neutrons and gamma rays, capture gamma rays, and fission product gamma rays.

During shutdown, the reactor core radiation sources are the fission product gamma rays. The core is treated as a cylindrical volume source for use in the NRN+(1) and QAD+(2) codes.

See Section 12.3.2 for details.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-98-117 12.2-2 Table 12.2-1 lists the basic data required for reactor core source computations including volume fractions, power peaking factors, and core power density.

Table 12.2-2 presents the neutron multigroup flux at the reactor core-reflector boundary, 238 cm from the core centerline. The NRN computer code (Reference 12.2-1) is used in computing the flux. This code is based on a multigroup slowing down and diffusion system with the deep penetration source corrected by a multigroup removal source.

Table 12.2-3 lists the gamma ray energy spectrum for the reactor core during normal operation. The operating core energy spectrum represents the sum of the spectra for prompt fission, fission product, and to capture gamma rays. The postoperation fission product gamma ray energy spectrum in the core immediately after shutdown is listed in Table 12.2-4.

12.2.1.2.2 Process System Radiation Sources 12.2.1.2.2.1 Introduction. The following process systems govern the shielding requirements within the reactor and turbine buildings:

a.

Recirculation (RRC),

b.

Reactor water cleanup (RWCU),

c.

Reactor core isolation cooling (RCIC),

d.

Residual heat removal (RHR),

e.

Fuel pool cooling and cleanup (FPC),

f.

Main steam (MS) and the reactor feedwater system (RFW),

g.

Traveling in-core probe (TIP), and

h.

Offgas system (OG).

The locations of the equipment, which are part of these systems, are shown on the radiation zone drawings, Figures 12.3-5 through 12.3-18.

12.2.1.2.2.2 Recirculation System Sources. The coolant activation products, principally 16N, are the dominant sources of radiation in the RRC system during normal operation. The 16N introduced into this system decays significantly by the time it returns to the core. Therefore, decay is considered in the portions of this system where elapsed transit time is significant.

For shielding purposes, the pipes in the RRC system are treated as equivalent line sources.

The 16N source strength in these lines varies from 0.10 Ci/cm at the line from the reactor vessel to 0.02 Ci/cm at the header which distributes coolant flow back to the reactor. These sources are located within the drywell of the primary containment of the reactor building, from approximately el. 501 ft to el. 540 ft.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-3 During shutdown, fission products become the dominant source. This shutdown source is considerably less than that of 16N during normal operation. Shielding design is based on the 16N source, which is more than adequate to shield against the fission product shutdown source.

12.2.1.2.2.3 Reactor Water Cleanup System Sources. The major sources of radiation in the portions of the RWCU system in the reactor building during normal operation are the coolant activation products, principally 16N. The 16N source strength (given in activity per unit length of line) in the RWCU system ranges from 1.00 x 10-3 Ci/cm in the lines flowing into the cleanup recirculation pumps to 4.68 x 10-8 Ci/cm at the exhaust of the tube side of the nonregenerative heat exchanger. Returning from the radwaste building, the 16N source strength ranges from 3.08 x 10-10 Ci/cm to negligible (less than 10-14 Ci/cm).

The 16N source strengths in the regenerative and nonregenerative heat exchangers are

a.

Tube side of the regenerative heat exchanger: 2.69 x 10-6 Ci/cm3,

b.

Tube side of nonregenerative heat exchanger: 6.24 x 10-8 Ci/cm3, and

c.

Shell side of the regenerative heat exchanger: 1.70 x 10-14 Ci/cm3.

These heat exchangers are treated as cylindrical sources for input into the QAD computer program (see Section 12.3.2). The portions of this system located in the radwaste building are discussed in Chapter 11. The RWCU pumps are located at el. 522 ft 0 in. and the RWCU regenerative and nonregenerative heat exchangers are located at el. 548 ft 0 in.

During shutdown, the fission products are the dominant radiation source. Since the shielding is designed to shield against the 16N source during normal operation, it is more than adequate to shield against the shutdown fission product source.

12.2.1.2.2.4 Reactor Core Isolation Cooling System Source. The dominant source of radioactivity in the RCIC system is the 16N present in the main steam used to drive the RCIC turbine. This 16N source occurs when the RCIC turbine is tested on a monthly basis. The shielding design is based on this source. The transit time is assumed to be negligible in this system.

The resulting 16N activity (given in activity/unit length of line) in the inlet line is 2.96 x 10-4 Ci/cm and in the outlet line, it is 6.57 x 10-5 Ci/cm. These sources are treated as equivalent line sources for shielding purposes.

The RCIC turbine source strength is 8.44 x 10-2 Ci of 16N. It is treated as a point source since the distance to the dose points is much greater than the turbine dimensions. The RCIC pump and turbine is located at el. 422 ft 3 in. of the reactor building.

12.2.1.2.2.5 Residual Heat Removal System Sources. The RHR system radiation sources consist of the fission and corrosion products. Table 12.2-5 lists the gamma ray energy

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-4 spectrum of the radionuclides in the RHR pumps, pipes, and heat exchangers 4 hr after shutdown. These sources are based on the RHR system operating in the normal reactor shutdown cooling mode. The fission and corrosion product isotope concentrations used are listed in Tables 11.1-2 through 11.1-4.

The RHR heat exchangers are located approximately from el. 559 ft 0 in. to el. 589 ft 0 in. on the west side of the reactor building. The RHR pumps are located at el. 422 ft 3 in. on the west side of the reactor building.

The pipes in this system are treated as equivalent line sources. The heat exchangers are treated as cylindrical sources.

12.2.1.2.2.6 Fuel Pool Cooling and Cleanup and System Sources. The primary sources of radioactivity in the spent fuel assemblies, which are stored in the fuel pool, are the fission products. Table 12.2-6 lists the gamma ray energy spectrum for the spent fuel sources for shutdown time of 2 days. The refueling area is located at el. 606 ft 10.5 in. in the reactor building.

These source terms are calculated using the Perkins and King data (Reference 12.2-2). The shielding calculations are done using the QAD point kernel code (Reference 12.2-3). The following assumptions are used in determining the shielding requirements:

a.

After radioactivity has reached equilibrium in the fuel assemblies, it is assumed that the reactor is shut down and the whole core is moved, within 2 days, into the spent fuel pool;

b.

The whole core and another one-fourth of a core from the last refueling are located by the north wall of the spent fuel pool to give the most conservative dose rate on the outside of the wall. Less water exists between the assembly racks and the north wall than between the assembly racks and any other side of the pool. The assemblies from past refuelings do not add to the shielding requirements because they have decayed for more than 1 year, they are shielded by pool water, and they provide self shielding; and

c.

The water, racks, spent fuel, and other constituents that are located within the array of spent fuel assemblies are homogenized for the purpose of determining the required values of the linear attenuation coefficients.

The minimum depth of water needed to adequately shield the refueling area from the spent fuel assemblies is calculated. It is found that the elevated fuel assembly during the fuel transfer operation controls the dose rate to the refueling area and, hence, the water depth in the spent fuel pool. The fuel assembly is treated as a cylindrical source geometry for the purpose of computing the water depth.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-5 The source strength used to determine the shielding requirements for the dryer-separator pool is based on a contact dose rate for the separator of 10 R/hr. The average gamma ray energy is approximately equal to 1 MeV.

12.2.1.2.2.7 Main Steam and Reactor Feedwater Systems Sources. In these systems, the dominant sources of radioactivity are the coolant activation products, principally 16N. The following equipment is considered:

a.

Moisture separators and reheaters (MSR),

b.

Main condenser and hotwell,

c.

Feedwater heaters, and

d.

The piping associated with these systems.

The moisture separator and reheaters govern the shielding requirements on the operating floor (el. 501 ft 0 in.) of the turbine building. The 16N source strengths for the first stage reheater tubes, the second stage reheater tubes, and the MSR plena are listed on Table 12.2-7. The first and second stage reheater tubes are approximated by rectangular parallelepipeds. The plena are divided into an array of rectangular parallelepipeds and cylinders, depending on their physical arrangement.

The 16N source strength in the main condenser is 6.0 x 10-8 Ci/cm3. This is based on the incoming flows from the exhaust of the reactor feed pump turbine, and the exhaust steam from the low pressure turbine. The main condenser is treated as either a truncated cone or infinite slab depending on the view angle and distance from the condenser to the dose point.

Since most of the 16N exists as a noncondensable gas, it comes out of solution in the main condenser. Thus, the radiation sources in the hotwell are the fission or corrosion product radionuclides. Table 12.2-8 lists the gamma ray volumetric source strengths calculated from the hotwell. The hotwell is treated as an infinite slab.

The 16N source strength of feedwater heater 6 listed in Table 12.2-9, governs the shielding requirements on the mezzanine floor of the turbine building (el. 471 ft 0 in.). The feedwater heater is treated as an array of rectangular parallel pipes and cylinders for input into QAD.

Table 12.2-10 lists the 16N source strengths in selected steam piping in the MS and RFW systems.

12.2.1.2.2.8 Offgas Sources in the Turbine Generator Building. Table 12.2-11 lists the source strength from the offgas system equipment in the turbine building. Nitrogen-16 is the dominant radionuclide present in this system. The offgas equipment is located at el. 441 ft 0 in. of the turbine building.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-6 12.2.1.2.2.9 Traveling In-Core Probe System Sources. The primary source of radiation in the TIP system is the 56Mn in the traversing in-core probe (TIP) cable. The average source strength per unit length of cable is 3.27 x 104 Ci/cm. This is calculated using an exposure time of 864 sec. The average radioactivity emitted per unit length is calculated using a cosine distribution for the neutron flux in the axial direction of the core. The cable is treated as a line source for shielding purposes. The TIP components are located at el. 501 ft 0 in. of the reactor building.

12.2.1.2.2.10 Sources Resulting From Crud Buildup. Fission and corrosion product deposition is complex to describe analytically. In estimating the effects of this source, operating experience from other BWR plants is used (Reference 12.2-4); Section 12.4 discusses this in greater detail.

12.2.1.3 Radwaste Building The radiation sources present in the radwaste building are discussed in Chapter 11.

12.2.1.4 Byproduct, Source and Special Nuclear Materials A list of all byproduct, source and special nuclear material in sealed sources with an activity greater than 100 mCi is given in Table 12.2-12a. Several areas outside the Reactor, Turbine, and Radwaste Buildings and within the site area have been established for use and storage of radioactive material in the form of activated components, sealed sources, and contaminated tools, equipment, and protective clothing. Those areas where the total quantity of activity could at times be greater than 100 mCi under normal conditions are listed in Table 12.2-12b.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 12.2.2.1 General Design features that limit the airborne radioactivity in normally occupied areas are incorporated into the plant. Areas that are designated as radiation Zones I and II are considered to be normally occupied areas.

The plant was designed so that the airborne radionuclide concentrations in normally occupied areas are well below the limits specified in 10 CFR Part 20. The criteria for Zones I and II are specified in 10 CFR 20, Appendix B to 20.1001-20.2401, Table 1, Column 3.

No radiation Zone I areas exist in the reactor or turbine generator building. The only Zone I areas found in the radwaste and control building are the main control room and the counting room shown on Figures 12.3-14 and 12.3-13, respectively. The main control room is located at el. 501 ft 0 in. It has a separate heating, ventilating, and air conditioning (HVAC) system which brings in fresh air from the outside. See Section 9.4.1 for further discussion. The

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 12.2-7 counting room is located at el. 487 ft 0 in. As seen in Figure 9.4-3, the incoming ventilation flow into this room consists of fresh air. The outgoing air flow is to the area surrounding the counting room. It is concluded that the airborne concentration in the counting room is small.

See Section 12.2.2.3.5 for discussion on the contribution of sampling and radiochemical analysis on airborne radioactivity levels within this area.

12.2.2.2 Model for Computing the Airborne Radionuclide Concentration in a Plant Area The model used for computing the airborne radionuclide concentration is based on the continuous leakage of a radioactive fluid into a plant area. The removal of radionuclides from this area is through ventilation exhaust and decay. The equation which yields the airborne radionuclide concentration in a plant area is:

C A q PF i

q q

V i

i s

i v

a i

a

(

)

exp (

/

)

1 t

(12.2-1) where:

Ci

= concentration of radionuclide i in a given plant area (ci/cm3)

Ai

= concentration of radionuclide i in the fluid (mCi/g) qs

= rate of radionuclide leakage into an area (g/minute)

(PF)i = partition factor for radionuclide i (dimensionless) i

= decay constant for isotope i (1/minute)

V

= volume of area (cm3) qa

= HVAC air flow rate out of area (cm3/minute) t

= time interval between start of leak and calculation of concentration (minute)

The equilibrium value of Ci is given by C

A q PF V

q i

i s i

i a

(

)

(12.2-2)

Note that in all analyses where these two equations are used, the partition factor is set equal to 1.0. This yields conservative results.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 LDCN-01-069 12.2-8 12.2.2.3 Sources of Airborne Radioactivity The potential sources of airborne radioactivity found in the plant are as follows:

a.

Leakage from process equipment in radioactive systems, such as valves, flanges, and pumps,

b.

Sumps, drains, tanks, and filter/demineralizer vessels which contain radioactive

fluid,

c.

Exhaust from relief valves,

d.

Removal of reactor pressure vessel (RPV) head and associated internals,

e.

Radioactivity released from sampling, and

f.

Airborne radioactivity released from the spent fuel pool water and spent fuel movement.

Sections 12.2.2.3.1 through 12.2.2.3.6 discuss each of these sources and their effect on the airborne radionuclide concentration in normally occupied areas of the plant. All design features that serve to reduce the airborne radionuclide concentration are also discussed.

12.2.2.3.1 Effect of Leakage from Process Equipment in Radioactive Systems Leakage into normally occupied plant areas from radioactive process systems is described by three parameters.

The first parameter is the location of the source of radioactive leakage. If this source is located in a Zone III or Zone IV radiation area, it does not contribute to the airborne radionuclide concentration in a normally occupied area because the air flow is from areas of potentially low radioactivity contamination to areas of potentially high radioactivity contamination. Any leakage of radioactive airborne contamination into a Zone III or Zone IV radiation area in the reactor, radwaste and control or turbine generator building is exhausted from that area through the HVAC system and is not transported into a normally occupied area. See the HVAC flow diagrams for the reactor, radwaste and control or turbine generator buildings, Figures 9.4-2, 9.4-3, and 9.4-6, and the radiation zone drawings, Figures 12.3-5 through 12.3-18.

Areas with multiple zone designation are regarded as having a high radioactivity contamination potential. Air from these areas is either exhausted to another Zone IV area or to the HVAC exhaust equipment. Thus, any radioactive leakage that occurs within an area with a multiple zone designation should not affect the airborne radionuclide concentration of a normally occupied area.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 LDCN-01-069 12.2-9 The second parameter is the rate at which the radioactive leakage is introduced into an area.

Any system that operates continuously is potentially a greater source of airborne radioactivity than a system that operates periodically when radionuclide concentration and leakage rate are the same for each system. The operating pressure of the system is another consideration which affects the leakage rate. A system, such as the main steam system, is expected to have a higher leakage rate than a system which operates at atmospheric pressure. However, radioactive systems which operate at high pressure are generally located in Zone IV areas.

Thus, these systems do not significantly contribute to the airborne radioactivity level in normally occupied areas. This is due to the HVAC air path which was discussed earlier.

The third parameter is the radionuclide concentration in the leaking fluid. A system that is leaking highly radioactive fluid such as reactor coolant or main steam is of greater concern, initially, than a system which is leaking condensate storage tank water. Those systems which can leak highly radioactive fluid are located in either a Zone III or Zone IV radiation area and the HVAC air flow paths do not transport this radioactivity to a low radiation area, as discussed earlier. The effect of systems which leak fluid with a low radionuclide concentration is discussed below.

A list of all radioactive systems found in the plant is provided in Table 12.2-13. Each radioactive system listed has been checked to determine its contribution to the airborne radioactivity levels in normally occupied areas. It is found that most of these systems are located in Zone III or Zone IV areas and do not contribute to the airborne radioactivity levels in normally occupied areas, due to HVAC design features as explained earlier. Some of the radioactive systems are located within areas which have a multiple radiation zone designation; e.g., Zone IV during system operation and Zone III during system shutdown. The systems in these areas do not contribute to the airborne radioactivity levels due to HVAC system design features, as explained earlier. Any system which is not entirely located in a Zone IV, Zone III, or multiple zone radiation area and which may contribute to airborne radionuclide levels in normally occupied areas is discussed in the following paragraphs. Those systems which are used only during loss-of-coolant accident (LOCA) conditions are not discussed.

These include the high-pressure core spray (HPCS), low-pressure core spray (LPCS), and main steam leakage control (MSLC) systems. These systems are not regarded as being significant sources of airborne radioactivity since they are used only in a test mode or in an accident situation.

The major source of control rod drive (CRD) leakage which can contribute to the airborne radionuclide levels in a normally occupied area are the CRD pumps located in the CRD pump room. The CRD pump room is located between column lines 3.4/6 and N/N.8 at el. 422 ft 3 in. of the reactor building. Figure 4.6-5 shows suction flow for the CRD pumps is from the outlet of the condensate filter demineralizers or the condensate storage tank. Only one pump is in operation at any time and the leakage rate from this pump is assumed to be 50 gal/day. The incoming HVAC air flow to this area is approximately 3500 ft3/minute. The

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-10 mathematical model described in Section 12.2.2.2 is used to evaluate the airborne activity.

The results are listed in Table 12.2-14. It is seen that the expected airborne radionuclide level is well below the derived air concentrations (DAC) values.

The major source of main condensate leakage which can contribute to the airborne radionuclide levels in a normally occupied area are the condensate pumps. These pumps are located between column lines D/F and 13/15 at el. 441 ft 0 in. of the turbine generator building. The suction flow from these pumps is taken from the main condenser hotwell. Since the suction side of the pumps is at a negative pressure to the surrounding area, no leakage from the pump seals should occur. However, leakage may occur from the valves and other equipment which are located on the discharge side of the pump. It is conservatively assumed that the leakage is 1 gpm. The airborne radionuclide concentration in the area where the condensate booster pumps and condensate pumps are located is listed in Table 12.2-15.

The offgas equipment which is radioactive is located in Zone IV areas. The exception is the afterfilter, which is located between column lines K.1/L.9 and 14.7/15.4 at el. 452 ft 0 in. of the radwaste building. This filter is not a source of airborne radioactivity. The reasons are that the filter operates at approximately atmospheric pressure and the radioactivity contained within the filter is small. See Section 11.3 for details on offgas system operation.

12.2.2.3.2 Effect of Sumps, Drains, Tank, and Filter Demineralizer Vents The equipment drain (EDR), floor drain (FDR), and miscellaneous radwaste (MWR) systems are designed to collect and process various types of liquid radioactive waste generated in the reactor, radwaste, or turbine generator buildings, as explained in Section 11.2. In each of the aforementioned buildings, there is a network of EDR, FDR, and MWR drains which conduct radioactive liquid waste to an EDR, FDR, and MWR sump, respectively. These drains and sumps are not significant sources of airborne radionuclides for the following reasons:

a.

Each of the EDR, FDR, and MWR sumps present in the reactor, radwaste, and turbine generator buildings is covered with a steel plate. The free volume in the sump is maintained at a negative pressure with respect to the surrounding area by use of a riser vent, which is connected to the HVAC system in the building of concern. The steel plate covering the sump does not provide an airtight seal. However, air is drawn into the sump, then through the riser vent and is exhausted to the HVAC system. Any radionuclides that escape into the free volume of the sump are discharged to the HVAC system and do not escape into the area surrounding the sump; and

b.

The EDR, FDR, and MWR drains are connected to their respective sumps through the same riser vents discussed in the preceding paragraph. Some of the drains employ loop seals, which prevent radioactive gases from escaping into the areas around the location of the drains. Other drains do not employ loop

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-11 seals, but since the riser vent is connected to the HVAC system, air will be drawn into the drain through the riser vent and out to the HVAC system.

The tanks and filter demineralizer vessels that contain significant inventories of radionuclides are vented to the HVAC system. These tanks and filter demineralizer vessels are located in Zone III or Zone IV radiation areas. Even if any airborne radionuclides were released from these tanks or filter demineralizers, there would be no effect on normally occupied areas due to the HVAC system design features, which are explained in Section 12.2.2.3.1.

12.2.2.3.3 Effect of Relief Valve Exhaust The relief valves found in the various plant systems which can exhaust radioactive fluids are not considered to be a significant source of airborne radioactivity in normally occupied areas.

The reasons are as follows:

a.

All relief valves (except the main steam safety relief valves), which relieve pressure in the turbine main steam or bleed systems, exhaust directly to the condenser, and

b.

All relief valves, which relieve pressure in a system which contains radioactive fluid, either exhaust directly into the suppression pool, an equipment or floor drain, or into a line which is part of the system in question.

With reference to the equipment or floor drain systems, the receiving point for relief valve discharges is in a radiation zone equivalent or higher than the equipment being relieved. For discharge back to the system, the same is true.

The main steam safety relief valves, which are located within primary containment, provide pressure relief to the main steam lines coming from the reactor. These valves exhaust directly to the suppression pool. The effect of main steam relief valve blowdown in normally occupied areas in secondary containment is analyzed by assuming that all radionuclides that are present in the main steam blowdown are released to the primary containment air. The radionuclide distribution within the free volume of the primary and secondary containment is assumed to be uniform. The radionuclide concentration in secondary containment becomes, in mCi/cm3:

C R q t A q

R V R

q V

t sc b

i v

i v

sc

,i (exp

(

)

/

) )

=

+

+

b sc i

t - exp - (

(12.2-3) where:

R

= primary containment leakage constant (1/minute) qb

= main steam blowdown flow (g/minute)

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-12 tb

= duration of blowdown flow (minute) qv

= ventilation flow rate out of secondary containment (cm3/minute)

Vsc

= volume of secondary containment (cm3) i

= decay constant for isotope i (1/minute) t

= time after blowdown event Csc,i

= airborne radionuclide concentration of radionuclide i in the secondary containment (µCi/cm3)

Ai

= radionuclide concentration in blowdown fluid (µCi/g)

The value of t which yields the maximum value of Csc,i is t

R q

V n

R q

V v

sc i

i v

sc

=

+

+

1 1

/

/

(12.2-4)

The calculated results are based on the occurrence of a main steam isolation valve closure.

This results in all 18 relief valves being actuated for a maximum duration of 40 sec. This event results in the maximum release of radionuclides to primary containment (see Table 5.2-3 and Sections 15.1 and 15.2). The values of the various parameters used in equations 12.2-3 and 12.2-4 are given as follows:

R = 0.5 vol. %/day (Section 3.8.2.3-1) qb = 1.6 x 107 lb/hr = 1.2 x 108 g/minute (Table 5.2-3) tb = 40 sec = 0.67 minute (Table 5.2-3) qv = 9.5 x 104 cfm (Table 11.3-6)

Vsc = 3.5 X 106 ft3 (Table 11.3-6)

The values of Ai are based on the information found in Section 11.1.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-13 The maximum airborne concentration of the most important radionuclides in secondary containment from a main steam relief valve blowdown is listed in Table 12.2-16. The concentrations are far below the DAC criteria given in 10 CFR Part 20. It is concluded that the main steam relief valve blowdown has a negligible effect on the airborne radionuclide level in secondary containment.

12.2.2.3.4 Effect of Removing Reactor Pressure Vessel Head and Associated Internals Experience at BWR plants has shown that an inventory of radioactive gases will accumulate inside the reactor vessel head between the time of shutdown and head removal. These gases consist primarily of the longer lived radioiodines and noble gases. To prevent these gases from being released to the refueling area, provisions are made so that the gases are vented to the HVAC system prior to reactor pressure vessel (RPV) head removal. These provisions include the ability to vent the RPV head through a 4-in. flexible hose connection to the reactor building sump vent filter units which contain particulate and charcoal filter components. The connection to the sump vent filter unit is shown in Figure 9.4-2.

Provisions are also made for clean water services to the RPV cavity area. This permits wetting of the RPV cavity and components to minimize airborne contamination. This is done prior to flooding the RPV cavity.

It is anticipated that RPV head and reactor internals removal will have a minimal effect on the airborne radionuclide level in the spend fuel area.

12.2.2.3.5 Effect of Sampling The possibility of releasing radionuclides which could become airborne during sampling operations is recognized. Design features are incorporated into the sample system to limit the radionuclide release. Radioactive liquids that require frequent grab sampling are piped via sample tube lines to fume hoods located in sample rooms. Grab sampling will normally be accomplished in the fume hoods. During sampling, an inflow face air velocity of approximately 100 ft/minute will be maintained to sweep any airborne radioactive particles to the exhaust duct. Administrative control is further exercised in the form of procedures that are followed when process fluids are sampled. This minimizes the release of radioactive fluids and, hence, exposure to personnel during the sampling process.

12.2.2.3.6 Effect of Spent Fuel Movement Experience at four operating BWR plants has shown that fuel movement does not present any unusual radiological problems. The expected level of radioactivity in the spent fuel pool water is listed in Table 3.5-7 of the Environmental Report. This activity is not expected to have any significant radiological effects on the airborne radionuclide levels above the spent fuel pool floor.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-14 12.2.2.3.7 Effects of Solid Radwaste Handling Areas The solid radwaste handling equipment contained Zone IV area between columns N.1 and S (Figure 12.3-11) are designed for remote operation. Entry for maintenance activities will normally entail shut down and flushing of systems and equipment. The solid radwaste handling equipment is operated primary from low background areas of Zone III between columns Q.1 and R.2 near the waste compactor.

The ventilation supply to this Zone III area is clean outside air with air flow into surrounding normally unoccupied areas. The only source of airborne radioactivity, the waste compactor, is vented directly to the filtered exhaust ventilation system.

Airborne radioactivity levels in the normally occupied solid radwaste handling areas will be ambient outside air concentrations.

12.2.2.3.8 Effects of Liquid Radwaste Handling Areas Normally occupied liquid radwaste handling areas include the valve corridor (a Zone III area),

the precoat room and the radwaste control room (Zone II areas) shown in Figure 12.3-12.

This valve corridor is supplied directly with outside air. Components that are operated from this corridor contain radioactive materials which are located in normally unoccupied valve and pump rooms which are served by separate ventilated supply and exhaust. The radwaste control room and the precoat rooms do not house components containing radioactive material.

Although not normally occupied, the possibility exists that entry into pump corridor (a Zone IV area between columns 11.2 and 12.2) (Figure 12.3-11) and valve and pump (East and West side) rooms (Zone IV areas of Figure 12.3-12) could be necessary while systems are operating.

The pump corridor of Figure 12.3-11 contains the highest concentrations of radioactive material and has the lowest air flow to volume ratio, so is taken as a worst case. Equilibrium airborne radioactivity concentration is calculated as described in Section 12.2.2.2 assuming a leak rate of 50 gal/day from the reactor water cleanup phase separator decant pump. The results are shown in Table 12.2-17.

12.

2.3 REFERENCES

12.2-1 Aalto, E. et al., A Users Manual for the NRN Shield Design, AE-145, June 1964.

COLUMBIA GENERATING STATION Amendment 57 FINAL SAFETY ANALYSIS REPORT December 2003 12.2-15 12.2-2 Perkins, J. F. and King, R. W., Energy Release from Decay of Fission Products, Nuclear Science and Engineering, Vol. 3, 1958 and Perkins, J. F.,

U.S. Army Missile Command Redstone Arsenal, Report No. RR-TR-63-11, July 1963.

12.2-3 Malenfant, Richard E., QAD: A Series of Point-Kernel General Purpose Shielding Program, National Technical Information Service, LA-3573, October 1966.

12.2-4 Butrovich, R. et al., Millstone Nuclear Power Station, Refueling/Maintenance Outage, GE NEDO-20761, Class 1, Fall 1974.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-17 Table 12.2-1 Basic Reactor Data for Source Computations (During Plant Operation)

Reactor thermal power 3486 MW Overall average core power density 51.6 w/cm3 Core power peaking factors At core center:

Pmax Pave Z (axial) 1.5 Pmax Pave R (radial) 1.4 At core boundary:

Pmax Pave Z (axial) 0.5 Pmax Pave R (radial) 0.7 Core volume fractions:

Material Density (g/cm3)

Volume Fraction UO2 10.4 0.254 Zr 6.4 0.140 H2O 1.0 0.274 Void 0

0.332 Average water density between core and vessel below the core 0.74 g/cm3 Average water-steam density above core In the plenum region 0.23 g/cm3 Above the plenum (homogenized) 0.6 g/cm3 Average steam density 0.036 g/cm3

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-18 Table 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary Energy Range (MeV)

Neutron Flux (Neutrons/cm2 sec) 17.9-14.0 7.13E9 14.0-12.0 2.37E9 12.0-10.0 1.79E10 10.0-9.0 2.37E10 9.0-8.0 4.69E10 8.0-7.0 1.17E11 7.0-6.0 3.45E11 6.0-5.0 6.57E11 5.0-4.0 1.23E12 4.0-3.0 2.34E12 3.0-2.5 2.04E12 2.5-2.0 1.27E12 2.0-1.5 2.97E12 1.5-1.0 5.63E12 1.0-0.7 3.18E12 0.7-0.5 3.92E12 0.5-0.3 4.15E12 0.3-0.1 5.62E12 0.1-0.03 3.50E12 0.03-0.01 2.31E12 1.0(-2)-1.0(-3) 3.76E12 1.0(-3)-1.0(-4) 3.07E12 1.0(-4)-1.0(-5) 2.40E12 1.0(-5)-1.0(-6) 1.94E12 1.0(-6)-1.0(-7) 1.50E12 1.05(-7)-thermal 2.58E12

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-19 Table 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation Energy Range (MeV)

Mid-Range Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 10.0-7.0 8.5 5.97E11 7.0-5.0 6.0 4.00E11 5.0-3.0 4.5 4.77E12 3.0-2.0 2.5 6.83E12 2.0-1.0 1.5 1.02E13 1.0-0.0 0.5 1.25E13

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-20 Table 12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown Energy Range (MeV)

Average Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec)

>2.60 3.00 8.70E11 2.60-2.20 2.40 4.24E11 2.20-1.80 2.00 2.63E11 1.80-1.35 1.58 1.15E12 1.35-0.90 1.13 1.40E12 0.90-0.40 0.65 1.46E12 0.40-0.00 0.20 2.83E11

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-21 Table 12.2-5 Fission Product Source in RHR Piping and Heat Exchangers 4 Hours After Shutdown Energy Range (MeV)

Average Energy (MeV)

Energy Release (MeV/cm3 sec) 4.00-3.00 3.50 2.30E2 3.00-2.60 2.75 3.70E2 2.60-2.20 2.40 8.20E2 2.20-1.80 2.00 1.30E3 1.80-1.35 1.58 4.80E3 1.35-0.90 1.13 5.00E3 0.90-0.40 0.65 9.20E3 0.40-0.10 0.25 1.80E3

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-22 Table 12.2-6 Gamma Ray Energy Spectrum For Spent Fuel Sources (One Core)

Energy Range (MeV)

Average Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 2 Days After Shutdown

>2.60 3.00 2.54E8 2.60-2.20 2.40 9.89E9 2.20-1.80 2.00 4.93E9 1.80-1.35 1.58 1.34E11 1.35-0.90 1.13 2.59E10 0.90-0.40 0.65 2.97E11 0.40-0.00 0.20 5.17E10

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-23 Table 12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater Component Radioactivity Concentration (Ci/cm3)

Moisture separators and reheaters (MSR)

Plena 3.48E-7 First stage reheater tube bundle (east end of MSR) 7.23E-7 First stage reheater tube bundle (west end of MSR) 5.91E-7 Second stage reheater tube bundle (east end of MSR) 1.43E-6 Second stage reheater tube bundle (west end of MSR) 1.14E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-24 Table 12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell Group Average Group Energy (MeV)

Volumetric Energy Release Rate (MeV/cm3 sec) 1 3.50 3.82E1 2

2.80 7.92E1 3

2.40 1.43E2 4

2.00 1.24E2 5

1.57 3.94E2 6

1.12 3.00E2 7

0.65 6.71E2 8

0.20 8.26E1

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-25 Table 12.2-9 Nitrogen-16 Source Strength in Feedwater Heater 6 Radionuclide Concentration (Ci/cm3)

Feedwater Heater Steam Water 6

4.93E-7 8.40E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-26 Table 12.2-10 Nitrogen-16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems Point of Interest Line Source (Ci/cm)

Input lines to high pressure turbine 7.60E-3 Extraction steam line from high pressure turbine to FWH 6A 7.50E-4 Crossunder piping from high pressure turbine to main steam relief 5.50E-4 Crossover piping from main steam relief to low pressure turbine 3.80E-4 Extraction steam line from low pressure turbine to FWH 1A 7.50E-5 Extraction steam line from low pressure turbine to FWH 2A 1.20E-4 Extraction steam line from low pressure turbine to FWH 3A 1.40E-4 Extraction steam line from low pressure turbine to FWH 4A 1.50E-4 Heater drain line from FWH 6A to FWH 5A 2.30E-5 Heater drain line from FWH 5A to FWH 4A 1.01E-6

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-27 Table 12.2-11 Offgas System Sources in the Turbine Generator Building Component 16N Source Strength

(µCi/cm3)

Steam jet air ejector (first stage) 2.3E0 Intercondenser 4.2E1 Steam jet air ejector (second stage) 3.6E0 Preheater 2.8E0 Recombiner 2.3E0 Offgas condenser 3.7E1 Water separatora 2.7E1 a The preheater, recombiner, offgas condenser, and water separator are located in the same room.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-28 Table 12.2-12a Special Sources With Strength Greater Than 100 Millicuries Isotope Identification Form Quantity (mCi)

Use/Location 241AmBe 2-81-020 Solid 14,770 Calibration (EOF) 241AmBe 2-82-050 Solid 2,845 Neutron source (plant) 241AmBe 2-86-073 Solid 988 WNP-1 source (plant) 137Cs 2-79-033 Solid 524 Panoramic shepard cal (EOF) 137Cs 2-83-097 Solid 4,673 ARM calibration (plant) 137Cs 2-84-058 Solid 1,474 Shepard series 28 cal (EOF) 137Cs 2-88-002 Solid 133 Victoreen model 878-W cal (plant) 137Cs 2-93-026 Solid 909 MG calibrator (EOF) 137Cs 2-99-028 Solid 2,078 Mini-89 calibrator (plant) 137Cs08-132 Solid 358,600 Hopewell calibrator (EOF) 137Cs08-133 Solid 422 Hopewell calibrator (EOF) 137Cs13-230 Solid 12,940 ARM calibration (plant) 238PuBe 2-84-047 Solid 11,150 WNP-3 startup source (plant) 238PuBe 2-84-048 Solid 11,260 WNP-3 startup source (plant)

Table as of 9/9/2015.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-14-005 12.2-28a Table 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area Area Location Approximate Size (sq. ft.)

Normal Contents Normal Activity (mCi)

LSA Storage Pad Outside SW side of Radwaste Bldg 1800 Low-activity dry and compacted solid wast containers 930 Building 13 West portion of Bldg 13 2700 Contaminated protective clothing 300 Warehouse 5 NE portion of Bldg 80 at Snake River Warehouse Complex 4000 Radioactive &

contaminated equipment 590 Building 167

~0.5 miles E of Plant 6332 Radioactive &

contaminated equipment 1370 Building 167 Storage Yard

~0.5 miles E of Plant 68000 C-van storage 1620 Outage Storage Yard Fenced area W of Bldg 105 2400 Outage radioactive material & C-van storage 114 Kootenai HP Calibration Lab Kootenai (Bldg

34) Rms 102 &

102A 600 Calibrators/irradiators, calibration sources, radioactive HP instruments 377030

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-29 Table 12.2-13 List of Radioactive Piping and System Designations Air removal (AR)

Bleed steam (BS)

Condensate filter/demineralizer (CPR)

Condenser vents and drains (CND)

Control rod drive (CRD)

Equipment drains radioactive (EDR)

Exhaust steam (ES)

Floor drains radioactive (FDR)

Fuel pool cooling (FPC)

Heater drains (HD)

Heater vents (HV)

High pressure core spray (HPCS)

Low pressure core spray (LPCS)

Main condensate before condensate demineralizers (COND)

Main steam (MS)

Main steam isolation valve leakage control system (MSLC)

Miscellaneous waste radioactive (MWR)

Offgas (OG)

Process sample radioactive (PSR)

Process vents (PVR)

Process waste radioactive (PWR)

Reactor core isolation cooling (RCIC)

Reactor recirculation (RRC)

Reactor water cleanup (RWCU)

Relief valve vents radioactive (VR)

Residual heat removal (RHR)

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-03-040 12.2-30 Table 12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building)

Radionuclide Airborne Concentration Ci (Ci/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 1.9E-11 2E-8 1E-3 132I 2.5E-12 3E-6 1E-6 133I 2.2E-11 1E-7 2E-4 134I 1.9E-12 2E-5 2E-7 135I 8.7E-12 7E-7 1E-5 83Br 3.3E-13 3E-5 1E-8 84Br 6.3E-14 2E-5 3E-9 85Br 1.3E-16 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-31 Table 12.2-15 Airborne Radionuclide Concentration in Condensate Pump Area (el. 441 ft. 0 in. turbine generator building)

Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 4.2E-10 2E-8 2E-2 132I 3.8E-9 2E-6 3E-3 133I 2.9E-9 1E-7 2E-2 134I 7.4E-9 2E-5 4E-4 135I 4.2E-9 7E-7 6E-3 83Br 4.8E-10 3E-5 2E-5 84Br 8.2E-10 2E-5 4E-5 85Br 3.2E-10 1E-7 3E-3 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-32 Table 12.2-16 Airborne Radionuclide Concentration in Secondary Containment from a Main Steam Relief Valve Blowdown Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 131I 3.0E-11 2E-8 2E-3 133Xe 5.0E-10 1E-4 5E-6 a 10 CFR 20, Appendix B to 20.1001-20.2401, Table I, Column 3.

COLUMBIA GENERATING STATION Amendment 53 FINAL SAFETY ANALYSIS REPORT November 1998 12.2-33 Table 12.2-17 Airborne Radionuclide Concentration in Liquid Radwaste Handling Area Radionuclide Airborne Concentration Ci (µCi/cm3)

Derived Air Concentration (DAC)a (mCi/cm3)

Ratio of Ci to DAC 140Ba 5.8E-10 6E-7 1E-3 140La 6.5E-10 6E-7 1E-3 239Np 2.2E-10 9E-7 2E-3 58Co 9.8E-10 3E-7 3E-3 89Sr 4.8E-10 6E-8 1E-2 99Mo 2.6E-10 6E-7 4E-4 99MTc 1.7E-10 6E-5 3E-6 132Te 1.5E-10 9E-8 2E-3 131I 9.2E-10 2E-8 4E-2 132I 2.4E-10 3E-6 1E-4 133I 4.1E-10 1E-7 4E-3 135I 1.8E-10 7E-7 2E-4 a 10 CFR 20.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-037 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES Columbia Generating Station plant incorporates the design objectives and the design features guidance given in Regulatory Guide 8.8 to the extent discussed in Section 12.1.1. Examples of these features are discussed in Section 12.3.1.3.

Figures 12.3-5 through 12.3-18 show the general arrangement for each of the plant buildings.

In addition, these figures show the shielding arrangement, radiation zone designations for both normal operation and shutdown conditions, controlled access areas, personnel and equipment decontamination areas, location of the health physics facilities, location of area radiation monitors, location of radwaste control panels, location of the counting room, and location of the onsite laboratory for analysis of chemical and radiochemical samples. The counting room is located at el. 487 ft 0 in. of the radwaste and control building (see Figure 12.3-13). The design basis radiation level within the counting room is 0.1 mrem/hr during normal operation.

Plant areas, as identified in Section 12.3 figures, are categorized as design radiation zones according to expected maximum radiation levels and anticipated personnel occupancy with consideration given toward maintaining personnel exposures ALARA and within the standards of 10 CFR 20.

12.3.1.1 Radiation Zone Designations The design basis criteria used for each zone are given below, and the plant layout including major equipment, locations, and radiation zone designations are shown in Figures 12.3-5 through 12.3-18.

For purposes of radiation exposure control, a restricted area is defined in 10 CFR 20, paragraph 20.1003, and plant procedures.

Maximum Dose Rate Zone (mrem/hr)

Design Bases Criteria I

1.0 Unlimited occupancy.

II 2.5 Unlimited occupancy for plant personnel during the normal work week.

III 100.0 Design base occupancy less than 1 hr per week.

Posted zones and controlled entries.

IV Unlimited Positive access control. Controlled entry and occupancy.

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-037 12.3-2 Each access point to every Zone IV area may be secured by locked door or other positive control method while it is a high radiation area. Occupancy of such areas is limited in frequency and duration and entrance must be authorized in advance by the Radiological Operations Supervisor or his representative.

An area survey of radiation levels will be conducted prior to first entry of Zone IV areas to determine the maximum habitation time.

12.3.1.2 Traffic Patterns Access control and traffic patterns in the plant have been evaluated to maintain personnel radiation exposures ALARA and to minimize the spread of contamination.

Normal entry into the plant is as follows:

a.

Personnel normally access the plant Protected Area through the security Protected Area Access Point (PAAP).

b.

The main Radiologically Controlled Area (RCA) normally includes the reactor building, turbine generator building, radwaste building, and diesel generator building. Normal access to these areas is through one of two Health Physics control points located at each end of the main plant corridor.

12.3.1.3 Radiation Protection Design Features Section 12.1.2 discusses the design objectives for the plant. Examples showing the incorporation of these objectives are provided in the following sections.

12.3.1.3.1 Facility Design Features Filters and Demineralizers Liquid radioactive waste and other process streams containing radioactive contaminants are processed through filters and demineralizers. The pressure-precoat type of filter is used in the major fluid processing systems. Cartridge type filters are used in a few select instances such as in the control rod drive (CRD) system. In the case of demineralizers, either the pressure-precoat or deep-bed type of demineralizer is employed.

Each filter and demineralizer is located in a shielded cubicle at el. 487 ft 0 in. of the radwaste and control building. These filters and demineralizers are accessible through the ceiling of the cubicle by removing the shielding plug at el. 507 ft 0 in. This minimizes the radiation

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 12.3-3 exposure to plant personnel from adjacent sources. After removal of the shielding plug, the filter or demineralizer can be serviced remotely by use of special tools designed for this purpose. If it is necessary, the filter or demineralizer can be removed from the cubicle to a work area for servicing. There are overhead cranes provided for the purpose of shielding plug and filter or demineralizer vessel removal.

Each pressure precoat type filter or demineralizer has its own support equipment such as a holding pump, process control valves, and precoating equipment. The holding pump and process control valves associated with each filter or demineralizer is located in the valve gallery. This valve gallery is located on the east side of the radwaste-control building, el. 467 ft 0 in., and is a Zone III radiation area (Figure 12.3-12).

The holding pump and motor-operated valves can be operated from control panels located in Zone III radiation areas. Manually operated process control valves are operated by use of reach rods that pass through the shielding walls into the corridor. This corridor is a Zone III radiation area. With the exception of instrument root valves, all pumps and valves can be remotely operated. Instrument root valves are normally open and are not used during normal plant operation. The filter or demineralizer precoat equipment and associated controls, which includes metering equipment, are located in a Zone III radiation area to provide for maintenance access. Each filter or demineralizer may be backwashed with condensate water, chemically cleaned, or purged with air from a remote control panel. Each deep-bed demineralizer has its own support equipment. A gravity feed subsystem transfers the fresh resins from the resin addition tank to the demineralizers. The spent resins are transferred hydropneumatically to the spent resin tank.

All piping routed to and from filter or demineralizer cubicles is located in the valve gallery area or in shielded pipe tunnels. The pipes are routed to avoid unnecessary bends, thus minimizing possible crud buildup.

Specific examples of filters or demineralizers that incorporate the aforementioned design features are the waste collector filter and waste collector demineralizer. A typical layout is shown in Figure 12.3-19.

Tanks All tanks that contain radioactive liquids and solids are categorized into different groups depending on their level of radioactivity. Tanks that contain significant levels of radioactivity are located at el. 437 ft-0 in. of the radwaste and control building.

The tanks that contain radioactive backwash and resins are located in shielded cubicles. These include the condensate phase separator tanks-condensate backwash receiving tank, spent resin tanks waste sludge phase separator tanks, and the reactor water clean up (RWCU) phase

COLUMBIA GENERATING STATION Amendment 55 FINAL SAFETY ANALYSIS REPORT May 2001 12.3-4 separator tanks. These tanks are constructed of either stainless steel or epoxy-lined carbon steel.

The tanks that contain unprocessed liquid radioactive waste, such as the waste collector, floor drain collector, and chemical waste tanks are not located in individually shielded cubicles.

However, as described in Section 12.1.1, administrative controls will be used to keep doses to plant personnel ALARA. The waste collector and the floor drain collector tanks are constructed of carbon steel. The chemical waste tanks are stainless steel.

To measure the liquid level in the aforementioned tanks, a bubbler-type level indicator, coupled with a pressure transducer, provides for remote monitoring of the liquid level.

All tanks described above are vented to the radwaste building heating, ventilating, and air conditioning (HVAC) exhaust system as described in Section 12.2.2. This limits the release of airborne radionuclides to the area surrounding the tank.

Pumps Pumps handling spent demineralizer resins are shielded from the phase separator tank. An example of this is the waste sludge discharge pump which is shielded from the waste sludge phase separator by concrete and block wall. A shutoff valve in the suction line keeps the pump isolated when it is not in use. Following the discharge of a sludge batch from the phase separator tanks, the pump and its associated piping is automatically flushed with condensate water. Thus, when it is not in use, the pump is free of sludge. A condensate supply, controlled by a valve electrically linked to the pump, maintains an external sealing barrier, preventing sludge leakage past the shaft seal during pump operation.

Heat Exchangers Heat exchangers handling radioactive fluids are designed to limit occupational exposures. An example is the cooler condensers whose function is to condense moisture from the offgas process stream. The cooler condensers are located in a separate cell in the radwaste building, as shown in Figure 12.3-11. No personnel access is required during plant operation. The associated valves, other than those used during maintenance operations, are remotely operated.

The glycol coolant which flows through the tube side is kept at a higher pressure than the offgas flowing through the shell side of the condenser. Thus, a tube failure will not result in radioactivity contaminating the glycol cooling loop. Condensate water that is separated from the offgas is returned, via a water loop seal, to the radioactive sump for processing. The 16-ft deep loop seal prevents escape of radioactive gas through the drain connection. An enlarged discharge section in the loop seal protects it against siphoning. The enlarged discharge section also provides for automatic loop seal restoration should its contents be displaced by a temporary pressure surge. Figure 12.3-20 shows schematically the cooler condenser loop seal arrangement.

COLUMBIA GENERATING STATION Amendment 55 FINAL SAFETY ANALYSIS REPORT May 2001 12.3-5 Recirculation Pumps The decontamination concentrator bottoms recirculation pump serves the decontamination solution concentrator. The pump is made of 316 stainless steel to minimize the corrosive effects of miscellaneous chemical waste fluid. The pump is located in a separate cell from the concentrating equipment it serves. A steam/air sparger at the bottom of the concentrator can be used to flush the pump prior to any maintenance operations. The internals of the pump may be removed without disturbing the connecting piping. A double mechanical seal with clean water circulating through it prevents leakage of process liquid past the shaft seal.

The decontamination concentrator bottoms recirculation pump is not used. There are no plans to use the pump.

Evaporators The decontamination solution concentrators use steam to boil off water from the incoming chemical waste, as described in Section 11.2. The heating steam is supplied from an evaporator. As shown in Figure 12.3-21, steam generated from demineralized water flows in a closed loop through the shell side of the evaporator and the shell side of the concentrator heating element. The steam is then circulated to the condensate return tank to be pumped back to the evaporator. The steam in the concentrator heating element is at a higher pressure than the decontamination solution flowing through the tube side. Furthermore, the auxiliary steam system, which provides heating steam to the tube side of the evaporator, is at a higher pressure than the shell side. This arrangement provides an effective barrier against contamination of the house auxiliary steam.

The decontamination solution evaporator system is deactivated. There are no plans to use the system.

Valve Gallery and Valve Operating Stations Valves handling radioactive fluids and requiring manual operation are located in a valve gallery at el. 467 ft 0 in. of the radwaste and control building. These valves are operated from behind a shielding wall with the aid of reach rods. Reach rod penetrations are not in the line-of-sight with major radiation sources, such as resin traps. In addition, the reach rod wall penetrations are grouted about the reach rod assembly, and steel plates are added on both sides of the penetration to minimize radiation exposure. A typical application of reach rods and details of a shielding wall penetration is shown in Figure 12.3-19.

The operating stations for motor-operated valves are located in Zone III radiation areas.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-05-002,05-007 12.3-6 Sampling Areas The location of the sampling areas within the plant is discussed in Section 9.3. Design features of sample areas that reduce occupational exposure are discussed in Section 12.2.2.3.5.

Ventilation Filters and Filter Trains Filters that are installed as part of the HVAC units in the Columbia Generating Station plant are located in an accessible area. Selected filter units are designed so that it is possible to introduce a plastic bag over the filter through a service access opening in the HVAC unit. For other filter units, the filter can be pulled into a plastic bag as the filter is removed from the HVAC unit.

Hydrogen Recombiners The hydrogen recombiners for the offgas system are located in the turbine-generator building.

These recombiners are single-pass devices which do not require process control valves. They are located in a shielded cell and do not require personnel access during operation.

Temperature and pressure in the recombiners are remotely monitored. The recombiners and associated piping are designed to withstand an internal explosion.

12.3.1.3.2 Design Features That Reduce Crud Buildup Design features and considerations are included to reduce radioactive nickel and cobalt production and buildup. For example, the primary coolant system consists mainly of austenitic stainless steel, carbon steel, and low alloy steel components. Nickel content of these materials is low. Nickel and cobalt contents are controlled in accordance with applicable ASME material specifications. A small amount of nickel base material (Inconel 600) is employed in the reactor vessel internal components. Inconel 600 is required where components are attached to the reactor vessel shell, and coefficient of expansion must match the thermal expansion characteristics of the low-alloy vessel steel. Inconel 600 was selected because it provides the proper thermal expansion characteristics, adequate corrosion resistance and can be readily fabricated and welded. Alternate low nickel materials which meet the above requirements and are suitable for long term reactor service are not available. Hardfacing and wear materials having a high percentage of cobalt are restricted to applications where no alternate materials of equally good characteristics are available.

To minimize crud buildup, lines carrying slurries in the reactor water cleanup system and in the radwaste system use extensively self-flushing valves. Furthermore, all of the pipe joints of 2.5 in. and above in size in the reactor water cleanup (RWCU) and radwaste systems are butt welded. Butt welding is also used in the equipment drain processing (EDR) and floor drain processing (FDR) systems for lines of 1 in. and above, except for two 1-in. flow control valves, one 1.5-in. flow control valve, two 2-in. socket welded ball valves, one 1-in. socket

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 LDCN-05-007 12.3-7 welded ball valve, and four 3-in. flanged butterfly valves. Lines of 2.5-in. and larger in the offgas system are also butt welded.

The recirculation system is equipped with decontamination flanges for decontamination of the recirculation pump and associated hardware. Also the cleanup, emergency core cooling system (ECCS), and radwaste systems are provided with decontamination connections to enable decontamination of their hardware. In addition, the RWCU and condensate process filter demineralizers reduce the amount of activation products in these systems. Boiling water reactors (BWRs) do not use high temperature filtration.

Depleted zinc is injected into the feedwater system to displace cobalt from the corrosion product layer in the recirculation system. Iron is injected into the feedwater system to cause the displaced cobalt to adhere to the surface of the fuel rods. This has caused a reduction of exposure rates from the recirculation system.

12.3.1.3.3 Field Routing of Piping All code Group A piping is dimensioned in detail. Other piping, 2 in. in diameter and smaller, carrying radioactive fluids, are not dimensioned in detail but are required to be in specified space envelops for shielding and in-plant exposure considerations. Such piping runs have their inception points and terminal points dimensionally located. These inception or terminal points may be an outlet or inlet of a pump, a junction with a larger pipe, or a tank nozzle. Also located dimensionally are all pipe penetrations through a wall, ceiling, or floor. Radioactive piping routed through lower radiation zones is enclosed within a shielded tunnel when warranted by high expected radiation levels. Radioactive piping 2.5-in. in diameter and larger is detailed dimensionally.

12.3.1.3.4 Design Features That Reduce Occupational Doses During Decommissioning Many of the design facilities which presently exist in the plant can be used to minimize occupational exposure during decommissioning, whether decommissioning is accomplished through mothballing, entombment, removal/dismantling, or any combination of the above alternatives. Such facilities include those used for handling and for offsite shipment of fresh fuel, spent fuel, more sources, contaminated filter elements, resins, and other radioactive wastes. The radioactively contaminated spent fuel pool water can be removed by a portable pump placed into the pool after all spent fuel has been removed from the site and any decommissioning use of the pool is finished. The portable pump will discharge into the skimmer surge tank from where the existing decontamination procedure can be implemented.

The number of man rems due to the airborne radioactivity, that may be introduced by the handling of radioactively contaminated systems, as well as the number of man rems due to contact with the same systems, can be reduced by first decontaminating them. Means exist in the present design where radioactively contaminated systems can be decontaminated chemically

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-8 by remote control and flushed. The plant has a hot machine shop and a hot instrument shop located in the radwaste building where contaminated equipment can be decontaminated under controlled conditions. Provisions have been made in the same building for a future decontamination facility with expanded features.

If decommissioning is accomplished by mothballing, the above provisions will reduce to low levels the occupational radiation exposure to personnel. According to Regulatory Guide 1.86, this method involves putting the facility in a state of protective storage. In general, the facility may be left intact except that all fuel assemblies and the radioactive fluids and waste should be removed from the site.

If entombment is chosen as the method of decommissioning, the previously described plant design facilities are adequate to accomplish the tasks with low occupational radiation exposure to personnel. The additional requirements described in Regulatory Guide 1.86 for sealing all the remaining highly radioactive or contaminated components (e.g., the pressure vessel and reactor internals) within a structure integral with the biological shield after having all fuel assemblies, radioactive fluids and wastes, and certain selected components shipped offsite can be met. The necessary modifications to the Columbia Generating Station primary containment structure are shown in Figure 12.3-22.

Low occupational radiation exposure to personnel can be achieved if the decommissioning method adopted is that of immediate removal/dismantling of the plant. For example, the laydown area, el. 606 ft 10.5 in., in the reactor building is provided with a drainage system.

There the drywell head and the reactor pressure vessel head can be decontaminated if it is required. Furthermore, there is ample space to erect a contamination control envelope around them during their segmentation for transportation and burial. The plant design accommodates the cutting of the reactor internal under water and within contamination control envelopes.

The turbidity of the water can be reduced by using the RWCU system. The only modification that might be required to the existing system could be the installation of a strainer for the removal of large filings or other large size contaminants. The highly radioactive pieces can be transferred under water to the cask loading area in the spent fuel pool by methods similar to loading spent fuel. The airborne radioactivity generated by the dismantling process can be removed by the filtration systems that the contaminated control envelopes can have and/or by the standby gas treatment system (SGTS).

12.3.1.4 Radioactive Material Safety 12.3.1.4.1 Materials Safety Program Columbia Generating Station has a program to ensure the safe storage, handling, and use of sealed and unsealed special nuclear source and byproduct materials. Included in the program are procedures for the following:

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-13-039 12.3-9

a.

Receiving and opening shipments as required by 10 CFR 20.1906,

b.

Storage of licensed materials as required by 10 CFR 20.1801 and 20.1802,

c.

Inventory and control of radioactive materials,

d.

Posting of radioactive material storage areas and tagging of source,

e.

Leak tests - sources are checked for leakage or loss of material at least semiannually, and

f.

Disposal - all licensed material disposals are in accordance with 10 CFR Part 20 requirements or by transfer to an authorized recipient as provided in 10 CFR Parts 30, 40, or 70.

12.3.1.4.2 Facilities and Equipment Facilities are provided for handling unsealed sources, such as the liquid standard solutions used for calibration of plant instrumentation. The radiochemical laboratory is equipped with a negative pressure fume hood with filtered exhaust. The hood work surface is designed to withstand heavy weights so that shielding can be provided in the form of lead brick. Drains from the fume hood are routed to the liquid radwaste system.

Remote handling tools are used as needed for movement of the high level sealed sources from their normal storage containers. Shielding to reduce personnel exposure is provided for these sources when they are not in use and to the extent practicable while they are in use.

Portable radiation and contamination monitoring instrumentation is provided, as described in Section 12.5.2, for surveillance to maintain control of the sources.

12.3.1.4.3 Personnel and Procedures The Columbia Generating Station Radiological Services Manager/Radiation Protection Manager (RPM) is responsible for the control and monitoring of sealed and unsealed source and byproduct materials. The Nuclear Material Manager appointed by the Engineering Manager is accountable for special nuclear materials (SNM). The Chemistry Technical Supervisor is responsible for the minimization of radioactive waste and the preparation, offsite shipment, and disposal of radioactive materials and radwaste. Monitoring during handling of these materials is provided by Radiation Protection. Experience and qualifications of Radiation Protection personnel are described in Section 13.1.

Health Physics requirements and instructions to personnel involved in handling byproduct materials are included in implementing procedures.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-10 12.3.1.4.4 Required Materials Any byproduct, source, or special nuclear materials in the form of reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrument and radiation monitoring equipment calibration, or as fission detectors, will be limited to the amounts required for reactor operation or specific calibration purposes except as noted in the facility operating license.

12.3.2 SHIELDING 12.3.2.1 General The radiation shielding design is in compliance with all NRC regulations concerning permissible radiation doses to individuals in restricted and nonrestricted areas. The guidance provided in Regulatory Guide 1.69 on concrete radiation shields is followed to the extent discussed in Section 1.8. The guidance provided in Regulatory Guide 8.8 on radiation protection is followed to the extent discussed in Section 12.1.2. The plant design and layout optimizes personnel and/or equipment protection. Shielding is supplemented with whatever administrative control procedures are necessary to ensure regulatory compliance. These control procedures include area access restrictions, occupancy limitations, personnel monitoring requirements, and radiation survey practices. Other criteria and considerations are listed in Section 12.1.2.

The shielding design is evaluated under the following conditions of plant operation:

a.

Operation at design power, including anticipated operational occurrences,

b.

Shutdown conditions, with radiation from the subcritical reactor core, spent fuel assemblies, and other sources discussed in Section 12.2, and

c.

Postaccident conditions, including those accident occurrences analyzed in Chapter 15. Emphasis is placed on control room habitability.

The majority of the shielding calculations performed are of the bulk shielding type.

Ordinary concrete, having a density of about 150 lb/ft3, is used for shielding except for special applications. In special applications, water, steel, high density concrete, lead, and permali JN P/3% boron are used.

The effects of mechanical or electrical penetrations in shield walls on radiation exposure to personnel is minimized by locating penetrations to preclude direct view of radiation sources through the penetration. The effect of penetrations in shield walls is also minimized by keeping penetration openings to the smallest practicable size. Penetrations are located away

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-11 from immediate areas with personnel access. When these criteria cannot be implemented, penetrations are offset.

Access into shielded areas is, in general, by labyrinths. Labyrinths are located to preclude direct personnel radiation exposure. Where labyrinths are not practicable, shield doors are used. Knock-out walls for equipment removal are constructed of brick arranged in staggered rows to preclude direct streaming.

Portable and removable shielding devices are used when practical and feasible. Portable shielding devices are easily moved from one location to another. Removable shielding devices are normally used at specific locations and can be removed when necessary. The reactor vessel to fuel pool transfer passage is a location where removable shielding is employed primarily for the protection of personnel working in the drywell. Personnel evaluation of the affected drywell area may be employed instead of, or in conjunction with, the above mentioned shielding.

12.3.2.2 Methods of Shielding Calculations Standard methods are used in computing the required shielding thickness for a given source.

These methods are described in References 12.3-1 through 12.3-4. Specific methods of calculation and the computer codes used in the shielding design are discussed below.

The NRN computer code (Reference 12.3-5) is used to determine the shielding requirements for the core generated neutrons and to calculate the thermal neutron flux used to determine captured gamma sources outside the core. This code is based on a multigroup slowing down and diffusion system corrected by a multigroup first flight or removal neutron source. The neutron cross sections used with this code are from Oak Ridge National Laboratory with modifications which have resulted from comparisons with data in BNL-325 (Reference 12.3-6) and the ENDF/B data libraries.

The QAD-BR computer code, which is based on the QAD-P5 code (Reference 12.3-7), is the basic code used to determine shielding requirements for gamma ray sources. This code provides gamma flux, dose rate, energy deposition, and other quantities which result from a point by point representation of a volume distributed source of radiation. Attenuation coefficients for water, iron, and lead, used in this program are taken from the Engineering Compendium for Radiation Shielding (Reference 12.3-1). Concrete attenuation coefficients are taken from ANL-6443 (Reference 12.3-8).

Any shielding requirements which are not determined with the QAD code were determined by using the methods discussed in the Reactor Shielding Design Manual (Reference 12.3-2). The various sources are reduced to their basic geometric configuration (line, disc, cylinder, sphere, etc.) and the corresponding equations are solved to find the dose. The Taylor exponential form

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-12 of the buildup factor is used in these equations. All required data is taken from Reference 12.3-1.

The criteria for penetration acceptability is based on the radiation zone levels in the areas separated by the wall where the penetration is located. The penetration is analyzed if the penetration passes from Radiation Zone IV to Zones III, II, or I; or, if the penetration passes from Radiation Zone III to Zones II or I. This analysis considers both the scattered as well as the direct radiation. The direct component is calculated using the point source attenuation equation, as described in the Reactor Shielding Design Manual (Reference 12.3-2). The scattered component is calculated using the Chilton-Huddleston equations (Reference 12.3-9).

Compensatory shielding (e.g., labyrinths, steel plate, lead wool) is used as needed, to reduce radiation streaming through penetrations and to protect against localized hot spots.

The general dose rate in each plant area, including contributions from radiation streaming, satisfies the design dose rate specified for that area.

Entrances to shielding cubicles are designed to prevent source radiation from passing directly through the opening. Whenever possible, a labyrinthine entrance is designed to reduce the emerging radiation to a level compatible with the access requirements outside the cubicle. The scattered and direct components of the emerging radiation are calculated by the above methods.

12.3.2.3 Shielding Description 12.3.2.3.1 General The description of the shielding throughout the entire plant is summarized within the following sections. These descriptions are to be used in conjunction with the radiation zone maps, Figures 12.3-5 through 12.3-18, to locate the process equipment which is shielded and to determine the design dose rate.

12.3.2.3.2 Reactor Building The sacrificial shield is an ordinary concrete structure 2 ft thick, lined on the inside and outside by steel plates of a minimum thickness of 0.5 in. each. The sacrificial wall extends between el. 519 ft 2.25 in. and 567 ft 4.5 in.

The biological shield wall protects station personnel in the reactor building from radiation emanating from the reactor vessel. The dose rate at the outer face of the biological shield as well as above the shield plug (above the reactor vessel) is, except at penetrations, less than 2.5 mrem/hr during normal reactor operation. The reactor core is the primary source of radiation, and it is used in computing the above dose rate. The wall is in the shape of a shell of the frustum of a cone, and its composition is ordinary concrete at least 5 ft thick. Inside the biological wall exists the primary containment vessel which has the same shape as the wall.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-13 The primary containment vessel is made of 0.75 in. minimum steel plate. The 16N contained in the recirculation system, the main steam lines, and the water in the vessel below the core along with the fission process in the core constitute the major sources of radiation used to determine the radial dose rate. The shielding arrangement for the other major sources in the reactor building is shown on Figures 12.3-15 through 12.3-18.

Personnel evacuation of the affected drywell area(s) and/or employing removable shielding at the fuel pool passage are two methods used for personnel protection in the drywell during fuel handling operations. The shielding is designed such that radiation levels are no greater than 100 R/hr at contact. Portable locally alarming radiation monitors and/or direct Health Physics monitoring are employed for additional personnel protection.

12.3.2.3.3 Turbine Building In the turbine building, 16N constitutes the major source of radiation and basis for shielding design. It is contained in the turbines, moisture separator reheaters, and the feedwater heater system that are located in the turbine access areas at el. 501 ft 0 in. and 471 ft 0 in. These areas are surrounded by ordinary concrete walls at least 3.5 ft thick. The dose rate to the areas outside these walls is less than 2.5 mrem/hr.

The walls which surround the turbine-generator access area at el. 501 ft 0 in. extend up to 524 ft 0 in. This minimizes the effects of direct radiation streaming at the site boundary.

The shielding arrangement for the major sources of radiation in the turbine building, including those discussed above, are shown in Figures 12.3-5 through 12.3-10.

12.3.2.3.4 Radwaste Building The shielding arrangement for the major sources in the radwaste building are shown in Figures 12.3-11 through 12.3-14.

12.3.3 VENTILATION The plant ventilation systems for the different areas of the plant are designed to meet the requirements of 10 CFR Parts 20 and 50. Gaseous wastes will be released in a controlled manner to environs such that during plant operation, onsite and offsite radioactivity levels are ALARA. The design features which limit and reduce the airborne radioactivity are as follows:

a.

In the reactor, radwaste, and turbine generator buildings the air flow is from areas of low airborne radioactivity potential to areas of higher airborne radioactivity potential. This serves to isolate and segregate airborne radioactivity which may be released due to equipment failure or malfunction and leakage from fluid systems;

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-14

b.

To prevent radioactivity buildup, all ventilation air is supplied to the reactor, turbine, and radwaste buildings on a once through basis;

c.

All cubicles housing equipment which handles radioactivity contaminated material are ventilated at a minimum rate of three volume air changes per hour;

d.

All sinks and chemical laboratory work areas where radioactive samples or materials are handled are provided with exhaust hoods to protect operating personnel from airborne contaminants;

e.

All liquid equipment leaks which are potential sources of airborne radioactivity in the reactor building are collected in the reactor building equipment drain system. The drain system is maintained at a negative pressure with respect to the balance of the building to protect operating personnel from airborne leakage generated in the system sumps. All exhaust air drawn from the reactor building equipment drain system is filtered by absolute particulate and charcoal filters.

The particulate and charcoal filters minimize the release of contaminated particulates and iodine; and

f.

The primary containment purge system reduces airborne radioactivity within the drywell to acceptable levels prior to entry of working personnel. The level of radioactivity released to the environment during normal primary containment purge will not exceed the requirements of 10 CFR Part 20 Appendix B. When airborne radiation levels in the primary containment are too high to allow direct purging to the atmosphere through the reactor building exhaust, purge air at a reduced flow rate is passed through the SGTS prior to exhaust. In this latter mode, airborne iodine and particulates are removed from the purge exhaust air prior to release; The air cleaning systems which utilize special filtration equipment to limit airborne radioactive contaminants are

a.

Standby gas treatment system (see Section 6.5),

b.

Control room emergency filtration system (see Sections 9.4 and 6.4),

c.

Reactor building sump vent exhaust filter system (see Section 9.4), and

d.

Radwaste building exhaust filtration system (see Section 9.4).

In addition, small local absolute particulate filters are used to locally filter the effluent from sample sink hoods and chemical hoods. These small filter units are all described in Section 9.4.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-15 The SGTS filter units discussed in Section 6.5.1 and the control room emergency filter units discussed in Sections 6.4 and 9.4.1 are the only engineered safety feature (ESF) filtration systems. A detailed evaluation of these units with respect to Regulatory Guide 1.52 is given in Section 1.8. The balance of the filtration systems meet the intent of Regulatory Guide 1.52 with the following deviations:

a.

Reactor building sump vent filter units These units are composed of demisters, an electric heater, a medium efficiency prefilter, an absolute particulate (HEPA) filter and tray type adsorber filters composed of 2-in. deep charcoal beds in a sheet metal housing. These units do not have absolute particulate filters downstream of the adsorber section as described in Regulatory Guide 1.52.

The 5-ft spacing between filter frames, as required in Regulatory Guide 1.52, is not provided in these units. These units are of low capacity (1000 cfm), and personnel access into the units for servicing and testing is not possible. Access panels are provided for servicing and removal of all unit components. Sufficient space is provided between elements to permit removal of any element without disturbing any other element.

b.

Radwaste building exhaust filter units These three units are composed of medium efficiency prefilters, absolute particulate (HEPA) filters and centrifugal fans in a sheet metal housing.

Unit capacity is 42,000 cfm, which is in excess of the 30,000 cfm maximum capacity recommended in Regulatory Guide 1.52. These units are composed of a 5 filter high by 8 filter wide array. Permanent service platforms are provided on both sides of the HEPA filters at an intermediate height for operating personnel during filter testing and service.

Absolute particulate (HEPA) filter and charcoal adsorber filters will be tested periodically to ensure continued filter efficiency as discussed in Sections 6.5.1.4, 9.4.2.4, and 9.4.3.4.

Figure 12.3-23 (el. 572 ft 0 in.) shows the layout of the SGTS filter units. The concrete wall between the units serves as both a fire wall and missile barrier between the two filter trains.

Access doors, 20 in. x 50 in., are provided into each plenum section between unit elements.

Ample aisle space is provided outside the units for ease of access for personnel to perform tests and maintenance. Charcoal test canisters are provided as shown in Figure 12.3-23. There are

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-05-033 12.3-16 12 test canisters per 4-in. deep filter bed. Dioctylphthalate (DOP) and freon injection and detection ports are provided as shown.

12.3.4 IN-PLANT AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION 12.3.4.1 Criteria for Necessity and Location The objectives of the in-plant area radiation and airborne radioactivity monitoring systems are to

a.

Warn of excessive gamma radiation levels in areas where nuclear fuel is stored or handled,

b.

Provide operating personnel with a record and indication in the main control room of gamma radiation levels at selected locations within the various plant buildings,

c.

Provide information to the main control room so that decision may be made with respect to deployment of personnel in the event of a radiation accident or equipment failure,

d.

Assist in the detection of unauthorized or inadvertent movement of radioactive material within the various plant buildings,

e.

Provide local alarms at selected locations where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area,

f.

Monitor areas having a high potential for increase gamma radiation levels where personnel may be required to work,

g.

Supplement other systems including process radiation leak detection or building release detection in detecting abnormal migrations of radioactive materials from process streams,

h.

Monitor the general conditions in the reactor building following an accident, and

i.

Furnish information for making radiation surveys.

No credit is taken for the operability of the in-plant area radiation and airborne radioactivity monitors in the event of an accident. However, the probability is high that many or all of the 13 area monitors in the reactor building will be operable. These monitors have local sensors in separate physical locations within the reactor building. The wiring from the local sensors is

COLUMBIA GENERATING STATION Amendment 62 FINAL SAFETY ANALYSIS REPORT December 2013 LDCN-12-031 12.3-17 run to the main control room in cable runs that have Seismic Category I qualified supports.

The electrical supply for the area monitor indicator and control room alarms in the main control room are powered from a critical 120-V ac supply. The local audible alarms are powered from a local ac supply and its loss would not impact control room readings. The sensors are designed to maintain operability and accuracy in atmosphere up to 95% RH, and temperature to 50°C.

12.3.4.2 Description and Location

a.

Area radiation monitors Area radiation monitors consist of local detector alarm units and main control room mounted indicator trip units, alarms, and recorders. Redundant criticality monitors are located in the reactor building new fuel storage pit as recommended by Regulatory Guide 8.12. When practicable and feasible the guidance in Regulatory Guide 8.12 has been followed. Major items in Regulatory Guide 8.12 have been addressed and include

1.

Employing two detectors in the new fuel vault,

2.

Emergency planning that includes both the Emergency Plan and the Emergency Plan Implementing Procedures, evacuation routes, assembly areas, and yearly drills, and

3.

Surveillance testing of criticality alarm systems, including procedures that address methods and frequency of testing.

10 CFR 50.68(b) requires radiation monitors in storage and associated handling areas when fuel is present to detect excessive radiation levels and to alert personnel to initiate appropriate safety actions.

Other detector locations have been selected in accordance with good operating practice and from past operating experience with similar plants. Detector locations are shown in Figures 12.3-5 through 12.3-18. Annunciations are given in the main control room and locally at the sensors when radiation levels exceed a predetermined level. Point indication and recording are provided for in the main control room. Local detectors are wall-mounted approximately 7 ft off the floor. The detectors have sufficient cable length to be taken from their normal positions to floor level for insertion into calibrating chambers to verify instrument accuracy. Direct-reading dosimeters, worn by individuals in

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-18 radiologically controlled areas, provide protection in areas where area radiation monitors have not been installed.

An additional area radiation monitor is installed on the refueling bridge during bridge operation to provide personnel protection. This monitoring system provides local indication, visual, and audible alarm.

There is no area radiation monitor in the area of the radwaste building where the waste containers are filled and stored. Waste containers will normally be processed either in cask or in the shielded waste storage bay.

The location and ranges of the 31 area radiation monitors are given in Table 12.3-1. Table 12.3-2 lists the maximum background radiation levels for the area radiation monitors in the reactor building based on design basis calculation.

b.

Airborne radiation monitors Airborne radioactivity monitoring for plant personnel protection and surveillance uses fixed location, continuous particulate monitors which include continuous iodine samplers; portable continuous particulate monitors with continuous iodine samplers positioned at specific work sites; and particulate and iodine grab samples taken before and during specific jobs.

Movable local alarming continuous air monitors are placed at predetermined plant locations for personnel protection and to substantiate the quality of the plant breathing atmosphere. The monitors have local readouts (charts) and have radioiodine sampling capabilities.

The installed continuous particulate monitoring system was designed for responsive personnel protection and plant surveillance. The three installed particulate monitors measure the airborne particulate activity levels in the radwaste and reactor building ventilation exhaust and furnish recording signals to the main control room. These units draw approximately 3 cfm air sample through the particulate filter which is monitored by a shielded beta detector with an efficiency of approximately 30%. The resultant response of the system is an increase of about 350 cpm for 1 hr of sampling at a 1 x 10-10 Ci/cm3 concentration. External gamma radiation will increase the background by 70 cpm/mrem/hr.

COLUMBIA GENERATING STATION Amendment 60 FINAL SAFETY ANALYSIS REPORT December 2009 LDCN-07-050 12.3-19 The actual ability of a ventilation exhaust monitoring system to detect the airborne particulate concentration in a specific space is dependent on the following factors:

1.

Flow rate ratio (flow of air from a specific confined space/flow rate of bulk ventilation system exhaust),

2.

Particulate activity and its half-life of the bulk ventilation system exhaust

air,

3.

Radionuclide composition in the specific confined space, and

4.

The energy of the beta radiation from the radionuclide composition.

Normal plant conditions are expected to yield a bulk ventilation exhaust air concentration (primarily short-lived fission product daughters and natural activity half-life about 20 minutes) of 1-3 x 10-10 Ci/cm3. This will reach an equilibrium on the sample filter of about 500 cpm.

The MPCa for normal plant airborne contamination is expected to be greater than 6 x 10-8 Ci/cm3. At this MPCa concentration a 1-hr accumulation (one MPCa-hr) will equal 2.0 x 105 cpm. Applying a dilution factor of 270:1, the 1-hr accumulation will equal 750 cpm.

This is a worst case dilution that considers the reactor building TIP Drive room at 400 cfm in the 105,000 cfm building bulk exhaust flow. Therefore, the ventilation monitoring system will easily detect 10 MPCa-hr on all locations.

Local particulate constant air monitoring instruments and a comprehensive (particulate, noble gases, and iodine) grab air sampling program complement the plant air sampling program.

Under these conditions, corrective actions will be taken and an assessment by portable sampling system results and portable monitoring activities will establish activity levels in all occupied areas which have potential for abnormal airborne activity.

In the radwaste building, the potentially contaminated areas normally entered by people would be those corridors adjacent to radioactive liquid and gaseous waste processing systems equipment such as demineralizers, concentrators, waste storage tanks, recombiners, dryers, moisture separators, and charcoal holdup vessels. Assuming that exfiltration from any one of the process systems to a normally entered corridor was sufficient to attain MPCa levels for 137Cs in that corridor, the dilution ratio would approach a factor of 10 to 100. For the worst case (100 to 1 ratio of bulk ventilation flow rate to corridor flow rate), 137Cs at MPCa (6 x 10-8 Ci/cm3) would be detected within 1 hr on the continuous particulate monitor. If exfiltration from a process vessel cubicle was sufficient to produce MPCa levels in an adjoining corridor, it is more probable that the normal cubicle flow rate input to the bulk ventilation flow would produce a prior distinguishable countrate ramp.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-20 In the turbine building, airborne contamination is most likely to arise from nuclear steam leaks or offgas processing systems piping. Heater bay areas, steam jet air ejector rooms, turbine areas, and steam-driven feedwater pump areas have the potential for airborne contamination.

Regular radiological surveys in the turbine building are not augmented by continuous particulate monitoring of the turbine building ventilation exhaust.

Each of the continuous particulate monitors has an associated iodine sampling cartridge which is counted regularly for baseline and surveillance information. This cartridge and iodine sampled/collected with portable sampling devices will be analyzed in the plant laboratory counting facility when abnormal airborne activity levels are signaled by a continuous particulate monitoring system. At a 5 cfm air flow rate through an iodine sampling cartridge, iodine present at an occupational MPCa concentration of 9 x 10-9 Ci/cm3 would be quantitatively observable within a 1-minute sample interval. Bases for this assessment are a 10-minute count time on a 12-15% Ge(Li) detector system having an overall efficiency of about 1% when source and geometry considerations are included. The information presented for detecting one MPCa concentration for 137Cs in areas having a low ventilation flow rate can also be applied to the iodine case. One MPCa of iodine can be ascertained within a 1-hr sampling period with a dilution factor of greater than 100. The routine weekly analysis of integrated iodine samples of radwaste, reactor, and turbine building ventilation air will permit observation of small iodine inputs. When these inputs are significant, a particulate and iodine sampling program is initiated to establish the source point.

Continuous particulate monitoring is reasoned to be the most responsive personnel protection and internal plant surveillance mechanism available. In addition, all tasks with potential for generating airborne contamination will be performed only when authorized by a radiation work permit (RWP).

The RWP assesses the radiological hazards, establishes additional monitoring and sampling requirements and, if necessary, specifies required engineering control and/or respiratory protection.

During outages, the above airborne monitoring system will be augmented by additional iodine sampling (continuous and grab) on the refueling floor since airborne iodine concentrations are known to become significant at this time.

12.3.4.3 Specification for Area Radiation Monitors The area radiation monitoring system is shown as a function block diagram in Figure 12.3-24.

Each channel consists of a sensor and a converter unit, a combined indicator and trip unit, a shared power supply, and a shared multipoint recorder. All channels also have a local meter and visual alarm auxiliary unit mounted near the sensor.

COLUMBIA GENERATING STATION Amendment 63 FINAL SAFETY ANALYSIS REPORT December 2015 LDCN-10-013 12.3-21 Each monitor has an upscale trip that indicates high radiation and a downscale trip that indicates instrument failure. These trips actuate alarms but cause no control function. The trip circuits are set so that a loss of power annunciates in the control room.

The type of detector used is a Geiger-Mueller tube responsive to gamma radiation over an energy range of 80 KeV to 7 MeV. Detector ranges are given in Table 12.3-1.

The calibrating frequency is once every 18 months using standard sources with National Institute of Standards and Technology (NIST) traceability. This ensures accuracies of (+) or

(-) 20% over the detection interval.

An internal trip test circuit, which is adjustable over the full range of the trip circuit, is provided. The test signal is fed into the indicator and trip-unit input so that a meter reading is provided in addition to a real trip. High-range radiation alarm trip circuits for high level and criticality monitors are of the latching type and must be manually reset at the front of the control room panel. The trip circuits for all other area radiation monitors are of the nonlatching type.

12.3.4.4 Specification for Airborne Radiation Monitors The airborne particulate monitors contain scintillation detectors with count ratemeters. Means for remote recording are provided for in the main control room. Sample collectors consist of shielded, fixed particulate, filter-type air collectors. The calibration frequency will occur at least annually and after major maintenance. Instrument response checks will be made at least monthly. Monitors will be calibrated using standard radioactive sources in the same geometry as the location of the particulate filter or by collection and analysis of mixtures present at known air flow rates. Particulate monitors are provided in the reactor and radwaste buildings.

The monitors are located so as to monitor the exhaust air from that building prior to any filtration. In addition, charcoal sampling cartridges are installed in each monitor for laboratory analysis of iodine.

Each of the three channels of the airborne radioactivity monitors has an independent local visual and audible alarm. High radioactivity or equipment failure will generate an alarm signal. No automatic system functions are performed by the alarm signals.

12.3.4.5 Annunciators and Alarms The annunciators are of the window type with pushbutton switches for sound acknowledgment and light reset functions. There are no automatic system actions performed by the area radiation monitors. The annunciator alarm windows are located in the main control room.

COLUMBIA GENERATING STATION Amendment 59 FINAL SAFETY ANALYSIS REPORT December 2007 LDCN-06-000 12.3-22 Area monitors have local/remote alarms that sound on exceeding their high radiation alarm settings and remote alarms to indicate instrument failure (see Figure 12.3-24). Monitors located in the reactor building near the fuel pool and in the new fuel areas have individual high radiation alarm windows. The remainder of the area monitor units in the reactor building have a common annunciator window for high radioactivity. Area monitors in the turbine building and the radwaste building each have a common building high radioactivity alarm window. All the area monitors have one common alarm window for instrument failure.

The two area monitors that are used as criticality detectors are located in the new fuel vault.

These monitors have a range of 10+2-10+6 mrem/hr and alarm on either of a fully adjustable upscale or downscale trip point. The alarm setpoint and bases are given in the Licensee Controlled Specifications.

12.3.4.6 Power Sources, Indicating and Recording Devices The area radiation monitor power supply units, indicating devices (except local alarms), and the recorders are all located on panels in the control room and receive power from the 120-V ac instrument power supply bus. The local audio alarms are supplied from a local 120-V ac instrument distribution panel. The recorder is a multipoint, strip chart type, which compiles a permanent record of inputs from the area radiation monitors.

12.

3.5 REFERENCES

12.3-1 Jaeger, R. G. et al., Engineering Compendium on Radiation Shielding, Volume 1, Shielding Fundamentals and Methods.

12.3-2 Rockwell, T., Reactor Shielding Design Manuals, 1st Edition, D. Van Nostrand Co., Inc., New York, 1956.

12.3-3 Goldstein, H., Fundamental Aspects of Reactor Shielding, Addison-Wesley Publishing Co., Inc., Reading, 1959.

12.3-4 Blizard, E. P., Reactor Handbook, Vol. III, Part B, Shielding.

12.3-5 Aalto, E. et al., A Users Manual for the NRN Shield Design, AE-145, June 1964.

12.3-6 Hughes, D. J. and Schwartz, R. B., Neutron Cross Sections, BNL 325, 2nd Edition, 1958.

Hughes, D. J., Magurno, B. A. and Brussel, M. K., Neutron Cross Sections, BNL 325, wnd. ed., Supplement 1, 1960.

COLUMBIA GENERATING STATION Amendment 58 FINAL SAFETY ANALYSIS REPORT December 2005 12.3-23 Stehn, John R. et al., Neutron Cross Sections, BNL 325, 2nd. Edition, Supplement 2, 1964.

12.3-7 Malenfant, Richard, QAD-A Series of Point Kernel General Purpose Shielding Program, U.S. Department of Commerce, LA-3573, Springfield, VA, 1966.

12.3-8 Walker, R. L., and Grotenhuis, M., A Summary of Shielding Constants for Concrete, ANL-6443, November 1961.

12.3-9 Chilton, A. B. and Huddleston, C. M., A Semiemperical Formula for Differential Dose Albedo for Gamma Rays on Concrete, Nuclear Science and Engineering, 17, 419-424, 1963.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 Table 12.3-1 Area Monitors Station Location Building Level (ft)

Range (mrem/hr)

LDCN-98-117 12.3-25 1

Reactor building fuel pool area 606 102-106 2

Reactor building fuel pool area 606 1-104 3

Reactor building new fuel area 606 102-106 3A Reactor building new fuel area 2 606 102-106 4

Reactor building control rod hyd equipment area E 522 1-104 5

Reactor building control rod hyd equipment area W 522 1-104 6

Reactor building equipment access area S 572 1-104 7

Reactor building neutron monitor system drive mechanical area 501 1-104 8

Reactor building SGTS filters area 572 1-104 9

Reactor building northwest RHR pump room 422 1-104 10 Reactor building southwest RHR pump room 422 1-104 11 Reactor building northeast RHR pump room 422 1-104 12 Reactor building RCIC pump room 422 1-104 13 Reactor building HPCS pump room 422 1-104 14 Turbine building turbine front standard 501 1-104 15 Turbine building entrance 441 1-104 16 Turbine building reactor feed pump area 1A 441 1-104 17 Turbine building reactor feed pump area 1B 441 1-104 18 Turbine building condensate pump area 441 1-104

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 Table 12.3-1 Area Monitors (Continued)

Station Location Building Level (ft)

Range (mrem/hr)

LDCN-98-117 12.3-26 19 Main control room 501 1-104 20 Radwaste building valve room E 467 1-104 21 Radwaste building valve room W 467 1-104 22 Radwaste building sample room 487 1-104 23 Reactor building CRD pump room 10 422 1-104 24 Reactor building equipment access area (W) 471 1-104 25 Radwaste building hot machine shop 487 1-104 26 Radwaste building contaminated tool room 467 1-104 27 Radwaste building waste surge tank area 437 1-104 28 Radwaste building tank corridor area north 437 1-104 29 Radwaste building tank corridor area south 437 1-104 30 Radwaste building radwaste control room 467 1-104 32 Reactor building NE entrance 471 10-1-104 33 Reactor building NW entrance 501 10-1-104 34 Reactor building eastside 606 10-1-104 35a Reactor building refueling bridge 606 0.1-2000 a Item 35 is installed at its dedicated location on the refueling bridge prior to bridge operation.

Alarm settings for all of the above monitors will be selected to provide indication of any abnormal increase in radiation levels while minimizing false alarms.

COLUMBIA GENERATING STATION Amendment 54 FINAL SAFETY ANALYSIS REPORT April 2000 LDCN-98-117 12.3-27 Table 12.3-2 Maximum Design Basis Background Radiation Level for Area Monitors ARM Building Level (ft)

Maximum Design Basis Background Level (mrem/hr)

ARM-RE-1 606 100 ARM-RE-2 606 100 ARM-RE-3 606 100 ARM-RE-3A 606 100 ARM-RE-4 522 3000 ARM-RE-5 522 500 ARM-RE-6 572 100 ARM-RE-7 501 100 ARM-RE-8 572 100 ARM-RE-9 422 4000 ARM-RE-10 422 4000 ARM-RE-11 422 100 ARM-RE-12 422 3000 ARM-RE-13 422 100 ARM-RE-23 422 3000 ARM-RE-24 471 100 ARM-RE-32 471 100 ARM-RE-33 501 100 ARM-RE-34 606 100