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Amendment 61 to Final Safety Analysis Report, Chapter 12, Radiation Protection
	ML12011A166
Person / Time
Site:	Columbia
Issue date:	12/14/2011
From:	; Energy Northwest
To:	; Office of Nuclear Reactor Regulation
References
	GO2-11-201
	Download: ML12011A166 (136)
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C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 12 RADIATION PROTECTION

TABLE OF CONTENTS Section Page LDC N-0 2-0 0 0 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 CONS IDERATIONS......................................................12.1-1 12.1.2 DESIGN CO NSIDERATIONS......................................................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 CONTAINE D 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 Recirc ulation System Sources................................................12.2-2 12.2.1.2.2.3 Reactor Wa ter 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 H eat Removal System Sources....................................12.2-3 12.2.1.2.2.6 Fuel Pool Coo ling 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 Special Nuclear Materials..........................................................12.2-6 12.2.2 AIRBORNE RADIOACTI VE MATERIAL SOURCES........................12.2-6 12.2.2.1 General................................................................................12.2-6 12.2.2.2 Model for Computing the Ai rborne 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 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 Sp ent 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 DE SIGN FEATURES..................................................12.3-1 12.3.1.1 Radiati on 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 Redu ce Crud Buildup..................................

12.3-6 12.3.1.3.3 Field Rou ting of Piping..........................................................12.

3-7 12.3.1.3.4 Desi gn Features That Reduce O ccupational 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 Require d Materials................................................................12.3-10 12.3.2 SHIELDING............................................................................

12.3-10 12.3.2.1 General................................................................................12.

3-10 12.3.2.2 Met hods 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 Turbin e Building..................................................................12.3-13 12.3.2.3.4 Radwas te Building................................................................12.3-13 12.3.3 VENTILATION........................................................................

12.3-13 12.3.4 IN-PLANT AREA RADIA TION AND AIRBORNE RADIOACTIVITY MONITORING INSTRU MENTATION...........................................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 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 12 RADIATION PROTECTION

TABLE OF CONTENTS (Continued)

Section Page LDCN-05-056 12-iii 12.3.4.3 Specification fo r 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, I ndicating 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 Fact ors, Dose Rates, and Es timated Personnel Exposures.....12.4-2 12.4.3 INHALATION EXPOSURES.......................................................12.4-4 12.4.4 SITE BOUND ARY 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 R easonably Achievable Procedures.............................12.5-10 12.5.3.3 Radiological Survey Procedures...................................................12.

5-12 12.5.3.4 Procedures for Radi oactive 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 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 Chapter 12 RADIATION PROTECTION

LIST OF TABLES Number Page LDCN-03-040 12-iv 12.2-1 Basic Reactor Data fo r 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 Ener gy 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 Pi ping 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 th e Turbine Generato r Building..................12.2-27

12.2-12 Special Sources With Strength Greater Than 100 Millicuries...............12.2-28 12.2-13 List of Radioactive Pipi ng 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 12.2-15 Airborne Radionuclide Concen tration in Conden sate Pump Area (el. 441 ft. 0 in. turbine generator building)....................................

12.2-31 C OLUMBIA G ENERATING S TATION Amendment 56 F INAL S AFETY A NALYSIS R EPORT December 2001 Chapter 12 RADIATION PROTECTION

LIST OF TABLES (Continued)

Number Title Page LDC N-0 1-00A 12-v 12.2-16 Airborne Radionuclide Concen tration in Secondary Containment from a Main Steam Relief Valve Blowdown...................................

12.2-32 12.2-17 Airborne Radi onuclide Concentration in Liquid Radwaste Handling Area........................................................................12.

2-33 12.3-1 Area Monitors........................................................................

12.3-25 12.3-2 Maximum Design Basis Bac kground Radiati on 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 Pr ocessing......................12.4-13

12.4-6 Occupational Dose Estimat es 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 Informati on Reported by Commercial Boiling Water Reactors.......................................................................12.

4-17 12.5-1 Health Physics In strumentation...................................................12.

5-21 C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 Chapter 12 RADIATION PROTECTION

LIST OF FIGURES Number Title LDC N-0 2-0 0 0 12-vi 12.3-1 Access Control System Ground Floor Plan 12.3-2 Access Control System Mezzanine Floor Plan 12.3-3 Access Control System Operating Floor Plan

12.3-4 Access Control System R&C Building at El. 487 ft 0 in. and 525 ft 0 in.

12.3-5 Radiation Zones - Turbine Generator Building

12.3-6 Radiation Zones -

Ground Floor Plan - Turbine Generator Building

12.3-7 Radiation Zones - Mezzani ne Floor Plan - Turbine Generator Building, East Side

12.3-8 Radiation Zones - Mezzani ne Floor Plan - Turbine Generator Building, West Side

12.3-9 Radiation Zones - Opera ting Floor Plan - Turbine Generator Building, East Side

12.3-10 Radiation Zones - Oper ating 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 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 Chapter 12 RADIATION PROTECTION

LIST OF FIGURES (Continued)

Number Title LDCN-05-056 12-vii 12.3-19 Arrangement of Filtration a nd Demineralization Equipment (Typical)

12.3-20 Schematic Arra ngement 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

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT 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 occupati onal and public radi ation exposures as far below regulatory dose limits as is practical while performing all activities related to the operation of Columbia Generati ng Station (CGS) and the Inde pendent Spent Fuel Storage Installation (ISFSI). This commitment is reflec ted in the Radiation Protection Program (RPP) that meets the requirements of 10 CFR 20 and provides for eff ective 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 radi ation control practices, in cluding preplanning and the proper use of appropriate equipment by qualified, well trained personnel.

The radiation protection practices are based, when practicab le 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 radiati on protection are prepared consis tent 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 ar eas described above. The following is a description of the applicable activities conducted by individuals or groups having responsib ility for radiation protection.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-042 12.1-2 a. The Plant General Manager, has the overall responsibility for approving the RPP and ALARA policy cons istent with Energy No rthwest and regulatory requirements, and for the ra diological safety of all on-site personnel. This includes the responsibility for implemen tation of the ALARA program by plant staff. The Plant General Manager has ensured consideration of exposure reduction methods by adopti on of the ALARA review practices described in Section 12.1.3. The Plant General Manager is responsible for the management of plant operational activ ities and for providing th e Radiological Services Manager the resources and support necessary to implement the RPP. This includes the responsibility for ensuri ng that the ALARA program is not adversely affected by pr oduction oriented goals;

b. The Radiological Services Manager reports to the Pl ant General Manager and is responsible for implementing the RPP w ith the exception of those radiation safety duties for which the Assistant Ra diological Services Manager/Radiation Protection Manager (RPM) is responsible. This individual provides organizational leadership and direction to the Assistant Radiological Services Manager/RPM in the mana gement of the Radiati on Protection department;

c. The Assistant Radiological Services Manager/RPM reports to the Radiological Services Manager. The Assistant Ra diological Services Manager/RPM has direct access to the Plant General Manager in all matters re lating to radiation safety, and has the responsibility and authority for ensuring th at plant activities meet applicable radiation safety regulations and RPP requirements. Specific responsibilities are pr ovided is Section 12.5.1;

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

e. The Health Physics (H P) Craft Supervisors report to the Radiological Operations Supervisor and implement the RPP through direct supervision of the plant HP Technicians. Ar eas 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, an d temporary shielding installation;

f. The Radiological Planni ng Supervisor reports to the Radiological Services Manager and is responsib le for radiation exposur e reduction. This is accomplished, in part, by providing input to the planning and coordination of plant activities. The Radiological Pla nning Supervisor ma kes recommendations C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-3 for the control/elimination of radiologi cal conditions that increase personnel exposure or the releas e of radioactivity;

g. The Radiological Support Supervisor reports to the Radi ological Services Manager and provides technical s upport to the Radiation Protection organization. This individual provide s 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) ha s been established a nd is functional. Its purpose is to serve as a review an d advisory organization to the Plant General Manager in several areas, incl uding radiological and nuclear safety.

The RPM is a member of the POC and ha s 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 pr ocedures, 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 re view described in Section 12.1.3 provides for this consideration in all plant procedures, quality aud its 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) provide s a description of this group's responsibilities and authority. By its charter, the CNSRB reviews radiological safety policies and programs and determines that these policies and pr ograms are in compliance with NRC requirements. The CNSRB has th e 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 occupati onal exposure to

personnel. Committee membership, res ponsibilities, author ities, and records are prescribed in plant procedures.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-4 The commitment to ALARA is implemented by employee training; audits, assessments, and reviews of the program; pro cedure development and reviews; enforcement of rules; and modifications to plant equipment or facilities where they will substantially reduce exposures at a reasonable cost. Management

's commitment to the ALARA po licy 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 polic y considerations.

12.1.2 DESIGN CONSIDERATIONS

To ensure that personnel occ upational radiation expos ures are ALARA, extensive consideration is given to equipment design and locations, accessibility requireme nts, and shielding requirements. Many of these desi gn objectives and considerations were estab lished prior to the issuance of Regulatory Gu ide 8.8. However, the design of th e plant substantially incorporates the recommendations provided in the regulatory guide. Design c onsiderations that ensure occupational radiation exposures to personnel during no rmal operation and anticipated operational occurrences are ALARA are the following:

a. The facility is separated into c ontrolled and uncontro lled areas based on anticipated radiation levels. The cont rolled areas of the facility are further defined by radiation zones established by pers onnel access requirements and are intended to limit radiation exposure to ALARA. The radiation zones provide guidance for the shielding design, contam ination control, a nd 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 ma jor 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 doe s not exceed 0.1 mrem/hr. In C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-5 addition, redundant pump s and cross tie piping permit the transfer of tank contents should abnormally hi gh 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 en try into the cubicle.

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

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

2. Radioactive pipes are r outed so that radiation exposure to plant personnel is minimized. The extent to which radioactive pipes ar e routed through normally accessible areas is minimized. Shielded pipe chases are utilized in normally accessible areas. Whenever possible, radioactive and nonradioactive pipes are kept sepa rate 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 shie ld wall to a low radiation area.

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

5. Where practical, loca l instrumentation readout s are routed to points outside shielding walls.

6. To minimize maintenance time a nd 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 e nhance 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 appropriat e low radiation areas.

8. Access to corridor C-125 on the 43 7 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 C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-6 in the corridor to detect abnorma l radiological conditions and warn personnel if radiation leve ls 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 shie lding calculations. Shielding design is conservative since the design basis radia tion 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 la byrinths are used to eliminate radiation streaming through access openings in the cubicles.

d. Auxiliary systems that may become contaminated ar e designed with provisions for flushing or remote chemical cleani ng prior to maintenance. This is accomplished by the following:

1. Providing connections for the purpose of backflushing, 2. Providing water connecti ons to tanks containing spargers to allow for water injection to un cake 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 fa cilitated by the following:

1. Filter access doors, which are size d to enhance the ease of performing maintenance, and

2. Providing for periodic inservice test ing of the equipment and filters.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-7 f. Spread of contamination is minimi zed in the event spillage occurs by the following:

1. Drains are provided in areas wher e equipment with large volumes of radioactive fluid is loca ted. 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, thos e that carry radioac tive 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 cen terline (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 mi nimized 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 a nd piping elbows are used.

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

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 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 w ith 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 cha nged with the aid of tools to allow remote handling.

r. Operating experience from other BW R plants is periodically reviewed. Problems are reviewed and the plant desi gn is checked to ensure that similar problems will not occur.

s. Design changes are review ed by Radiation Protection.

12.1.3 OPERATIONAL CONSIDERATIONS

12.1.3.1 Procedures and Methods of Operation

A positive means of ensuring that occupational a nd public radiation e xposures are ALARA has been incorporated into the Plant Procedures Manual (PPM) a nd Site Wide Procedure (SWP) preparation program. Procedures are formally reviewed for ALARA consid erations as part of the approval process. The guidance provided by Regulatory Guide 8.8 is considered during this review. Additional information regarding ALARA consid erations in procedures is provided in Section 12.5.3.2.

In addition to the above process, the Radia tion Work Permit (RWP) process is used for individual tasks to identify radiological conditions and controls. This provides a means of prescribing precautionary measures, protectiv e equipment, and other exposure reduction methods in each situation. I ndividual exposures, as determin ed by dosimeters, are recorded and provide a means of determining exposures for various tasks and work groups. This information is used for prepla nning work, identifying sources, dete rmining radiation levels and otherwise evaluating exposure problems.

Administrative controls ensure that occupational and public radiation expos ures are as low as reasonably achievable. These are supplemented by the Radiation Protection staff through a C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT 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 determin e sources and trends of e xposure and for investigation to determine causes of nor mal and unusual exposure;

b. Plant procedures are formally revi ewed by Radiation Protection for ALARA considerations when required;

c. Plant modifications th at 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 ra diological concerns. For these tasks, a RWP is completed which describes the proposed actions. The RWP describes the conditions necessary such as survey re quirements, 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 take n, 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 AL ARA, and several design changes have been made since issuance of the PSAR and Regulatory Guide 8.8. These change s or additions were implemented as a result of review by both the architect-engineer a nd Energy Northwest personnel and include the following:

a. Revised offgas system va lve design to prevent releas e of radioactive gases to building atmosphere,

b. Relocation of the counting room for lower background leve ls and adequate shielding,

c. Revised effluent monitoring capabilities to provide for more efficient monitoring, C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-10 d. Increased capability for in-plant conti nuous airborne radioactivity monitoring with remote readout and recording features,

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

f. Inclusion of supplied air stations thr oughout the plant for ef ficient respiratory protection,

g. Space and services provisi ons made for a decontamina tion facility and hot shop to reduce contact maintenance exposur es and airborne radioactivity,

h. Revised penetra tion access design at sacrificia l 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 c onsumed in repacking,

k. Replaced hydraulic snubbers with m echanical snubbers to reduce maintenance requirements,

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

m. Made provisions for future connec tions to increase re actor 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 exposur e 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; C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.1-11 b. Respiratory protection procedures incorporate proven practices from other nuclear facilities;

c. Typical procedures on survey meth ods, personnel m onitoring, personnel dosimetry, and process/effluent radiologi cal monitoring have been observed in the implementation stage at several operati ng reactor facilities. In addition, the procedures used at operating BWR plants have been reviewed. All of these were evaluated and applied in th e procedure generating process;

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

1. The issuance, requirements, c onditions, and controls of RWPs,

2. The review process of plant pro cedures for ALARA considerations, and

3. Methods for minimizi ng personnel exposures duri ng RPV head removal, drywell entry, and conduct during emergencies.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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 co rrosion products) cont ained 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, an d fission product gamma rays. During shutdown, the reactor core radiation s ources 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.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 1 7 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 corr ected by a multigroup removal source.

Table 12.2-3 lists the gamma ray energy spectrum fo r 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 po stoperation 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 sh ielding requirements within the reactor a nd 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 re actor 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 th r o ugh 12.3-18. 12.2.1.2.2.2 Recirculation System Sources. The coolant activation products, principally 16 N, are the dominant sources of radiation in the RRC system during normal operation. The 16 N 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 containm ent of the reactor building, from approximately el. 501 ft to el. 540 ft.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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. Shie lding design is based on the 16 N source, which is more than adequate to shie ld against the fission pr oduct shutdown source.

12.2.1.2.2.3 Reactor Wa ter 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 16 N. The 16 N source strength (given in activity per unit length of line) in the RWCU sy stem 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 excha nger. Returning from the radwaste building, the 16 N source strength ranges from 3.08 x 10

-10 Ci/cm to negligible (less than 10

-14 Ci/cm). The 16 N source strengths in the regenerative and nonregenerative heat exchangers are

a. Tube side of the regenera tive heat exchanger: 2.69 x 10

-6 Ci/cm 3 , b. Tube side of nonregenera tive heat exchanger: 6.24 x 10

-8 Ci/cm 3 , and c. Shell side of the regenera tive heat exchanger: 1.70 x 10

-14 Ci/cm 3.

These heat exchangers are treat ed 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 exch angers are located at el. 548 ft 0 in.

During shutdown, the fission products are the do minant 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 shut down 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 16 N activity (given in activity/unit length of line) in the inlet line is 2.96 x 10-4 Ci/cm and in the outle t 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 16 N. 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 an d corrosion products.

Table 12.2-5 lists the gamma ray energy

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-4 spectrum of the radionuclides in the RHR pump s , pipes, and heat exchangers 4 hr after shutdown. These sources a r e ba s e d on the RHR system operating in the normal reactor shutdown cooling mode. The fission and corros i on p r oduct isotope concentrations used are listed in Tables 11.1-2 through 11.1-4. The RHR heat exchangers are l o cated approxi m a tely 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 sys t em are tre a ted as equivalent line sources.

The heat exchangers are treated as cy lindrical source

s. 12.2.1.2.2.6 Fuel Pool Cool i ng and Cleanup and System Sources. The primary sources of radioactivity in the s p ent fuel assemblies, which are sto r ed in the fuel pool, are the fission products.

Table 12.2-6 lis t s the gamma ray energy spectr u m for the spent f u el sources for shutdown t i me of 2 days. The refueling area is located at el. 606 ft 10.5 in. in the reactor building.

These source terms are ca l cula t e d using the Pe r k ins and King data (Re f erence 12.2-2). The shielding calculations are done using t h e QAD point kernel code (Reference 12.2-3). The following assumptions are used in det e rmining the shielding requirements:

a. After radioactivity has r eached equilibrium in the fuel asse m b lies, it i s assumed that the reactor is shut down and the who l e core is moved, within 2 days, into the spent fuel pool;

b. The whole core and ano t her one-fou r th of a core from the la st refueling are located by the north wa l l of the spent fuel pool to g i ve the most conservative dose rate on the outside of the wall. Less water exi s ts b e tween the assembly racks and the north wall than between the assembly racks and any other side of the pool. The assemblies from past r e fuelings do not add to the shielding requirements becau s e t h ey have decayed f o r more than 1 year, they are shielded by pool water, and they p r ov i de self shielding; and

c. The water, racks, spent fuel, and other constituents t h at are located within the array of spent fuel ass e mblies are homogenized for t h e pu r pose of determining the required values of the li n ear at t e nuation coeffic i ent s. The minimum depth of water needed to adequately shield t h e ref u e l ing area from the spent fuel assemblies is calculated. It is found that the elevated fuel a ssembly 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 sour ce geometry for the purpose of computing the water depth.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-5 The source strength used to dete rmine the shielding requirements for the dryer-separator pool is based on a contact dos e 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 Re actor Feedwater Systems Sources. In these systems, the dominant sources of radioactivity are the coolant activation pr oducts, 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 tube s, and the MSR plena are listed on Table 12.2-7. The first and second stage reheater tube s are approximated by rectangula r parallelepipeds. The plena are divided into an array of rectangular pa rallelepipeds and cylinders, depending on their physical arrangement.

The 16 N source strength in the main condenser is 6.0 x 10

-8 Ci/cm 3. 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 ma in condenser is treated as either a truncated cone or infinite slab depending on the view angle and dist ance from the condense r to the dose point.

Since most of the 16 N exists as a noncondensab le 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 16 N source strength of feedwater heater 6 listed in Table 12.2-9 , governs the shielding requirements on the mezzanine floor of the turbin e building (el. 471 ft 0 in.). The feedwater heater is treated as an array of rectangular parallel pipes and cylinde rs for input into QAD.

Table 12.2-10 lists the 16 N 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 th is system. The offgas equipm ent is located at el. 441 ft 0 in. of the turbine building.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 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. Th e average source strength per unit length of cable is 3.27 x 10 4 Ci/cm. This is calcu lated using an exposure time of 864 sec. The average ra dioactivity emitted per unit lengt h 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 Re sulting 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 Special Nuclear Materials A list of all special nuclear materials with an activity greater than 100 mCi is given in Table 12.2-12.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES

12.2.2.1 General Design features that limit the airborne radi oactivity 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 lim its 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, Tabl e 1, Column 3.

No radiation Zone I areas exist in the reactor or turbine genera tor 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 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 c oncluded that the airborne concentration in the counting room is small.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 12.2-7 See Section 12.2.2.3.5 for discussion on the contributi on of sampling a nd radiochemical analysis on airborne radioactiv ity 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 concentra tion is based on the continuous leakage of a radioactiv e fluid into a plant area. Th e removal of radionuclides from this area is through ventilation exhaust and decay. The equation which yiel ds the airborne radionuclide concentration in a plant area is:

CAqPF iq qV iisi v a ia()exp(/) 1 t (12.2-1) where: C i = concentration of radionuc lide i in a given plant area (ci/cm 3) A i = concentration of radionuc lide i in the fluid (mCi/g)

q s = rate of radionuclide leak age into an area (g/minute)

(PF)i = partition factor for radi onuclide i (dimensionless) i = decay constant for isotope i (1/minute)

V = volume of area (cm

3) q a = HVAC air flow rate out of area (cm 3/minute) t = time interval between start of leak and calculation of concentration (minute)

The equilibrium value of C i is given by CAqPF Vq i i s ii 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.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 12.2-8 12.2.2.3 Sources of Airborne Radioactivity The potential sources of airborne radioac tivity found in the pl ant are as follows:

a. Leakage from process e quipment in radioactive systems, such as valves, flanges, and pumps, b. Sumps, drains, tanks, a nd 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 releas ed from sampling, and

f. Airborne radioactivity released from the spent fuel pool wa ter 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 ra dionuclide 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 pr ocess 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 doe s 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 radio activity 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 tr ansported 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.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 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 potentia lly 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 anot her 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 signifi cantly contribute to the airborne radioactivity level in normally occupied areas. This is due to th e HVAC air path which was discussed earlier.

The third parameter is the radionuc lide 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 ta nk 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 lo w 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 th at most of th ese 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 e xplained 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 leve ls 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 wh ich may contribute to airborne radionuclide levels in normally occupied ar eas is discussed in the followi ng paragraphs. Those systems which are used only during loss-of-coolant accid ent (LOCA) conditions are not discussed. These include the high-pressure core spray (HPCS), low-pre ssure 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) le akage 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 be tween 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 demine ralizers or the condensat e 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 ft 3/minute. The

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT 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 concentra tion in the area where the condensate booster pumps and condensat e 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 lo cated between column lines K.1/L.9 a nd 14.7/15.4 at el. 452 ft 0 in. of the radwaste building. This fi lter 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 De mineralizer Vents

The equipment drain (EDR), floor drain (FDR), and miscellaneous radwaste (MWR) systems are designed to collect and pro cess 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 sour ces of airborne radionuclides for the following reasons:

a. Each of the EDR, FDR, and MWR sump s 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 in to 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 surrou nding 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 preven t radioactive gases from escaping into the areas around the location of the drains. Other drai ns do not employ loop

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 12.2-11 seals, but s i nce the ri s e r vent is connected to t h e HVAC system, air w i ll be drawn into t h e drain th r ough the ris e r vent and out to the HVAC system.

The tanks and filter demineralizer vessels that conta i n significant invento r ies of ra dionuclides are vented to the HVAC syste

m. These tanks and f ilter demineralizer vess e ls are located in Zone III or Zone IV radiation areas. Even if a n y airborne radionuclides were released from these tanks or filter demineralize r s, there would be no effect on norm a lly occupied areas due to the HVAC system desi g n feature s , 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 significa n t source of airborne radioactivity in normally occupied areas.

The reasons are as foll o ws: a. All rel i ef valves (e xcept the main s t eam safety relief valves), which relieve pressure in the turbine m a in 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 pa rt 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 th e equipment being relieved. For discharge back to the sy stem, 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 va lves 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 al l radionuclides that are present in the main steam blowdown ar e released to the pr imary containment air. The radionuclide distribution within the free volume of the primar y and secondary containm ent is assumed to be uniform. The radionuclide concentration in secondary containment becomes, in mCi/cm 3: CR qtAqR VRqVt sc bi vivsc ,i(exp()/))b sc i t-exp-( (12.2-3) where: R = primary containment leakage constant (1/minute) q b = main steam blowdown flow (g/minute)

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 12.2-12 t b = duration of blowdo wn flow (minute)

q v = ventilation flow rate out of secondary containment (cm 3/minute)

V sc = volume of secondary containment (cm

3) i = decay constant for isotope i (1/minute) t = time after blowdown event C sc,i = airborne radionuclide concentrati on of radionuclide i in the secondary containment (Ci/cm 3) A i = radionuclide concentration in blowdown fluid ( Ci/g) The value of t which yields the maximum value of C sc,i is tRqV n R qVvsc iivsc1 1// (12.2-4)

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

This results in all 18 relief va lves being actuated for a maximu m duration of 40 sec. This event results in the maximum release of ra dionuclides to primary containment (see Table 5.2-3 and Sections 15.1 and 15.2). The values of the various para meters 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) q b = 1.6 x 10 7 lb/hr = 1.2 x 10 8 g/minute (Table 5.2-3) t b = 40 sec = 0.67 minute (Table 5.2-3) q v = 9.5 x 10 4 cfm (Table 11.3-6) V sc = 3.5 X 10 6 ft 3 (Table 11.3-6) The values of A i are based on the information found in Section 11.1.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT 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 i nventory of radioactive gases will accumulate inside the reactor vessel head between the time of shutdown a nd 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 contaminati on. This is done prior to flooding the RPV cavity.

It is anticipated that RPV head and reactor internals removal w ill 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 fe atures are incorporat ed into the sample system to limit the radionuclide release. Radioactiv e 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 air borne radioactive particles to the exhaust duct. Administrative c ontrol 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.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 12.2-14 12.2.2.3.7 Effects of Solid Radwaste Handling Areas

The solid radwaste handling equipment contai ned 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 w ith air flow into surrounding normally unoccupied areas. The only source of ai rborne 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 valv e 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 s upplied 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 se parate ventilate d supply and exhaust.

The radwaste control room and the precoat rooms do not house co mponents containing radioactive material.

Although not normally occupied, the possibility exists that entry in to 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 de scribed 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.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 12.2-15 12.2-2 Perkins, J. F. a nd King, R. W., Energy Release from Decay of Fission Products, Nuclear Science and Engineeri ng, Vol. 3, 1958 and Perkins, J. F., U.S. Army Missile Comma nd 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 Nucl ear Power Station, Re fueling/Maintenance Outage, GE NEDO-20761, Class 1, Fall 1974.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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/c m 3) Volume Fraction U O 2 10.4 0.254 Zr 6.4 0.140 H 2O 1.0 0.274 Void 0 0.332 Average water density between core and vessel below the core 0.74 g/cm 3 Average water-steam density above core In the plenum region 0.23 g/c m 3 Above the plenum (homogenized) 0.6 g/c m 3 Average steam density 0.036 g/c m 3 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-18 Table 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary Energy Range (MeV) Neutron Flux (Neutrons/c m 2 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

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-23 Table 12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater Component Radioactivity Concentration (Ci/cm 3) 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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/c m 3 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

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-25 Table 12.2-9 Nitrogen-16 Source Strength in Feedwater Heater 6 Radionuclide Concentration (Ci/c m 3) Feedwater Heater Steam Water 6 4.93E-7 8.40E-6 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-27 Table 12.2-11 Offgas System Sources in the Turbine Generator Building Component 16 N Source Strength

( Ci/c m 3) 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 separato r a 2.7E1 a The preheater, recombi n er, offgas condenser, and water se parator are located in the same room.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-28 Table 12.2-12 Special Sources With Strength Greater Than 100 Millicuries Isotope Identification Form Quality (mCi) Use/Location 241 AmBe 2-81-020 Solid 15,200 Calibration (EOF) 241 AmBe 2-82-050 Solid 2930 Neutron source (plant) 241 AmBe 2-86-073 Solid 1,020 WNP-1 source (plant) 137 Cs 2-79-012 Solid 186 Eberline calibration (EOF) 137 Cs 2-79-013 Solid 4,870 Eberline calibration (EOF) 137 Cs 2-79-016 Solid 1,970 Eberline calibration (EOF) 137 Cs 2-79-017 Solid 101,000 Eberline calibration (EOF) 137 Cs 2-79-033 Solid 790 Panoramic shepard cal (EOF) 137 Cs 2-83-097 Solid 7,050 ARM calibration (plant) 137Cs 2-84-058 Solid 2,220 Shepard series 28 cal (EOF) 137 Cs 2-88-002 Solid 200 Victoreen model 878-W cal (plant) 85 Kr 2-84-019 Gas 200 Gas source-chemistry (plant) 238 PuBe 2-84-047 Solid 12,800 WNP-3 startup source (plant) 238 PuBe 2-84-048 Solid 12,900 WNP-3 startup source (plant) 55 Fe 2-94-020 Liquid 186 Chemistry calibration source (plant) Table as of 11/6/97.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-29 Table 12.2-13 List of Radioactive Pipi ng 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 conde nsate demineralizers (COND)

Main steam (MS)

Main steam isolation valve l eakage 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)

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-03-040 12.2-30 Table 12.2-14

Airborne Radionuclide C oncentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building)

Radionuclide Airborne Concentration C i (µCi/cm 3) Derived Air Concentration (DAC) a (mCi/cm 3)

Ratio of C i 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 83 Br 3.3E-13 3E-5 1E-8 84 Br 6.3E-14 2E-5 3E-9 85 Br 1.3E-16 --- ---

a 10 CFR 20, Appendix B to 20.1001

-20.2401, Table I, Column 3.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT 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 C i ( Ci/cm 3) Derived Air Concentration (DAC) a (mCi/cm 3)

Ratio of C i to DAC 131 I 4.2E-10 2E-8 2E-2 132 I 3.8E-9 2E-6 3E-3 133 I 2.9E-9 1E-7 2E-2 134 I 7.4E-9 2E-5 4E-4 135 I 4.2E-9 7E-7 6E-3 83 Br 4.8E-10 3E-5 2E-5 84 Br 8.2E-10 2E-5 4E-5 85 Br 3.2E-10 1E-7 3E-3 a 10 CFR 20, Appendix B to 20.100 1-20.2401, Table I, Column 3.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-32 Table 12.2-16 Airborne Radionuclide Co n centration in Secondary Containment from a Main Steam Relief Valve Blowdown

Radionuclide Airborne Concentration C i ( Ci/cm 3) Derived Air Concentration (DAC

)a (mCi/c m 3)

Ratio of C i to DAC 131 I 3.0E-11 2E-8 2E-3 133 Xe 5.0E-10 1E-4 5E-6 a 10 CFR 20, Appendix B to 20.100 1-20.2401, Table I, Column 3.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.2-33 Table 12.2-17 Airborne Radionuclide Concentration in Liquid Radwaste Handling Area

Radionuclide Airborne Concentration C i ( Ci/cm 3) Derived Air Concentration (DAC) a (mCi/cm 3)

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

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 LDC N-0 1-0 0 0 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES

Columbia Generating Station plant incorporates the design objectives an d 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 12.3-1 through 12.3-18 show the general arrangement for each of the plant buildings.

In addition, these figures show the shielding arrangement, radiation z one designations for both normal operation and shutdown c onditions, controlled access area s, personnel and equipment decontamination areas, location of the health physics facilities, locat ion 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 with in the counting room is 0.1 mr em/hr during normal operation.

Plant areas, as iden tified 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 AL ARA and within the standards of 10 CFR 20.

12.3.1.1 Radiation Zone Designations The design basis criteria used fo r each zone are given below, and the plant layout including major equipment, locations, and radia tion zone designati ons are shown in Figures 12.3-5 through 12.3-18.

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

Maximum Dose Rate Zone (mrem/hr) Design Bases Criteria I 1.0 Unlimited occupancy.

II 2.5 Unlimited occupancy for pl ant personnel during the normal work week. III 100.0 Design base occupancy le ss than 1 hr per week.

Posted zones and controlled entries. IV Unlimited Positive access cont rol. Controlled entry and occupancy.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 12.3-2 Each access point to every Z one IV area may be secured by locked door or other positive control method while it is a "hi gh 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 leve ls will be conducted prior to firs t entry of Zone IV areas to determine the maximum habitation time.

12.3.1.2 Traffic Patterns Access to plant areas is shown in Figures 12.3-1 through 12.3-4. Access control and traffic patterns in the plant have been evaluated to maintain personnel radi ation exposures ALARA and to minimize the spr ead 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 in cludes the reactor building, turbine generator building, ra dwaste building, a nd diesel generator building. Normal access to these areas is through on e of two Health Physics control points located at each end of the main plant corridor.

12.3.1.3 Radiation Prot ection 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 ot her process streams containing radioactive contaminants are processed through filters and demine ralizers. 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 demineralize r is employed.

Each filter and demineralizer is located in a shie lded cubicle at el. 487 ft 0 in. of the radwaste and control building. These filt ers 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

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 12.3-3 exposure to plant personnel from adjacent sources. After remova l of the shielding plug, the filter or demineralizer can be serviced remo tely 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 cr anes provided for the pur pose of shielding plug and filter or deminerali zer vessel removal.

Each pressure precoat type filter or deminera lizer has its own suppor t 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-operate d valves can be ope rated from control panels located in Zone III radiation areas. Ma nually 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 instrume nt root valves, all pum ps and valves can be remotely operated. Instrument root valves are normally open and are not used during normal plant operation. The filter or demineralizer pr ecoat equipment and asso ciated 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 ow n support equipment. A gravity f eed 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 f ilter 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 wa ste collector filter and waste collector deminerali zer. A typical layout is shown in Figure 12.3-19.

Tanks All tanks that contain radioactive liquids a nd 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 se parator tanks-condensate backwash receiving tank, spent resin tanks waste sludge phase separator tanks, and the reacto r water clean up (RWCU) phase C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 12.3-4 separator tanks. These tanks ar e constructed of either stai nless 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 desc ribed 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 chemic al 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 ra dwaste building heating, ventilating, and air conditioning (HVAC) exhaust syst em 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 th e phase separator tank. An example of this is the waste sludge discharge pump which is shielded from the waste sludge phase separator by concre te and block wall. A shutoff valve in the suction line keeps the pump isolated when it is not in us

e. Following the discharge of a sludge batch from the phase separator tanks, the pump and its associated pi ping is automatically fl ushed 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 sh aft seal during pump operation.

Heat Exchangers Heat exchangers handling radio active fluids are designed to lim it occupational exposures. An example is the cooler condenser s whose function is to condens e moisture from the offgas process stream. The cooler conde nsers are located in a separate cell in the radwaste building, as shown in Figure 12.3-11. No personnel access is require d 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 gl ycol 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 dr ain connection. An enlarged discharge section in the loop seal protects it ag ainst siphoning. The enlarged discharge section also provides for automatic loop seal restor ation should its contents be displaced by a temporary pressure surge.

Figure 12.3-20 shows schematically the c ooler condenser loop seal arrangement.

C OLUMBIA G ENERATING S TATION Amendment 55 F INAL S AFETY A NALYSIS R EPORT May 2001 12.3-5 Recirculation Pumps The decontamination concentrator bottoms r ecirculation 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 leakag e of process liquid past the shaft seal.

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

Evaporators The decontamination solution concentrators us e 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 demi neralized water flows in a closed loop through the shell side of the evaporator and the sh ell side of the concentrator heating element. The steam is th en circulated to the condensate return tank to be pumped back to the evaporator. The steam in the concentrator heating elemen t 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 si de 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 Valv e Operating Stations Valves handling radioactive fl uids and requiring manual operation are located in a valve gallery at el. 467 ft 0 in. of th e 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 radiati on sources, such as resin traps.

In addition, the reach rod wall penetrations are grouted about the reach rod as sembly, 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 locate d in Zone III radiation areas.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 re duce occupational exposure ar e discussed in Section 12.2.2.3.5. Ventilation Filters and Filter Trains Filters that are installed as pa rt of the HVAC units in the Co lumbia Generating Station plant are located in an accessible area. Selected filter units are de signed 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 o ffgas system are loca ted in the turbine-ge nerator building. These recombiners are si ngle-pass devices which do not require process control valves. They are located in a shielded cell and do not requi re personnel access during operation. Temperature and pressure in th e recombiners are remotely mon itored. The recombiners and associated piping are designed to w ithstand an internal explosion.

12.3.1.3.2 Design Features That Reduce Crud Buildup

Design features and considerations are incl uded to reduce radioac tive nickel and cobalt production and buildup. For exampl e, the primary coolant system consists mainly of austenitic stainless steel, carbon steel, and lo w alloy steel components. Nick el content of these materials is low. Nickel and cobalt contents are c ontrolled in accordance with applicable ASME material specifications. A sma ll amount of nickel base materi al (Inconel 600) is employed in the reactor vessel in ternal components. Inc onel 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, ade quate corrosion resistan ce and can be readily fabricated and welded. Altern ate low nickel materials which meet the above requirements and are suitable for long te rm reactor service are not availabl

e. 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 extensivel y self-flushing valves.

Furthermore, all of the pipe joints of 2.5 in. and above in size in the reactor wate r cleanup (RWCU) and radw aste 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. a nd 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 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 dec ontamination 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 th ese 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 dimens ioned in detail. Other piping, 2 in. in diameter and smaller, carrying radioactive fluids, are not dimensioned in de tail 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 poi nts 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, ce iling, or floor. Radioactive piping routed through lower radi ation zones is enclosed with in a shielded tunnel when warranted by high expected radiatio n 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 ex ist in the plant can be used to minimize occupational exposure during decommissioning, whether decommissioning is accomplished through mothballing, entombment, removal/dismantling, or a ny combination of the above alternatives. Such faci lities include those used for handling and for offs ite shipment of fresh fuel, spent fuel, more sources, contaminated filter elements, resins, and other radioactive wastes. The radioactively cont aminated 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 ra dioactivity, 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 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-8 by remote control and flushed.

The plant has a hot machine s hop and a hot instrument shop located in the radwaste building where contam inated equipment can be decontaminated under controlled conditions. Provisions have been made in the same building for a future decontamination facility w ith expanded features.

If decommissioning is accomplished by mothballi ng, 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 st ate of protective storag e." In general, the facility may be left intact excep t 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, th e previously described plant design facilities are adequate to accomplish the tasks with low occupationa l radiation exposure to personnel. The additional re quirements 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 a nd 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 ac hieved if the decommissioning method adopted is that of imme diate removal/dismantling of th e 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 co uld be the installation of a strainer for the removal of large filings or other large size contaminants. The highly radioactiv e pieces can be transferred under water to the cask loading area in the spent fu el pool by methods similar to loading spent fuel. Th e 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 treatm ent 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, handli ng, and use of sealed and unsealed special nuclear source and b yproduct materials. In cluded in the program are procedures for the following:

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-039 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 ar e checked for leakage or loss of material at least semiannually, and

f. Disposal - all licensed material dispos als are in accordance w ith 10 CFR Part 20 requirements or by transfer to an au thorized 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. Th e radiochemical laboratory is equipped with a negative pressure fume hood with filtered exhaust. The hoo d 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 m ovement 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 Assistant Radiological Services Manager/Radiation Protection Manager (RPM) is re sponsible for the control and monitoring of sealed and unsealed source and byproduct materials. Th e Nuclear Material Ma nager appointed by the Engineering Manager is accountable for special nuclear materials (S NM). The Chemistry Technical Supervisor is res ponsible for the minimization of radioactive waste and the preparation, offsite ship ment, and disposal of radioactive materials and radwaste. Monitoring during handling of these materials is provided by Radia tion Protection.

Experience and qualifications of Radia tion Protection personnel ar e described in Section 13.1.

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

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 fo r reactor instrument a nd radiation monitoring equipment calibration, or as fission detectors, will be limite d to the amounts required for reactor operation or specific calibration purpos es except as noted in the facility operating license. 12.3.2 SHIELDING 12.3.2.1 General The radiation shielding desi gn is in compliance with a ll NRC regulations concerning permissible radiation doses to i ndividuals in restricted and nonr estricted 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 equipm ent protection. Shielding is supplemented with whatever administrative control procedures are necessary to ensure regulatory compliance. These control procedures include area access restrictions, o ccupancy limitations, personnel monitoring requirements, and radiation survey practices. Ot her 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 ot her sources discussed in Section 12.2 , and

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

The majority of the shielding calculations pe rformed are of the "bulk shielding" type.

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

The effects of mech anical or electrical penetrations in shield walls on ra diation exposure to personnel is minimized by locating penetrations to preclude di rect view of radiation sources through the penetration. The ef fect of penetrations in shie ld walls is also minimized by keeping penetration openings to the smallest practicable size. Penetrations are located away C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-11 from immediate areas with pe rsonnel access. When these cr iteria cannot be implemented, penetrations are offset.

Access into shielded areas is, in general, by labyrinths. Labyrinths are located to preclude direct personnel radia tion exposure. Where labyrinths ar e not practicable, shield doors are used. Knock-out walls for equipment removal are constructe d of brick arrange d 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 loca tion to another. Rem ovable shielding devices are normally used at specific locations and can be removed when necessary. The reactor vessel to fuel pool transfer passage is a lo cation where removable shielding is employed primarily for the protection of pe rsonnel working in the drywell.

Personnel evaluation of the affected drywell area may be em ployed instead of, or in conjunction with, the above mentioned shielding.

12.3.2.2 Methods of Sh ielding Calculations Standard methods are used in computing the re quired shielding thickness for a given source. These methods are desc ribed in References 12.3-1 through 12.3-4. Specific methods of calculation and the computer codes used in the shielding design ar e discussed below.

The NRN computer code (Reference 12.3-5) is used to determine th e shielding requirements for the core generated neutrons and to calculate the thermal ne utron 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 represen tation 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 React or Shielding Design Manual (Reference 12.3-2). The various sources are reduced to th eir basic geometric c onfiguration (line, di sc, cylinder, sphere, etc.) and the corresponding equations are solved to find the dose. The Ta ylor exponential form C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-12 of the buildup factor is used in these e quations. 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 lo cated. 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 calcu lated using the Chilton-Huddleston equations (Reference 12.3-9). Compensatory shielding (e.g., la byrinths, steel plate, lead wool) is used as needed, to reduce radiation streaming th rough penetrations and to protect against lo calized "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 requiremen ts 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 summarize d 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 proce ss 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 th ickness 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 prot ects station personnel in the r eactor 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 (a bove the reactor vessel) is, excep t at penetrations, less than 2.5 mrem/hr during normal reac tor operation. The reactor core is the primary source of radiation, and it is used in co mputing the above dose rate. The wall is in the sh ape 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 primar y containment vessel which has the same shape as the wall.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-13 The primary containment vessel is made of 0.75 in. minimum steel plate. The 16 N 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 co re constitute the major sources of radiation used to determine the radial dose rate. The shie lding 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 affect ed drywell area(s) and/or employing removable shielding at the fuel pool passage are two methods used for personnel protecti on in the drywell during fuel handling operations. The shieldi ng is designed such that radiation levels are no greater than 100 R/hr at contact. Portable locally alarming ra diation 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 ra diation and basis for shielding design. It is contained in the turbines, moistu re 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 conc rete 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-generato r access area at el. 501 ft 0 in. extend up to 524 ft 0 in. This minimizes the effects of di rect radiation streami ng 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 buildi ngs 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; C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-14 b. To prevent radioactivity buildup, all ve ntilation air is supplied to the reactor, turbine, and radwaste buildi ngs on a once through basis;

c. All cubicles housing equipment whic h 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 poten tial 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 su mps. All exhaust air draw n 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 a nd iodine; and

f. The primary containment purge system re duces airborne radioactivity within the drywell to acceptable levels prior to entr y 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 r eactor building exhaust, purge air at a reduced flow rate is passed through the SG TS prior to exhaust.

In this latter mode, airborne iodine and particulates are removed fr om the purge exhaust air prior to release;

The air cleaning systems which utilize specia l 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 un its are all described in Section 9.4.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 detaile d 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 ab solute particulate (HEPA) filter and tray type adsorber filters composed of 2-in. d eep charcoal beds in a sheet metal housing. These units do not have ab solute 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 th e units for servicing and testing is not possible. Access panels are provided for servicing and removal of all unit components. Su fficient space is provided between elements to permit removal of any el ement without disturbing any other element.

b. Radwaste building exhaust filter units These three units are com posed of medium efficiency prefilters, absolute particulate (HEPA) filters and centrif ugal 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 filt er wide array. Permanent service platforms are provided on both sides of the HEPA filters at an intermediate height for operati ng personnel during f ilter 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 th e 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 be tween 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 C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-033 12.3-16 12 test canisters per 4-in. deep filter bed. Diocty lphthalate (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 a nd airborne radioactivit y 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 reco rd 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 inadverten t movement of radioactive material within the various plant buildings,

e. Provide local alarms at selected locati ons where a substantial change in radiation levels might be of immediate importa nce 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 proce ss radiation leak de tection 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 radia tion 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 m onitors have local sensors in separate physical locations within the reactor building. The wiring from the local sensors is C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.3-17 run to the main control room in cable runs that have Seismic Cate gory 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 w ould 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 ne w fuel storage pit as recommended by Regulatory Guide 8.12. When practicable and feasible the guidance in Regulatory Gu ide 8.12 has been followed. Major items in Regulatory Guide 8.12 have b een 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.

The criticality monitor alarm setpoints are determined to meet 10 CFR 70.24(a)(1) using the calculational methodology of ANS I/ANS 8.3-1979, Appendix B. However, the sensitivity design guideline of 1 msec for response and alarm discussed in Section 5.3 of ANSI/ANS 8.3-1979 involves highly enriched material. Thus, the 1 msec guideline is not practicable nor applicable

at Columbia Generating Station.

Other detector locations have been sele cted 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 leve

l. Point indication and recording are provided for

in the main control room. Local detect ors are wall-mounted approximately 7 ft off the floor. The detectors have suffici ent cable length to be taken from their normal positions to floor level for inserti on into calibrating chambers to verify instrument accuracy. Direct-reading dosimeters, worn by individuals in C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-050 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 refu eling 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 wast e 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 backgr ound radiation levels for the area radiation monitors in the reactor building ba sed 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 m onitors 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 protecti on and plant surveillance. The four installed particulate monitors measure the airborne particulate activ ity levels in the radwaste, reactor, and turbine building ventilation exhaust, and furnish alarm and recording signals to the main control room. These units draw

approximately 3 cfm air sample through the particulate fi lter 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/cm 3 concentration. Extern al gamma radiation will increase the background by 70 cpm/mrem/hr.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-050 12.3-19 The actual ability of a ventilation exha ust 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 radiati on from the radionuclide composition.

Normal plant conditions are expected to yiel d a bulk ventilation exha ust air concentration (primarily short-lived fission product daughters and natural activity hal f-life about 20 minutes) of 1-3 x 10-10 Ci/cm 3. This will reach an equilibrium on th e sample filter of about 500 cpm.

The MPC a for normal plant airborne contamination is expected to be greater than 6 x 10-8 Ci/cm 3. At this MPC a concentration a 1-hr accumulation (one MPC a-hr) will equal 2.0 x 10 5 cpm. Applying a dilution factor of 270:1, the 1-hr accumulation will equal 750 cpm. This is a worst case dilution th at considers the reactor building TIP Drive room at 400 cfm in the 105,000 cfm building bulk exhaust flow. Therefore, the ventilation mon itoring system will easily detect 10 MPC a-hr on all locations.

Local particulate constant air monitoring instruments and a co mprehensive (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 asse ssment by portable sampling system results and porta ble 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 no rmally 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 charco al holdup vessels. Assuming that exfiltration from any one of the process systems to a nor mally entered corridor was su fficient to attain MPC a levels for 137 Cs in that corridor, the dilution ratio would ap proach a factor of 10 to 100. For the worst case (100 to 1 ratio of bulk ventilation flow rate to corridor flow rate), 137 Cs at MPC a (6 x 10-8 Ci/cm 3) would be detected within 1 hr on th e continuous particulate monitor. If exfiltration from a process vessel cubicle was sufficient to produce MPC a levels in an adjoining corridor, it is more probable that the normal cubicle flow rate i nput to the bulk ventilation flow would produce a prior distinguishable countrate ramp.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 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.

The areas have individual ventilation exhaust rates in excess of 5000 cfm and 137 Cs MPC a concentrations originating in these areas would give a contin uous air monitor response ramp which is distinguishable within 1 hr.

Each of the continuous particulate monitors has an as sociated 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 ac tivity levels are si gnaled by a continuous particulate monitoring system. At a 5 cfm air flow rate through an iodine sampling cartridge, iodine present at an occupational MPC a concentration of 9 x 10-9 µCi/cm 3 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 e fficiency of about 1% when source and geometry considerations are included.

The information presented for detecting one MPC a concentration for 137Cs in areas having a low ventilation flow rate can also be applied to the iodine case. One MPC a of iodine can be asce rtained 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 si gnificant, a partic ulate 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 additi on, all tasks with potential for generating airborne cont amination 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 engineeri ng control and/or respiratory protection.

During outages, the above airborne monitoring system will be augmen ted by additional iodine sampling (continuous and grab) on the refueling floor since airbor ne 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 functio n 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 reco rder. All channels also have a local meter and visual alarm auxiliary un it mounted near the sensor.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-21 Each monitor has an upscale trip that indica tes 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-Muelle r tube responsive to ga mma 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 mont hs using standard sources with National Institute of Standards and Tec hnology (NIST) traceability. This en sures accuracies of (+) or (-) 20% over the detection interval.

An internal trip test circuit, which is adjustable ove r 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 tr ip. High-range radiati on alarm trip circuits for high level and criticality monitors are of the latching type a nd must be manually rese t 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 scint illation detectors with count ratemeters. Means for remote recording and alarm annunciation 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 annuall y and after major mainte nance. Instrument response checks will be made at least monthly. Monitors w ill 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, turbine, and radwaste buildings. The monitors are located so as to monitor the exhaust air from th at building prior to any filtrat ion. In addition, charcoal sampling cartridges are instal led in each monitor for labor atory analysis of iodine.

Each of the four channels of th e airborne radioactivity monitors has an independent local visual and audible alarm and all share a common annunciator alarm window in the main control room. High radioactivity or e quipment failure will ge nerate 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 wi ndows are located in th e main control room.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-000 12.3-22 Area monitors have local/remo te 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 n ear the fuel pool and in the new fuel areas have individual high radiation alarm windows. The re mainder of the area monitor units in the reactor building have a common annunciator window for high radioactivity. Area mon itors in the turbine building and the radwaste building each have a common building high radioactiv ity 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 lo cated 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 se tpoint and bases are given in the Licensee Controlled Specifications.

12.3.4.6 Power Sources, Indi cating and Recording Devices The area radiation monitor power supply units, indicating devices (exc ept 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 reco rder 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., Engineer ing Compendium on Ra diation Shielding, Volume 1, Shielding F undamentals and Methods.

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

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

12.3-4 Blizard, E. P., Reactor Handb ook, 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.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.3-23 Stehn, John R. et al., Neutron Cros s 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 Gr otenhuis, 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.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Table 12.3-1 Area Monitors Station Location Building Level (f t) Range (mrem/hr)

LDC N-9 8-1 1 7 12.3-25 1 Reactor building fuel pool area 606 1 0 2-1 0 6 2 Reactor building fuel pool area 606 1-1 0 4 3 Reactor building new fuel area 606 10 2-1 0 6 3A Reactor building new fuel area 2 606 10 2-1 0 6 4 Reactor building control rod hyd equipment area E 522 1-1 0 4 5 Reactor building control r od hyd equipment area W 522 1-1 0 4 6 Reactor building equipment access area S 572 1-1 0 4 7 Reactor bui l d ing neutron monitor system drive mechanical area 501 1-10 4 8 Reactor building SGTS filters area 572 1-10 4 9 Reactor building north w est RHR pump room 422 1-10 4 10 Reactor building southw est RHR pump room 422 1-10 4 11 Reactor building northeast RHR pump room 422 1-1 0 4 12 Reactor building R C IC pump room 422 1-1 0 4 13 Reactor building H P CS pump room 422 1-1 0 4 14 Turbine bui l d ing tu r b ine front standard 501 1-1 0 4 15 Turbine bui l d ing entrance 441 1-1 0 4 16 Turbine bui l d ing reactor feed pump area 1A 441 1-1 0 4 17 Turbine bui l d ing reactor feed pump area 1B 441 1-1 0 4 18 Turbine bui l d ing condensate pump area 441 1-1 0 4 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 Table 12.3-1 Area Monitors (Continued)

Station Location Building Level (f t) Range (mrem/hr)

LDC N-9 8-1 1 7 12.3-26 19 Main control room 501 1-1 0 4 20 Radwaste building valve room E 467 1-1 0 4 21 Radwaste building valve room W 467 1-1 0 4 22 Radwaste building sample room 487 1-1 0 4 23 Reactor building CRD pump room 10 422 1-1 0 4 24 Reactor building equipment access area (W) 471 1-1 0 4 25 Radwaste building hot machine shop 487 1-1 0 4 26 Radwaste building con t a m inated tool room 467 1-1 0 4 27 Radwaste building waste surge tank area 437 1-1 0 4 28 Radwaste building tank corridor a r ea north 437 1-1 0 4 29 Radwaste building tank corridor a r ea south 437 1-1 0 4 30 Radwaste building radwa s te control room 467 1-1 0 4 32 Reactor building NE en t r ance 471 1 0-1-1 0 4 33 Reactor building NW entrance 501 1 0-1-1 0 4 34 Reactor building eastsi d e 606 1 0-1-1 0 4 35 a Reactor building refu e ling br i dge 606 0.1-2000 a Item 35 is installed at its dedicated location on t h e refueling bridge pr i o r to bridge operation.

Alarm setti n gs for all of the above monitors will be selected to provide indication of any abnormal increase in radiation leve ls while minimizing false alarms.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 8-1 1 7 12.3-27 Table 12.3-2 Maximum Design Basis Background Radiation Level for Area Monitors

ARM Building Level (ft)

Maximum Design Bas i s 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 Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Amendment 54 April 2000 Arrangement of Filtration and DemineralizationEquipment (Typical) 900547.52 12.3-19 Figure Form No. 960690Draw. No.Rev.Valve Gallery El. 487' - 0" El. 507' - 0" El. 467' - 0" Corridor Filter Demineralizer Instrument Rack 2'-0" MinValve Reach Rod Corridor Filter Holding PumpWall Penetrations

Not in Line of Sight

of Radiation Source Removable Shield PlugsSteel Plate (Typ.)Wall Sleeve Penetration Detail Grout Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Schematic Arrangement of the Cooler Condenser Loop Seal 900547.53 12.3-20 Figure Form No. 960690Draw. No.Rev.Notes:1. Valves and instrumentation are not shown to prevent clutter.

2. Overflow volume in the enlarged discharge section is sufficient to restore loop seal following a pressure surge.Off-gas Stream See Note 2To Sump 16'Seal(Typ.)Cooler Condenser Glycol Loop(Typ.)Moisture Separator Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000 Figure Form No. 960690Draw. No.Rev.Decontamination Concentrator Steam Supply Arrangement 900547.54 12.3-21 DistillateTank Evaporator Cond ReturnChem Waste Concentrate

Discharge Auxiliary Steam Concentrator Heating Element Columbia Generating StationFinal Safety Analysis Report Figure Not Available For Public Viewing Amendment 54 April 2000 6' - 4"Layout of the Standby Gas Treatment SystemFilter Units 900547.56 12.3-23 Figure Form No. 960690Draw. No.Rev.4.1 H.3 J K 5 5.2 6 6.8 7.7 8.3 5' - 0" Stand-by GasTreatment Unit 1A Dop and Freon Injection Port Moisture Separator Electric Heating Coils PrefilterHigh Efficiency Filter (HEPA)Carbon Test Canisters(Total 12)

Stand-by GasTreatment Unit 1B Activated Carbon Iodine AdsorberCarbon Test Canisters(Total 12)High Efficiency Filter (HEPA)Access Door 20x50 (Typical)Butterfly Damper (Typical)

Dop and Freon Detection PortExhaust Fan with AutomaticInlet Vanes (Typical)

Reactor Building El. 572' - 0" Columbia Generating StationFinal Safety Analysis Report

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-024 12.4-1 12.4 DOSE ASSESSMENT

12.4.1 DESIGN CRITERIA The criteria for the dose to plant personnel during normal opera tion and anticipat ed operational occurrences including refueling, are based on the requirements discussed in 10 CFR Part 20.

The design radiation levels during normal operation and refueling are shown in Figures 12.3-5 through 12.3-18. In areas such as the control room and offices, the maximum dose rate does not exceed 1.0 mrem/hr (Zone I radiation level). For pers onnel who work in controlled radiation areas, radiation Zone II through IV in Figures 12.3-5 through 12.3-18, administrative controls ensure that doses do not exceed the requirements of 10 CFR Part 20.

12.4.2 PERSONNEL DOSE ASSESSMENT BASED ON BWR OPERATING DATA

The italicized information is historical and was provided to support the application for an operating license.

12.4.2.1 General In general, data (Reference 12.4-1) from operating boiling water reactors (BWRs) have shown that the man-rem exposures to plant personnel are primaril y due to the corrosion product isotopes. Of the corr osion product isotopes, 60Co is believed to be the single most important radionuclide.

A review of the data from operating re actors was performe d in References 12.4-6 and 12.4-7. Based on this it was concluded that the shield ing design, which assumes the GE BWR design base source terms, was adequate to account for the additional radi oactivity that will deposit in the lines due to crud.

Chemical cleaning connections were also installed on a number of systems. A chemical cleanup can be performed to reduc e the deposits of crud and minimi ze the increase in radiation levels if needed. Section 12.3.1.3.2 addresses the design features that were incorporated to reduce the buildup of crud.

The variables that have been found to affect plant personnel exposure in clude the following:

a. The BWR plants show an increase in total personnel exposure during the first few years of operation,

b. The need to minimize plant downtime requires that inspection and repair tasks must be started immediately after plant shutdown when the dose rates from short-lived radionuclid es can be significant, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-2 c. Plant design and equipmen t layout has a significant e ffect on personnel dose.

Section 12.3.1 discusses the design features used to mi nimize plant personnel exposure, d. Training and experienc e of plant workers, e. The extent of maintenance operations required for a specific year, and

f. The extent that a utility uses non-regular or contractor personnel.

12.4.2.2 Personnel Dose fr om Operating BWR Data References 12.4-1 through 12.4-5 provide a tabulation of pers onnel exposures for operating BWRs. Table 12.4-9 tabulates the average personnel exposure for operating BWRs for the period 1969 through 19

80. References 12.4-4 and 12.4-5 provide more recent information.

The assessments of personnel ex posures summarized in Section 12.4.2.3 include this more recent information.

12.4.2.3 Occupancy Factors, Dose Ra tes, and Estimated Personnel Exposures A summary of the total estimated man-rem doses broken down by major function is given in Table 12.4-1. More detailed break downs are presented in Tables 12.4-2 through 12.4-8 for each of the seven major functions given in Table 12.4-1. These tables are based on the more recent information obtained from indus try operating experience. The data from Table 12.4-9 is given for comparison purposes only.

The results of the to tal estimated man-rem doses will be discussed with reference to six occupational groups as follows:

a. Group 1 - This group includes mainte nance personnel such as mechanical, electrical, instrument cr aftsmen, and Foreman.

There are approximately 128 people is this group.

Tables 12.4-4 and 12.4-8 provide the functional breakdown of exposures for this occupational group. As can be seen from the tables, 433 total man-rem may be expected.

Routine and special mainte nance operations which include control rod drive repairs, residual heat removal (RHR) repairs, snubber maintenance, etc., account for approximately 60% of the a verage annual personnel dose. One to two rem per year per person is projected for the station maintenance personnel for a maximum total of 256 man-rem per year. Accordingly, the remaining 175 man-rem per year would be expected to be received by non-station maintenance personnel. As discussed in Section 12.3.1 , the equipment layout C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-3 and design and shielding design are such that the exposur es are as low as is reasonably achievable (ALARA).

b. Group 2 - This group includes plant operations personnel composed of supervisors, control room staff and plant equipment operators. There are approximately 60 people in this group.

Tables 12.4-2 , 12.4-3 , 12.4-5 , and 12.4-6 show the total estimated man-rem fo r this group. As can be seen, the total is approximately 130 man-rem per year or approximately 2.2 rem per year per man. Personnel in this group will be performing routine and non-routine operation and surveillance, waste proces sing and refueling opera tions. In plant operations, personnel are expected to rece ive approximately one to two rem per year per man for a maximum total of 120 man-rem per year. The remaining 10 man-rem per year may be expected to be received by non-station personnel.

As part of this total, the supervisors and control room sta ff are expected to receive an exposure of less than 500 mrem/yr.

c. Group 3 - This group includes health physics and chemistry personnel. There are approximately 53 people in this group.

If the plant chemistry personnel spend 1% of their time collecting samp les in Zone III sampling stations. They will receive a maximum dose of 723 mrem/yr. Assuming the remainder of their time is spent in Zone I and Zone II areas, the total dose is between 1 and 2 rem per person. The health physics pers onnel conduct radiation surveys and support maintenance activities which require con tinuous and pre-job radi ation surveys.

The exposure to these health physics personnel ranges from 2 to 3 rem/yr. This is based on experience from operating pl ants. Assuming a dose of 3 rem per person per year and consider ing 35 health physics people in the group, the total is 105 man-rem per year. Since this group covers virtually all functions delineated in Tables 12.4-2 through 12.4-8, this 105 man-rem is considered to be spread out across all the functions.

d. Group 4 - This group includes engineers and technical supervisors. There are approximately 27 people in this group.

Personnel in this group will spend most of their time in Zone I areas where exposures are less than 500 mrem/yr. Table 12.4-7 indicates approximately 153 man-rem per year will be experienced for inservice inspection.

Plant technical personnel will have a supervising roll in this operation with non-stati on personnel performing the inspection operations. This supervisory roll will take the personnel into all zone levels during ISI activities and this roll is expec ted to result in exp osure from 1 to 2 rem/yr. Thus, the projected dose estimate for the 27 people in this group is 54 man-rem per year, the balance bei ng accounted for in the non-station personnel.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-4 e. Group 5 - This group includes station s upervisors such as health physics and chemistry supervisors, shift supervisors, etc. There are approximately 24 people in this group. Station personnel will supervise Group 1 and Group 2 personnel.

Their dose is approximately the same as personnel in these groups. With a projected dose estimate of 1 rem per year per person with 24 people in the group, the total dose is 24 man-rem per year.

f. Group 6 - This group includes adminis trative and manageme nt personnel.

There are approximately 31 people in this group. Personnel in this group spend their time in Zone I radiation areas. The projected dose estimates will be less than 500 mrem/yr. With 31 people in this group and a 500 mrem per man per year the total dose is 15.5 man-rem per year.

As seen from Table 12.4-1 , the total estimated man-rem expos ure is 715 man-rem. Groups 3, 5, and 6 are considered to be spread over all the functions. These groups constitute only 15%

of the total exposure in any case.

12.4.3 INHALATION EXPOSURES

Airborne radionuclide concentrations in norma lly occupied areas are, as discussed in Section 12.2.2 , well below the limits set by 10 CFR Part 20 and thus inhala tion exposures are negligible. In areas where engineering contro ls or operational procedures do not reduce the airborne radionuclide concentrations sufficiently, additional measures such as access control, limiting exposure time (DAC hours), and respirator y protection devices are used to maintain the total effective dose e quivalent (TEDE) ALARA.

12.4.4 SITE BOUNDARY DOSE

Steam handling equipment on the turbine operati ng floor can contribute to the site boundary dose in two ways: through a direct component and through an air-scattered "skyshine" component. Since the 16N bearing equipment is known, it can be shielded to reduce the direct component. The "skyshine" component reaches the site boundary as a result of those gamma rays which are directed such th at they bypass any inte rcepting shield walls and are scattered by the air to the site boundary.

The calculated results show that the skyshine dose will have its greatest effect on a dose point approximately 1950 m north of the turbine buildi ng. The skyshine dose at this point will be approximately 4 mrem/yr. This result is based on a plant capacity factor of 80%.

The main contributors to this dose and their contribution (in percent) are the south moisture-separator reheater (MSR) which contributes 60%, the north MSR which contributes 20%, the cross over lines which contribute 10%, and the turbines and feedwater heaters which contribute 10%.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-5 The dose estimate was computed from a model that represents the 16 N gamma leakage by point isotopic sources. This model uses the out put from the COHORT Code (Reference 12.4-3) which gives the airscattered dose as a func tion of distance a nd source ray angle.

The site boundary dose from liquid and gase ous effluents are disc ussed in Sections 11.2.3 and 11.3.3.

12.

4.5 REFERENCES

12.4-1 Atomic Industrial Fo rum, Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Oper ating Nuclear Power Plants, September 1974.

12.4-2 Ninth Annual Occupati onal Radiation Exposure Report, NRC, NUREG-0322, Washington, D.C., October 1977.

12.4-3 Tenth Annual Occupational Radiation Exposure Report, NRC, NUREG-0463, Washington, D.C., October 1978.

12.4-4 Occupational Radiation Exposure at Light Water Cooled Power Reactors, Annual Report 1977, NRC, NUREG-0482, Washington, D.C., April 1977.

12.4-5 Occupational Radiation Exposure at Commercial Nuclear Power Reactors, Annual Report 1979 and 1980, Volumes 1-2, NRC, NUREG-0713, December 1981.

12.4-6 NRC Seventh Annual Occupati onal Radiation Exposure Report 1974, NUREG-75/108, November 1975.

12.4-7 Atomic Industrial Fo rum, Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants, September 1974.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-7 Table 12.4-1 Summary of Occupati o nal Dose Estimates Man-rem/yr

1. Routine op e ration and s u rveillance 53 2. Nonroutine operation and surveill a nce 15 3. Routine ma i n tenance 288 4. Waste processing 15 5. Refueling 48 6. Inservice inspection 153 7. Special maintenance 145 Total 717 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-8 Table 12.4-2 Occupational Dose E s timates During Routine Operations and Surveillance Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. Walking 0.5 0.5 2 1/shift = 0.54 2a. Checking Railroad acce s s Change rooms Relay room Motor generator sets Battery room Computer ro om Switch gear r o om Air conditi o ni ng equip. Recirc. motor gen.

RBCCW heat Exchangers Emergency air comp RBCCU pumps RBCCW expansion

Tank 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 2.2 2b. Mech. vac. pumps CRD pumps

CRD hydraulic Cont. units

Refueling floor CRD filters RUCV demmo resin Tanks RNP pumps SRMP pumps

Air coolers IVST racks 10 10 10 10 10 10 10 10 10 10 10 10 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 2 2 2 2 2 2 2 2 2 2 2 2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 11 2c. CRD storage and repair SGTS HPCI turbine and pump 15 15 15 0.2 0.2 0.2 2 2 2 1/shift 1/shift 1/shift = 3.3 2d. RWCU heat exchangers RHR heat exc h angers Acid purple and turbine 50 50 50 0.1 0.1 0.1 1 1 1 1/shift 1/shift 1/shift = 5.5 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-9 Table 12.4-2 Occupational Dose E s timates During Routine Operations and Surv e illance (Continued)

Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 2e Checking (continued)

Demin precoat tank Precoat pump Waste sample pump Floor drain s a mple room Waste surge pump Equip. drain s u mp pump Waste surge pump Waste precoat pump Waste sludge d i sch. pump Waste filter aid pump Chemical waste pump Floor drain co l l. pump 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1 1 1 1 1 1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 0.1 2f. Chemical waste tank Spent resin p u mp Cond. phase d e cant pump Cond. phase sludge Discharge mix i ng pump 50 50 50 50 50 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1/shift 1/shift 1/shift 1/shift 1/shift = 1.3 2g. Floor drain d e min. Waste hopper Floor drain fi l ter 8 8 8 2 2 2 1 1 1 1/shift 1/shift 1/shift = 0.8 2h. Turbine inst. and controls Gen. C O 2 units Station air co m p. Heater feed p u mps Demin. pumps and valves MTG lubricati o n system Hatch area a b o ve demin.

tanks H 2 seal 2.1 eq u i p. Health shell pu l l space BCCW heat expansion

and pumps 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1 1 1 1

1 2 2 2 2 2 2 2 2 2

2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 1.1 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-10 Table 12.4-2 Occupational Dose E s timates During Routine Operations and Surv e illance (Continued)

Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 2i Checking (continued)

TBCCW expansion tank Ventilation equipment Demin. precoat and resin tanks Demin. precoat pumps

Sump pumps Reactor feed pump turbine

Lub. system MTG lub oil cooler Main gen. and exciter

MTG utilizer activators Stop and throttle valves

Circ. water isol. valves 5

5 5

5 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 1 1

1 1

1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 1.6 2j. Heater drain pumps Heater drain flash tanks Condense water box Reactor feed pumps and turbines 50 50 50 50 0.2 0.2 0.2 0.2 1 1

1 1 1/shift 1/shift 1/shift 1/shift = 11.0 2k. Drain coolers Feed water heaters Reheater seal tank Gland steam condenser

Main turbine

Reheater separators 15 15 15 15 15 15 0.5 0.5 0.5 0.5 0.5 0.5 1 1

1 1

1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 14.6 Total 53.04 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-11 Table 12.4-3 Occupational Dose Est i mates During Nonroutine Operations and Surveillance Activity Average Dose Rate (mrem/h r) Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. 1a.

1b.

1c.

1d. Operation of e q uipment: Traversing in-core probe system

Safety i n jection system Feedwater pumps and turb ine Instrument cal i bration 2 5

1 2 2 1

1 1 2 1

1 1 3/yr 1/month 1/week 1/day 0.02 0.06 0.05 0.73 2. Collection of r adioactive sam p les: 2a. 2b.

2c.

2d.

2e.

2f. Liquid system

Gas system

Solid system

Radiochemistry

Radwaste operation

Health physics 10 5 10 1 3

5 0.5 0.5 0.5 1 8

2 1 1

1 2

3 2 1/day 1/month 4/yr 1/day 1/week 1/day 1.83 0.03 0.01 0.73 3.75 7.30 Total 14.50 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-12 Table 12.4-4 Occupational Dose E s timates During Routine Operations and Surveillance Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. 2.

3.

4. 5. 6. 7.

8.

9.

10.

11.

12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28. Minor repairs reactor building Ventilation and air conditioning Control rod drive repair Reactor water cleanup pump Reactor water cleanup valve and heat exc h anger Residual heat removal system Safety relief valves

Main steam isol. valves

Recirc. pumps Snubber inspector and repair

Misc. turbine bldg. repairs Reactor feed pumps and turbine

Drain coolers

Steam jet air ejectors

Offgas system

MTG actuator Heater drain flash tanks

Condenser water box Annual turbine inspection Misc. radwaste pump repairs

Misc. radwaste valve repairs Filter and demin.

Centrifuge

Evaporation

Turbine instr. and control Waste solidification

Area monitors Operate laundry facility 1 0.5 15 180 110 200 80 75 200 75 2 10 2 10 2 5 2 5

3 25 10 65 50 85 2 2 20 0.5 20 20 200 35 45 27 30 100 50 100 8 40 40 40 40 40 40 20 120 40 40 30 8 50 10 40 40 40 2 1

6 3 6 8 5 6

3 5

1 2

2 2

2 1

1 1 10 2 2

3 2

3 1

2 2

3 1/week 1/week 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/day 2/yr 1/yr 2/yr 6/yr 1/yr 1/yr 1/yr 1/yr 4/yr 6/yr 1/yr 4/yr 1/yr 1/week 2/yr 2/yr 1/day 2.1 0.5 18 19 30 43 12 45 30 37.5 5.8 0.8 0.16 1.6 0.96 0.24 0.08 0.1 3.6 8.0 4.0 5.9 3.2 12.8 1.0 0.32 0.32 2.2 Total 288.2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-13 Table 12.4

-5 Occupational Dose Estim a tes During Waste Processing Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)Number of Worke r s Frequency Dose (man-rem/yr) Radwaste c o ntrol room

.5 8 1 1/shift 4.4 Sampling and filter changing 15 8 1 1/week 6.2 Panel operator insp. and testing 1 2 1 1/day 0.73 Operation of waste and

packaging equipment 2 16 2 1/week 3.3 Total 14.6 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-14 Table 12.4-6 Occupational Dose Estimates During Refueling Activity Aveg. Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/yr) 1. Opening/closing reactor pressure vessel 60 40 10 1/yr 24 2. Fuel preparation 10 24 2 1/yr 0.48 3. Refueling 10 100 15 1/yr 15 4. Fuel handling 2.5 100 4 1/yr 1.0 5. Fuel sipping 10 120 6 1/yr 7.2 Total 47.7 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-15 Table 12.4-7 Occupational Dose Estimat e s During Inservice I n spection Activity Average Dose Rate (mrem/hr)

Exposure Time (hr) Number of Workers Frequency Dose (man-rem/yr) 1. Removal/replacement of insulation 150 80 4 1/yr 48 2. Installation/removal and ladders 50 40 4 1/yr 8 3. Inspecting inside drywell 150 80 6 1/yr 72 4. Recorder data 50 80 6 1/yr 24 5. Inspecting outside drywell 5 50 2 1/yr 0.5 Total 153 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-16 Table 12.4-8 Occupational Dose Estimat e s During Special Maintenance Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/yr) Sparger replacement 800 60 5 Should not be nece s sary --- CRD replacement 260 35 5 1/yr 45.5 Turbine overhaul 5 250 20 1/5 yr 5 Servicing in-core det e ctors 15 50 3 1/yr 2.3 Offgas charcoal sys.

overhaul 20 100 2 1/20 yr 0.2 Special maintenance reactor water clearup sys.

150 100 8 1/10 yr 12 Misc. piping repairs 80 100 10 1/yr 80 Total 145.0 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000 LDC N-9 9-0 0 0 12.4-17 Table 12.4-9 Summary of Annual Information Reported by Commercial Boiling Water Reactors 1969 - 1980 Year Numb e r of Reactor s Included Annual Collective Doses (man-rems)

No. of Workers With Measurable Dose Gross MW-Yrs Electric Generated Average Dose Per Worker (rems) Average Collective Dose Per Reactor (man-rems)

Average No.

Person n el With Measurable Doses Per Reactor Average Man-rems Per MW-y r Average MW-y rs Generated Per Reactor Average Rated Capacity (MWe) Net 1969 3 (2) 586 (3 00) 290 a 192 1.03 a 195 145 a 3.1 64 112 1970 6 (4) 764 (5 10) 1,321 a 912 0.39 a 127 330 a 0.8 152 267 1971 7 (5) 1,784 (1 , 06 9) 1,873 a 1,038 0.57 a 255 375 a 1.4 187 339 1972 10 (7) 2,858 (2 , 13 0) 2,258 a 3,058 0.94 a 286 323 a 0.9 306 434 1973 12 4,564 5,340 3,394 0.85 380 445 1.3 283 459 1974 14 7,095 8,769 4,059 0.81 507 626 1.7 290 513 1975 18 12,611 14,607 5,789 0.86 701 812 2.2 321 611 1976 23 12,626 17,859 8,586 0.71 549 776 1.5 373 647 1977 23 19,042 21,388 9,098 0.89 828 930 2.1 396 645 1978 25 15,096 20,278 11,774 0.74 604 811 1.3 471 669 1979 25 18,322 25,245 11,671 0.73 733 1,010 1.6 467 669 1980 26 29,530 34,094 10,868 0.87 1,136 1,311 2.7 418 664 a During the y ears 1 9 69 t h rough 19 7 2, a ll plants rep o rted collecti v e doses b u t a few did not s ubmit the n u mber of per s onnel that re ceived measurable doses. The number of reac t ors that did report doses a nd nu m ber of work e r s is given in parentheses in the second c o l u m n. The collective doses shown in parentheses in the third column, as w e ll as the noted numb e rs in the r e maining columns, are all based on the data submitted by the number of reactors shown in parentheses. This c o rrection, and others, changed so m e of the values from those a ppearing in earlier NU R EG documen ts.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-002 12.5-1 12.5 RADIATION PROTECTION PROGRAM

12.5.1 ORGANIZATION

Radiation Protection, under the di rection of the Radiological Se rvices Manager, implements the Radiation Protection Program (RPP).

Health Physics (HP) is audited for compliance to regulations a nd to ensure that occupational and public radiation exposures are as low as is reasonabl y achievable (ALARA). Regulatory Guide 1.8 and ANSI 18.1-1971 have been followed in the selec tion of HP personnel. Energy Northwest pre-employment practices include screening to determine that plant employees are trustworthy, fit, and qua lified to perform their duties safely. The experience and qualifications of the personnel responsible for the RPP and for handling and monitoring radioactive materials including special nuclear source and byproduct materials, ar e described in Sections 12.5 and 13.1. Also, Section 13.1 describes the minimum qualificatio n requirements for specific plant personnel, using the criteria outlined in Regulatory Guide 1.8 and ANSI 18.1-1971.

The Plant General Manager reports to the Vice President, Operations and has the overall responsibility for the RPP. Th e Plant General Mana ger is responsible for ensuring that personnel, facilities, and other resources required to implement the RPP are avai lable. This includes ensuring that the authority to implemen t an effective RPP is delegated through the management structure, ensuri ng the program receives the active support of all Energy Northwest personnel, and ensu ring production goals, maintenance activities, and work schedules do not adversely affect the ability to provide proper ra diological controls. In turn, all plant personnel share the responsibility for en suring personal radiolog ical safety and are required to follow the rules and procedures established for ra diological safety. Specific responsibilities regarding ALARA are described in Section 12.1.

The Operations Manager reports to the Plant General Manage r and has the responsibility for ensuring the independence of the RPP from pl ant operational pressures. The Operations Manager provides the Radiological Services Manager the support necessary for the effective implementation of the RPP.

The Radiological Services Manage r reports to the Plant Genera l Manager and is responsible for the implementation of the RPP with the exception of those ra diation safety duties for which the Radiation Protection Manager (RPM) is res ponsible. The RPM has direct access to the Plant General Manager in all matters relati ng to radiation safety. The RPM meets the qualifications defined in Regulatory Guide 1.8, and provides the experience and expertise necessary to implement the RPP. The RPM is an assigne d duty and not a defined position in the organization. RPM duties and responsibilities may be assigned to any of the Radiation Protection management or supervisory positions desc ribed in this Chapter.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-10-002,11-016 12.5-2 The Radiological Services Manage r supports the safe, reliable, and economic operations of the plant within applicable laws, st andards, codes, regulations, and Energy Nort hwest policies.

The Radiological Planning Supervis or reports to the Radiological Services Manage r and directs the activities and monitors the performance of the Radiological Planning Group.

This group is responsible for performing AL ARA reviews and evaluations to support HP Operations.

The Radiological Support Supervis or reports to the Radiological Services Manage r and directs the activities and monitors th e performance of the Radiologi cal Support Group. This group provides technical support for all aspects of the RPP.

The Radiological Operations Supervisor reports to the Radiological Services Manager and directs the activities and monitors the performance of the Health Physics Craft Supervisors and Rad Material Control/

Rad Waste Supervisor.

The Health Physics Craft Superv isors report to the Radiological Operations Supervisor and direct the activities and monitor the performance of HP Tech nicians. Health Physics Craft Supervisors are responsible for ensuring conditions that have th e potential for causing exposure to radiation are identified, posted, and controlled.

The Rad-Material Control and Rad-Waste Supervisor reports to the Radiological Operations Supervisor and is responsible for providing immediate supervisi on, leadership and technical support to the laborers in the areas of equi pment and area decontam ination, radioactive material control and inventory and anti-contamination laundry; pr ocess, package and transport of radioactive waste materi al, including mixed waste.

Each individual who has unescorted access to Colu mbia Generating Station restricted areas is responsible for ensuring personnel radiation safety. This in cludes strict compliance with radiation protection re quirements, procedures , and good radiological work practices. In addition, individuals with escort duties are responsible for the radiological safety of visitors.

12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES

This section describes the e quipment, instrumentation, and facilities available for implementation of the RPP and the criteria used for selection of the instrumentation and equipment. The guidance pr ovided by Regulatory Guides 8.

3, 8.4, 8.6, and 8.28 has generally been followed with exceptions noted as follows:

a. Regulatory Guide 8.3, "Film Badge Performance Criteria" will be followed if film badges are used in the plant progra m; however, other dosimeter types, such as optically stimulated luminescent (OSL) dosimeters, are used as the Dosimeter of Legal Record (DLR) for complian ce with 10 CFR 20.150 1, 20.1502, and 20.2206; C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 12.5-3 b. Regulatory Guide 8.4 is implemented fo r the selection of direct-reading pocket dosimeters as defined in S ection 2 of ANSI N13.5-197 2 except for C.2.b, which states, "The calibration/response te st result should not exceed +/-10% ofan exposure from a source traceable to the National Bureau of Standards."

This is accepted on the minus side, but is considered excessively stringent on the positive side. Since the error on the positive side results in exposure conservatism to the worker, +20% is a mo re reasonable limit for rejection of a pocket dosimeter. Vendor literature will be accepted as documentation that performance standards specified in Regul atory Guide 8.4 are met. Continued use of direct-reading pocket dosimeters w ill be based on their ability to perform acceptably under test conditions for temperature and humidity described in approved Health Physics Inst ructions as follows: +/-2% drift per 24 hr at -10°C and any percent humidity; +2% drift per 24 hr at 50°C and 95% humidity; and

+20% and -10% of 80% of calibrated full scale;

c. Regulatory Guide 8.6, "Standard Test Procedure for Geig er-Mueller Counters," will be used as applicable. This guide references ANSI N42.3-1969 (ANSI/IEEE Standard 309-1970) for twelve different tests to Geiger-Mueller counters. Energy Northwest will develop tests and procedures to ensure that Geiger-Mueller tube characteristics are appropriate for planned (or intended) applications. These tests may incorporate plateau characteristics, dead time, efficiency, and operating environment;

d. The majority of direct-reading dosimet ers at Columbia Generating Station are electronic dosimeters with audible-alar m capabilities. A program for their appropriate use requires that conditions under whic h they may not perform adequately be discussed, as well as describing performance specifications which are met.

Electronic dosimeters will not be used to circumvent the initial meter survey required prior to work in an area. Ra diation Protection will assign electronic dosimeters only when the working envir onment is suitable for their use.

Individuals required to wear an electronic dosimeter will be provided appropriate instructions either in training or at the time of issue to minimize the risk of improper use. Regulatory Guid e 8.28 endorses, with one exception, the performance specifications indicated in ANSI N 13.27 "Performance Specifications for Pocket - Sized Alarming Dosimeters/Ratemeters."

National Institute of Standards and Technology.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 12.5-4 Since credit is taken for the audibl e-alarm capabilities of the electronic dosimeter, the Energy Northwest program for use of audible-alarm dosimeters complies with Regulatory Guide 8.

28 with the following exceptions:

1. Requirement: Section C.2.c of the Regulatory Guide requires a source check of audible-alarm dosimeters each day before use.

Energy Northwest position: Electro nic dosimeters equipped with an automatic electronic test to ensure de tector function are not subject to the requirement for a source ch eck each day before use.

2. Requirement: Section C.2.b(1) of the Regulatory Guide specifies that alarm dosimeters should not be used when the alarm may not be heard, such as (a) in a high noise envir onment, (b) when the user has a pronounced hearing loss, (c) when the user is wearing mufflers over the ears, or (d) when the sound from the dosimeter would be muffled by heavy clothing worn over the dosimeter.

Energy Northwest position: Electro nic dosimeters will be allowed for use when their alarm may not be h eard except when th ey are used to fulfill the alarming dosimeter f unction described in Technical Specification 5.7. When used in accordance with the Technical Specification, alte rnative methods of warning are required when the audible alarm may not be heard.

Alternative methods include, but are not limited to the following: (1) vibrator, (2) ear phone, (3) flashing

light clearly visible to the worker.

The program outlines the performance requirements for electronic dosimeters and details the exception to the AN SI N13.27 criteria, while ensu ring that reliable electronic dosimetry is used to facilitate expos ure control and the ALARA concept.

12.5.2.1 Criteria for Selection

a. Radiation and contamination survey instrumentation: This equipment was selected to cover the wide range requi rements extending from picocurie quantity measurements in the laboratory to th e thousand R/hour rang es necessary for emergency dose rate determinations. The laboratory instrumentation was chosen to provide capability for the quantitative and qualita tive analyses required to identify and measure the radionuclides encountered in a power reactor. The portable instrumentation includes low level detection capabilities for alpha, beta, and gamma contamina tion and wide rang es of dose rate measuring instruments for beta, gamma, a nd neutron radiation.

The criteria for quantity selection were to provide adequate available counting time for anticipated demand in the la boratory and sufficient porta ble instruments to cover C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-5 normal operational and emerge ncy requirements in all areas of the Columbia Generating Station facility;

b. Airborne radioactivity monitoring: The basic criteria for selection of this equipment were to provide a means fo r determining radioactive airborne effluents released from the plant, and to effectively monitor airborne radioactivity levels within the plant e nvirons. Provisions have been made for continuing response monitoring of noble gases discharged from gaseous release points from the reactor, radwaste, and turbine building, an d for continuous sampling of radioiodines and particulates at these same locations. Internal plant air monitoring instrumentation is used within these buildings with readout locally and in the control room;

c. Area radiation monitoring: This system was designed to provide continuous surveillance of radiation levels throughout the plant with local alarm at predetermined levels, local indication, and control room annunciation and recording. Functions of the system include warning of excessive gamma radiation levels in fuel st orage and handling areas, de tection of unauthorized or inadvertent movement of radioactive materials in the plant, local alarms to warn personnel in an area of a s ubstantial increase in radi ation levels, provision for supervisory information in the control room so that correct decisions may be made in the event of a radiation incide nt, backup to other systems for detection of abnormal migrations of radioactive materials in or from the process streams, and providing a permanent record of gamma radiation levels at selected locations within the various plant buildings; and

d. Personnel monitoring: Pe rsonnel dosimetry devices we re chosen to provide a record of exposure receive d by occupationally exposed individuals at the site who are likely to receive, under normal or accidental conditions, exposures greater than 10% of applicable 10 CFR 20 limits.

Personnel dosimeter badges (DLRs) cont aining an OSL dosimeter or other acceptable dosimetry provide the primar y legal record of exposure received by personnel. Each person requiring monitoring for record is assigned a badge, which is recorded with the wearer's identification.

Results from the badge and the period of exposure are recorded on a document kept as a legal Energy Northwest record. Badges used will be capable of recording exposure over a range of at least 40 mrem to greater than 1000 rem.

Persons being monitored may be required to wear other dosimetry assigned by the Radiation Protection staff, such as direct reading dosi meters, integrating dose meters, extremity ba dges, and finger rings.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 12.5-6 12.5.2.2 Facilities Radiation Protection facilities at Columbia Generating Stati o n include the following:

a. Personnel decontamination showers a nd sinks are located in the radwaste building (487 ft leve l). Te mporary change areas are s e t up as nece s s ary in areas of the plant to local i ze and prevent the spread of contamination while performing maintenance activ i ties. Small inventories of protective clothing are stored in the emergency reloca t ion ce n ters, operation and radwaste control rooms, and strategic locat i ons throughout the plant; b. Monitoring equipment f o r personnel [e.g., frisk e rs and installed personnel monitors (IPMs)] and tools

/personal items [fri s kers and small article m o nitors (SAMs)] are prov i ded at the radiological acce s s control areas and various areas within the plant to survey f o r radio a ctive contamination; c. Facilities for personn e l exposure m onitoring and pr otection, which include:

1. Internal dosimetry,

2. Respiratory protection testing;

d. Medical fir s t aid faci l it i es a r e equ i pped to p r ov i de care for injuries, including those with radioactive c ontamination involved; and

e. Faci l it ies for equipment and tool d e contamination exis t in the radwaste, turbine, and reactor buildings. The locations and facil i t i es a r e

1. Radwaste building

The general decontamin a tion area is shown in F i gure 12.3-1 2 , approximate column location Q.4-13.

6 at the 467 ft 0 in. level.

Facilities include curbing, sink, monorail hoist, and drains. At the 487 ft 0 in. level, Figure 12.3-1 3 , column coo r dinates R.2-14.0, tools and small equipment can be deconta m inated in the hot machine shop.

Facilities in the hot machine shop include a b e nch space and drains.

Also, there is a personnel (male/fem a le) decontamination station at the 487 ft 0 in. level, column coord i nate s K.1-15.9. This facility contains sinks, showers, and a decontamination kit.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-056 12.5-7 2. Turbine building Figure 12.3-5 , columns H-9.5, el. 441 ft 0 in. identifie s the turbine building decontamination area. Faciliti es include a monorail, curb, sink, shower, and drains.

3. Reactor building

The head washdown area is shown in Figure 12.3-18 at column coordinates N-5.8 at th e 606 ft 10 in. level and contains a curb and drain.

The CRD room area, Figure 12.3-16 , columns M-3.4, 501 ft 0 in.

elevation contains a sink, monora il, bench, and storage vault.

4. The office of the RPM is located in the service building.

Health Physics Craft Supervisors and HP Technicians are located in locations to provide for ready access by other plant work ers and an area to generate and process records.

5. A hot machine shop and a hot in strument shop are provided in the radwaste building for work on contaminated equipment under controlled conditions. A HEPA-filtered vacuum system is installed in the hot machine shop to control airborne radioactivity while working on radioactive equipment. Portable HEPA-filtered vacuum systems are also available.

6. A laboratory complex is provided in the radwaste building consisting of a sample room, hot radiochemistry laboratory, and a counting room where radioactive samples will be qualitativ ely and/or quantitatively analyzed.

f. A protective clothing storage and distribution facility inside the protected area fence, but outside the power block.

Radiation Protection facilities at the Plant Support Facility/Eme rgency Operations Facility:

The Energy Northwest Plant Support Facility (PSF) is located 0.75 mile southwest of Columbia Generating Station.

It is designed and e quipped to provide emer gency capabilities in support of Columbia Generating St ation and for support of Columb ia Generating St ation during normal plant operations and main tenance. Support facilities important to HP include:

a. Portable radiation mon itoring equipment calibration, b. Radiological training.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-056 12.5-8 The instrument calibration laboratory is located in the extreme northwest corner of the lower level. It contains an irradiation cubicle that is shielded on all sides and above by 2 ft of concrete. The cubicle entrance is protected by a labyrinth and a lockable gate. Larger open sources are stored and used in the cubicle. The shielded cubicle together with administrative controls such as procedures, radiation work permits, and surveys ensure that calibration laboratory operation will not result in radiation areas in surrounding spaces.

Calibrations are performed in accordance with a pproved procedures and are traceable, either directly or indirectly, to the National Institut e of Standards and Tec hnology (NIST). Available sources are listed in Table 12.2-12. 12.5.2.3 Equipment Radiation Protection equipment, other than instrumentation, is described in the following:

a. Protective clothing and accessories are provided for personnel required to work in contaminated areas. Clothing requirements for a particular task or area are prescribed by Radiation Protection based on the actual or potential conditions. Available clothing includes, but is not limited to:

1. Coveralls and laboratory coats, 2. Gloves - rubber and/or cotton, 3. Head covers, 4. Foot protection, and

5. Plastic suits - with or without supplied air.

b. Respiratory protection equipment is provided for personnel when it is not practicable to apply proce ss controls or other engin eering controls to control airborne radioactive contamination. The decision to use respiratory protection equipment is based on maintaining the TEDE ALARA. The respiratory protection program is conducted per the requirements of 10 CFR 20.1701, 1702, 1703, and 1704.

Exposure is limited to deri ved air concentrations (DAC) and annual limit on intake (ALI) values sp ecified in Appendix B, Table 1 of 10 CFR 20. Allowance is made for use of respiratory protective equipment, as prescribed in 20.1703, in limiting an individual's intake of airborne radioactive materials. Among the types of equipment used are:

1. Full face air purifying respirators, 2. Airline supplied full face masks (pressure demand regulated), and

3. Self contained breathi ng apparatus (pressure demand regulated).

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-024 12.5-9 c. Air sampling equipment, in addition to the continuous air monitors, includes high and low volume portable air sample rs, low volume consta nt air samplers, and air samplers with a se lf-contained power source.

Collection media (filters) employed are capable of collecting par ticulate and radioiodine samples; and

d. Emergency equipment and supplies are maintained in lockers at strategic locations within the plant. These lock ers are to be used for a rapid initial response and are not intended to provide the resources for a long term recovery operation. Equipment is stored for field team use at the EOF. Four-wheel drive vehicles, automobiles, and survey kits are avai lable for use by the field team. Locations and types of emergency equipment are listed in the Emergency Plan. Other than emergency supplies, the primar y storage areas for radiation protection equipment are the two Radiation Protection control areas located in the service and radwaste buildings. Temporary storage fa cilities are set up in localized areas as required.

12.5.2.4 Instrumentation Typical plant portable radiological instrumentation is described in Table 12.5-1. All of this instrumentation is calibrated at least semiannually when in us e except the Condenser R-meter which is calibrated annually. Electronic calibra tions of instrument co mponents are performed using test equipment traceable to the NIST. Overall calib ration of radiation measuring instruments is performed using radioactive standards traceable to a recognized source in a known, reproducible geometry.

Calibration of low level radia tion detection instruments is done with a pulse generator.

12.5.3 PROCEDURES

Section 12.1.3 described a process that was incorporat ed into the preparation and revision of plant procedures which provide a positive method of ensuring Radiati on Protection input and ALARA consideration into radiation exposure related activities. Th e intent of this process is to incorporate the general guidance of appr opriate regulatory guides plus the previous experience of power reactor radiation protection work into all applicable plant procedures.

12.5.3.1 Personnel Control Procedures The Plant Procedures Manual contains the admini strative procedures for control of access to radiation areas, high radiation ar eas, and very high radiation ar eas. This includes control of time spent within these areas by all plant workers. Basica lly, the procedures limit entry to these areas to time required fo r necessary operational maintena nce and surveilla nce activities only. The primary tool used to ensure cont rol and to maintain TEDE ALARA at Columbia C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-10 Generating Station is the Radiation Work Permit (RWP). All work performed at Columbia Generating Station in ra diologically controlled areas is performed in accordance with an RWP, with the exception of specific activities identified in the impl ementing plant RWP procedure. The RWP provides current data on radiation levels within the area of interest, any restrictions on allowable work time, protective clothing and respiratory protective requirements, information on special tools or equipment needed, special radiation safety and personnel monitoring requirements, and any other special instructions or radiological hold points necessary. A section of the RWP is used to incorporate the criteria given in Regulatory Guides 8.2, 8.8, and 8.10 into each individua l task, even though it has already been included in job procedures through the system for ALARA consideration de scribed in Section 12.1.3. All RWPs require approval from Radiation Protection supervision prior to starting work. In addition, all personnel who perform activities covered by th e RWP are required to read, understand, and document their und erstanding as specified by implementing plant procedures. There are two types of RWPs:

a. Specific RWP is issued for the perfor mance of a particular task or function which falls outside the limitations imposed for General RWPs, and

b. General RWP may be issued to cove r repetitive (routine) functions in areas where radiological conditions are known and stable.

In addition to the administrative controls used at Columbia Generating Sta tion, certain physical controls are established which restrict entry to radiation ar eas, high radiation areas, and very high radiation areas. Radiati on areas are posted as required by 10 CFR 20, and high radiation areas and very high radiation ar eas are locked or otherwise controlled as specified by this same regulation and Technical Specifications.

The plant security control system complement s both the administrativ e and physical entry restraints by allowing access only to personnel with authorization to be in speci fic plant areas.

12.5.3.2 As Low As Is Reas onably Achievable Procedures The procedures and processes described belo w are in addition to those described in Sections 12.1.3 and 12.5.3.1. They have been developed to ensure that o ccupational radiation exposures are maintained ALARA. A primary goal of the Columbia Generating Station RPP is the control and reduction of individual and collective radi ation exposures. This goal is achieved through training and comprehensive job planning and reviews as follows:

a. Training - ALARA training is required by plant procedures to be included in all applicable radiation training courses.

The training applies to individuals whose duties require working with radioactive materials, entering radiation areas, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-11 directing the work of others in radiologically controlled areas, or planning work and preparing procedures for work in radiologically contro lled areas. ALARA training is commensurate with the individual's duties, responsibilities, and radiation exposure potential. Workers w ho may enter restricted areas are given specific instructions about prenatal exposure risks to the developing embryo and fetus. This instruction includes th e information provi ded by Regulatory Guide 8.13;

b. Job Planning and Reviews - Plant proce dures specify that each job, involving exposure to radiation and/or radioactive materials, receives job pl anning and/or an ALARA review. The extent of the re view is determined by an evaluation of the radiological risks involved. The preliminary job planning and review identifies the need for pr e-job briefings and/or j ob planning meetings to coordinate work efforts and to familiari ze personnel with th e work and exposure reduction techniques.

Pre-job planning may in clude the following:

1. Job history reviews, 2. Determination of radiological conditions, 3. Determination of exposure estimate, and

4. Interface with planners, schedulers, job supervisor, ALARA.

Additional ALARA reviews are performed by the Senior Site ALARA Committee. The Senior Site ALARA Co mmittee has been developed to ensure participation by a range of plant personnel and provide for an appropriate level of management invol vement and direction in ALAR A issues. The Senior Site ALARA Committee serves as a review a nd advisory organization to the Plant General Manager on occ upational radiation exposure to personnel.

Plant procedures provide require ments for committee membership, responsibilities, authority, and records of meetings and actions. The Senior Site ALARA Committee is responsible for the review of plant and departmental exposure goals and reviewi ng and assessing the effectiv eness of the radiation exposure control program and the ALARA Program. The Senior Site ALARA Committee may create Workin g Groups to provide dos e reduction methods for tasks which have a significant potential for dose reduction.

As part of pre-job revi ew, ALARA job planning m eetings are conducted when significant exposure savings or increased contamination control may result. The planning meetings may include the job supervisor, job planner, Radiation Protection supervision, HP technicians, and key workers. These meetings are used to ensure worker familiarity with procedures, work locations, RWP requirements, unusual hazards, and jo b-specific ALARA techniques to be employed. The use of mock-ups or dry runs may result from these meetings.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-12 In addition to pre-job ALARA reviews, plant procedures have provisions for work-in-progress ALARA reviews and post-job ALARA reviews. These reviews are coordinated by Radiation Protection and may include discussions with individuals who performed the wo rk, HP technicians, engineers, job supervisors, designers, or others as appropriate;

c. Remote handling tools and/or equipmen t - use of special tools/equipment for remote handling of radioactive equipment is factored into each applicable work activity; and

d. Exposure records are maintained in a manner that will allow Radiation Protection to tabulate and correlate expos ure results to iden tify problem areas with individuals or activities.

Plant procedures are evaluated to determine the need for an ALARA review. An ALARA review is required for new procedures in whic h the actions take plac e in a radiologically controlled area or involve handling radioactive material.

An ALARA review is required fo r procedure revisions which:

a. Cause entry into a radiation area, hi gh radiation area, or high-high radiation area, b. Cause opening a contaminated or potentially contaminated system, c. Significantly increase dose rates, or d. Significantly increase exposure received.

Radiation Protection requirements, prerequisites, precautions , and ALARA considerations are incorporated into these procedures during the procedure review and approval process. In addition, an RWP is issued for those activiti es having radiological implications. Special activities such as inservice inspection (ISI), outage, and refue ling are reviewed by the Senior Site ALARA Committee and Radiation Protecti on as appropriate. Special precautions, prerequisites, or requirements may be incorporated into the plant procedures based on these reviews.

12.5.3.3 Radiological Survey Procedures A radiological survey is defi ned as an evaluation of radiological c onditions and potential hazards. When appropriate, such an evalua tion includes a physical survey (e.g., direct radiation and/or cont amination surveys).

Routine surveys are conducted in various areas throughout the site to identify, monitor, and control sources of radiation and contamination. Routine surveys are performed on a frequency

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-13 based on a consideration of potential radiological hazar ds, personnel occupancy and radiological stability. Incl uded in routine surveys are the following daily checks:

Check the plant area radiation monitoring system, and

Check inservice portal, air and other continuous ly operating radiation/contamination monitors.

Abnormal changes in background and abrupt unexplained increases are investigated.

Nonroutine surveys are performe d as the need and conditions dictate. The frequency and extent of these surveys should be determined based on histori cal data, on the conditions and activities taking place in the area, and on ALARA considerations.

Surveys of normally inaccessible, unoccupied areas are performe d after each shutdown or prior to entry into these area

s. Postings and survey record sheets are updated as conditions dictate.

Instructions relating to radi ation surveys are provided in the Energy Northwest Radiation Protection Procedures.

12.5.3.4 Procedures for Radioactive Contamination Control This section describes the bases and methods used for the monitoring and control of radioactive contamination on personnel, material, and surfaces.

a. Bases: The methods used for the monitoring and control of Columbia Generating Station licensed radioactive material are based on 10 CFR 20.1101(b), 20.1501, NRC Circul ar 81-07, and industry-accepted practices. Tools, equipment, and ot her items with dete cted quantities of licensed radioactive materials will not be unconditionally rele ased. Detection levels will be based on the ALARA principle.

b. Methods: Personnel a nd materials will be surveyed in accordance with 10 CFR 20.1501. When physical surv eys are performed, they will be conducted using industry-accep ted, calibrated, detecti on instruments and with techniques that are appropriate to the level of risk. Other sections of Chapter 12 cover the selection criteria for contamination survey instruments, contamination monitoring facilities, prot ective clothing, contamination and radiation controls established through the RWP program, contamination monitoring surveys, ALARA with respect to contamination, and control of airborne radioactive material.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-14 Instructions for monitoring and control of contamination are maintained in the Radiation Protection Procedures.

12.5.3.5 Procedures for Control of Airborne Radioactivity Evaluation of airborne radioac tivity concentrations is done pr ocedurally by se veral methods.

Routine airborne surveys consist of observing the continuous air monitors located in various areas of the plant and al so the effluent monitors. These observations are supplemented by grab samples taken on a routine basis an d by laboratory analysis of selected particulate and charcoal filters used on the continuous monitors. Special airborne surveys are made with portable samplers when a continuous air monitor indicates increases in airborne radioactivity or to evaluate conditions in a specific area or on a specific job.

The portable air sampling equipment consists of both high and low volume collectors with appropriate media for collecting particulates and radioiodines. These samplers are used for both spot evaluations by collecti on of grab samples and longer te rm evaluations by use of low volume samplers to collect over the period of a specific job or activity.

Laboratory analysis is made of air samples for gro ss radioactivity and, where warranted, for specific isotope identification and quantification to determ ine and record airborne concentrations.

Selected numbers of the routine air samples collected are analyzed for specific isotope content to ensure that the DAC levels are not being approached. Special samples are taken for this purpose whenever unexplained increases occur on continuous air monitors or when gross activity levels indicate there is a potential for exceeding the value specified in 10 CFR 20, Appendix B, Table 1, Column 3 of a ny isotope present in the mixture.

Airborne radioactive iodine monitoring includes integrated sa mple collection and laboratory analysis plus portable sampling and analysis. Portable and stationary sampling encompasses iodine collection on charcoal and/or silver zeolit e cartridges of nominal di mension of 2-in. disc diameter by 1-in. thickness at calib rated flow rates. Duration of sampling is determined by the anticipated ambient concentration levels whereas a nominal sampling period in excess of 5 minutes is selected to mi nimize sampling errors. Where gross noble gas concentrations exist, the sample cartridge may be purged in the laboratory with clean filtered air to minimize noble gas interferences. The cartridge will be s ealed in a clean plastic bag and taken to the analytical laboratory counting room for analysis.

Areas are barricaded and posted as airborne radioactivity areas whenever average concentrations in that area exceed 0.3 DAC of the values specified in 10 CFR 20, Appendix B to Parts 20.1001-20.2401, Table 1, Column 3. The use of resp iratory protection equipment is evaluated when a significant potential for an airborne hazard exists, or when entering an area of unmonitored, unknown airborne contamination.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-042 12.5-15 Various methods for control and reduction of airborne activity are incorporated into Energy Northwest Radiation Protection Procedures, wh ich include proper use of the ventilation system, use of specially designed equipment to collect radioactive airborne contaminants, methods for reducing and contai ning contamination to preven t it becoming airborne, and methods for cleanup of pr imary water prior to opening this system.

The respiratory protection program is designed to meet the requirements of 10 CFR 20.1701, 1702, 1703, and 1704.

Procedures for fitting, training, maintenance, and testing of the respiratory protection equipment are included. All equipment is required to have appropriate National Institute for Occupati onal Health and Safety (NIO SH)/Mine Safety and Health Administration (MSHA) approval if available. Unless the requirements are met, the protection factors are not used. Unapproved equipment ma y be used in some instances where reduction of intake of radioactive material will result, but no protection factor is taken for its use. An example of this is use of charcoal cartridges in atmospheres where radi oiodines are present to reduce the inhalation of these materials.

12.5.3.6 Radioactive Material Control Including Special Nuclear Materials (SNM)

Columbia Generating Station has implemented a program to ensure safe radioactive material control which include:

a. Procedures and training for receiving and shipping radioactive materials in accordance with 10 CFR 20.1906,

b. Procedures and training for storing licensed materials in accordance with 10 CFR 20.1801 and 1802,

c. Procedures and training for shielding, handling, and inventor y control of sealed and unsealed radioactiv e sources and SNM,

d. Procedures and training for posting and/or labeling radioactive materials in accordance with 10 CFR 20 requirements,

e. Procedures and training for leak testing sealed radioactive sources in accordance with Technical Specifications, and

f. Procedures and training for disposal of all licensed radioactive materials in accordance with 10 CFR 20, 10 CFR 30, 10 CFR 40, 10 CFR 61, or 10 CFR 70.

g. Procedures and training for activities associated with dry storage cask loading and unloading of spent fuel and the ope ration of the Independent Spent Fuel Storage Installation for storage of sp ent fuel in accordan ce with 10 CFR 72.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-039,09-002 12.5-16 Inherent in the above mentioned procedures is direction for handling liquid standard solutions used for calibration of plant inst rumentation which include ventila tion control, shielding, waste collection, contamination control, and monitoring.

Plant procedures assign the responsibility for c ontrol and monitoring of sealed and unsealed sources and byproduct materials to the RPM. The Engineeri ng Manager is responsible for overall implementation of control of SNM. This is accomplished throug h a Nuclear Material Manager who is appointed in writing by the Engineering Manager. The Chemistry Technical Supervisor is responsible for minimization of radioactive wast e and the preparation, offsite shipment, and disposal of radioactive materials and radwaste.

Monitoring during handling of nuclear materials is provided by Radi ation Protection, as appropriate.

12.5.3.7 Personnel Dosimetry Procedures

Section 12.5.2 describes the monitoring de vices used to provide the primary legal records of exposure incurred by pers onnel and additional equipment used to backup and supplement this data. Records of radiation exposure are maintained for each individual for whom personnel monitoring is required by 10 CFR 20.1502. Reports of required monitoring are documented on NRC Form 5 or electronic media containing a ll the information required by NRC Form 5. Energy Northwest provides these individual radiation exposure records pursuant to the provisions of 10 CFR 19.13.

For monitored individuals, a determination of prior occupa tional dose is made per the requirements of 10 CFR 20.2104. This includes the dose received during the current year at Columbia Generating Station and other nuclear facilities. This exposure history is documented on NRC Form 4 or equivalent.

All individuals who are monitored for external radiation exposure are monitored for internally deposited radioactivity as follows:

a. Initial, performed prior to the individual entering any radi ologically controlled area. Monitoring may be either quantitative (whole body count) or qualitative (passive monitoring).

b. When a worker formally declares pregna ncy, a whole body c ount is performed.

c. At termination of employment at Columb ia Generating Station, if the individual has been monitored for external radiati on exposure. Monitoring may be either quantitative (whole body count) or qualitative (passive monitoring).

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-17 d. Whenever an individual causes an alarm of the passive whole body contamination monitors (portal monitors) and internal deposition is suspected, a quantitative bioassay (whole bod y count) is performed.

e. Special bioassays as de termined by the RPM.

Energy Northwest monitors all adult workers who are likely to exceed 10% of the 10 CFR 20 annual occupational radiation e xposure limits for adults. Radi ation exposure monitoring for minors (individuals less than 18 y ears of age) is required if they are likely to exceed 0.1 rem deep dose equivalent in a year. Monitoring for declared pregnant women is required if they are likely to exceed 0.1 rem deep dose equivale nt during the entire pre gnancy. Monitoring is not required for visitors who enter a restricted ar ea, since they are not lik ely to exceed 10% of the annual limit. However, confirmatory monitoring of visitors may be performed if directed by Radiation Protection s upervision. The determination of whether an individual is likely to exceed 10% of the 10 CFR 20 limit, and thus re quire monitoring, is based on a prospective evaluation. An evaluation is not required for each individual, but is based on employees with similar job functions.

For internal exposure, monitoring is required if an adult worker is likely to receive, in 1 year, an intake in excess of 10% of the applicable 10 CFR 20 annual limit. The need for internal exposure monitoring of i ndividuals is based on a prospectiv e evaluation whic h will be updated whenever there is an indication that there has been significant fuel failure.

Since 90 Sr and 3 H are not measurable by whole body counting, in-vitro bioassay (urinalysis) will be performed when the plant radiation surv eillance program indicates a potential need.

All results obtained from in-vivo and in-vitro bioa ssay will be evaluated a nd become part of the individual's record, as appropriate.

Energy Northwest complies with the adult occupational dose limits identified in 10 CFR 20.1201. An individual is allowed to exceed these 10 CFR 20 exposure limits only in exceptional situations wh ere the dose received is in accordance with the conditions of a planned special exposure, as specified in 10 CFR 20.1206. Records of planned special exposures are maintained and retained per th e requirements of 10 CFR 20.2105.

Written reports of planned special exposures are submitted to the NR C per the requirement s of 10 CFR 20.2204.

In addition to the 10 CFR 20 do se limits, Energy Northwest us es administrative exposure hold points to maintain exposures ALARA. Plant procedures allow an individual to exceed an administrative hold point but, only if a prio r approved dose exte nsion is obtained.

Procedurally, DLR badges are processed for radiation workers semiannually at a minimum but may receive interim processing if an abnormal exposure is suspected.

Pocket dosimeters and other auxiliary monitoring devices are used to maintain an estimate of an individual's dose during the interim period between processing of DLRs. The use of auxiliary monitoring C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-18 devices as a permanent record of an individual's dose is restricted to times when DLRs are lost or damaged or give a false result. When a large discrepancy exists between the two devices, it must be established that the DL R is in error before the auxiliary monitoring device result is assigned as the permanent record.

Plant supervisors are notified of their assigned worker's exposure status and are responsible for maintaining these and their own exposure to ALARA and within specified limits.

Personnel dosimeters, that require processing to determine the radiation doses, are processed and evaluated in accordance with the requirements of the National Voluntary Laboratory Accreditation Program (NVLAP).

12.5.3.8 Radiation Protection Surveillance Program

The practices incorporated into the overall structure to ensure that the RPP is maintained at a high level and upgraded to meet new requi rements and problems are the following:

a. Section 12.1.1 describes the organization structured to provide assurance that the ALARA policy is effective. It is also pointed out in this section that the plant RPP has several levels of re view from a performance standpoint;

b. Section 12.1.3 describes the process for review of plant procedures for ALARA consideration;

c. The RWP program and other records pr eviously described provide a valuable source of information and are used to determine where the occupational radiation exposures are occu rring and as a means of re view for possible methods of exposure reduction;

d. The Radiological Servic es Manager and his staff work on an individual and group basis with other plant organizati ons to determine what their principal sources of exposure are and to look fo r methods of reducin g these exposures;

e. Procedures provide for routine mainte nance, calibration, and testing of all radiation instrumentation and equipment. New equi pment will be added as necessary for replacement and to supplement that existing. Written procedures are provided for use of equipment where required;

f. Plant facilities are routinely reviewed for possible improvements from a radiation protection standpoint. Section 12.1.3 describes several changes that have been incorporated into plant design for this purpose. Ot her considerations are additional shielding where practi cable, improved ventilation control, additional equipment, and incr eased physical restriction; C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.5-19 g. The routine and special surveys previously descri bed point out levels of radioactive contamination in plant areas.

The Columbia Generating Station staff is committed to maintaining a clean plan t and considers it routine procedure to reduce levels of contamination whenever such action will not result in an increase of occupational radiation exposure to personnel;

h. One aspect that is considered important and used in implementing the RPP is the incorporation of previous reactor and pow er reactor experience in this area. Previously successful methods, procedur es, and equipment are used whenever possible; and

i. Training of all personnel w ho work in the plant in radi ation safety practices is mandatory and given a high priority by Energy Northwest and Columbia Generating Station Ma nagement. The Training Mana ger, in conjunction with the Radiological Services Manager, is responsible for development of all

training programs, including radiation safety indoctrination.

Radiation Protection assists in this training by providing instructors for some phases. The degree of training provided each plant worker is dependent on his function and degree of responsibility; however , all radiation work ers in the plant are provided training considered necessary or required for their position. The training programs provided are desi gned to meet the requirements of 10 CFR 19.12 and the guidance of Regulator y Guides 1.8, 8.8, 8.10, 8.27, and 8.29. Clarifications, ela borations, and exceptions in using the above mentioned regulatory guides are located in the Energy Northwest Procedures.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 12.5-21 Table 12.5-1 Health Physics Instrumentation Type Number Available Radiation Detected Sensitivity Range Ion chamber dose rate survey meter 9 Beta, gamma 0-5E4 mR/hr (gamma) 0-2.E4 mrad/hr (beta)

cfx 5R/hr High range ion chamber dose rate survey meter 2 Gamma 0-1.999E7 mR/hr Telescoping dose rate survey

meter 3 Gamma 0-1.0E3 R/hr Neutron do s e rate sur v ey meter 2 Neutron 0.1-5.0E3 mrem/hr Contamination survey meter with end window or pancake GM

probe 20 Beta, gamma 0-5.0E4 cpm or 0-5.0E5 cpm Contamination survey meter with

scintillation detector 2 Alpha 0-5.0E5 cpm 0 Condenser R-meter 1 Gamma 0-100 R Direct reading pocket dosimeters 300 Gamma 0-999 rem Direct reading pocket dosimeters 200 Gamma Various ranges 0-500 mR

@@ Line 2: / Line 2: @@
 | number = ML12011A166
 | issue date = 12/14/2011
-| title = Columbia, Amendment 61 to Final Safety Analysis Report, Chapter 12, Radiation Protection
+| title = Amendment 61 to Final Safety Analysis Report, Chapter 12, Radiation Protection
 | author name =
 | author affiliation = Energy Northwest
@@ Line 685: / Line 685: @@
 C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT 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 re duce occupational exposure ar e discussed in Section 12.2.2.3.5. Ventilation Filters and Filter Trains Filters that are installed as pa rt of the HVAC units in the Co lumbia Generating Station plant are located in an accessible area. Selected filter units are de signed 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.
+The location of the sampling areas within the plant is discussed in Section
+===9.3. Design===
+features of sample areas that re duce occupational exposure ar e discussed in Section 12.2.2.3.5. Ventilation Filters and Filter Trains Filters that are installed as pa rt of the HVAC units in the Co lumbia Generating Station plant are located in an accessible area. Selected filter units are de signed 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 o ffgas system are loca ted in the turbine-ge nerator building. These recombiners are si ngle-pass devices which do not require process control valves. They are located in a shielded cell and do not requi re personnel access during operation. Temperature and pressure in th e recombiners are remotely mon itored. The recombiners and associated piping are designed to w ithstand an internal explosion.

ML12011A166: Difference between revisions

Latest revision as of 18:47, 4 April 2019

Contents

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2.3 REFERENCES

3.5 REFERENCES

4.5 REFERENCES

2.3 REFERENCES

9.3. Design

3.5 REFERENCES

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ML12011A166: Difference between revisions

Latest revision as of 18:47, 4 April 2019

Text

2.3 REFERENCES

3.5 REFERENCES

4.5 REFERENCES

2.3 REFERENCES

9.3. Design

3.5 REFERENCES

4.5 REFERENCES

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