ML15355A511
| ML15355A511 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/21/2015 |
| From: | Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15356A148 | List: |
| References | |
| GO2-15-159 | |
| Download: ML15355A511 (77) | |
Text
C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Chapter 12 RADIATION PROTECTION
TABLE OF CONTENTS
Section Page LDCN-13-039 12-i 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) .............. 12.1-1 12.1.1 POLICY CONS IDERATIONS ...................................................... 12.1-1 12.1.2 DESIGN CONS IDERATIONS ...................................................... 12.1-4 12.1.3 OPERATIONAL CO NSIDERATIONS ............................................ 12.1-8 12.1.3.1 Procedures and Me thods 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 In troduction
...................................................................... 12.2-2 12.2.1.2.2.2 Recirculati on System Sources ................................................ 12.2-2 12.2.1.2.2.3 Reactor Water Cleanup System Sources .................................... 12.2-3 12.2.1.2.2.4 Reactor Core Isolation Cooling System Source ........................... 12.2-3 12.2.1.2.2.5 Residual Heat Re moval System Sources .................................... 12.2-3 12.2.1.2.2.6 Fuel Pool Cooling a nd Cleanup and System Sources ..................... 12.2-4 12.2.1.2.2.7 Main Steam and Re actor Feedwater Systems Sources
.................... 12.2-5 12.2.1.2.2.8 Offgas Sources in th e 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 Resulti ng From Crud Bu ildup .................................... 12.2-6 12.2.1.3 Radwaste Building ................................................................... 12.2-6 12.2.1.4 Byproduct, Source, and Special Nuclear Materials ............................ 12.2-6 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES ........................ 12.2-6 12.2.2.1 General ................................................................................ 12.2-6 12.2.2.2 Model for Computing the Ai rborne Radionuclide Concentration in a Plant Area .......................................................................... 12.2-7 12.2.2.3 Sources of Air borne 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 for Area Radiation Monitors......................................12.3-20 12.3.4.4 Specification for Airborne Radiation Monitors.................................12.3-21 12.3.4.5 Annuciators and Alarms............................................................12.
3-21 12.3.4.6 Power Sources, 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 12 RADIATION PROTECTION
LIST OF TABLES (Continued)
Number Title Page LDCN-14-005 12-iv 12.2-1 Basic Reactor Data for Source Computations .................................. 12.2-17 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary
............................ 12.2-18 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation ............ 12.2-19
12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown ...................................................................... 12.2-20
12.2-5 Fission Product Source in RHR Pi ping and Heat Exchangers 4 Hours After Shutdown ...................................................................... 12.2-21
12.2-6 Gamma Ray Energy Spectrum fo r Spent Fuel Sources ....................... 12.2-22
12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater ...... 12.2-23
12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell .............................................................
12.2-24 12.2-9 Nitrogen -16 Source Strength in Feedwater Heater 6 ......................... 12.2-25
12.2-10 Nitrogen -16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems ................................... 12.2-26
12.2-11 Offgas System Sources in the Turbine Generator Building .................. 12.2-27
12.2-12a Special Sources With Strength Greater Than 100 M illicuries ............... 12.2-28 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Cont rolled Area
................................................... 12.2-28a 12.2-13 List of Radioactive Piping and System Designations .......................... 12.2-29
12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building) ................................................ 12.2-30
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 12 RADIATION PROTECTION
LIST OF TABLES (Continued)
Number Title Page LDCN-14-005 12-v 12.2-15 Airborne Radionuclide Concentration in Condensa te Pump Area (el. 441 ft. 0 in. turbine generator building) .................................... 12.2-31 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 Processing
...................... 12.4-13
12.4-6 Occupational Dose Estimates During Refueling ............................... 12.4-14 12.4-7 Occupational Dose Estimates During Inservice Inspection
................... 12.4-15 12.4-8 Occupational Dose Estimates During Special Main tenance ..................
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 strumentati on ...................................................
12.5-21 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Chapter 12 RADIATION PROTECTION
LIST OF FIGURES
Number Title LDCN-12-037 12-vi 12.3-1 DELETED 12.3-2 DELETED 12.3-3 DELETED
12.3-4 DELETED
12.3-5 Radiation Zones - Turbine Generator Building
12.3-6 Radiation Zones -
Ground Floor Plan - Turbine Generator Building
12.3-7 Radiation Zones - 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-061 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 Chemistry and Radiological Safety Manager and is re sponsible for implementing the RPP. This position meets the requirements of the Radiation Protection Manager (RPM) as described in Reg Guide 1.8. This individual provides organi zational leadership and direction to the Radiati on Protection department;
- c. The Radiological Servic es Manager has direct access to the Plant General Manager in all matters rela ting to radiation safety, a nd has the responsibility and authority for ensuring that plant activ ities meet applicable radiation safety regulations and RPP requi rements. Specific res ponsibilities are provided in 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 Support Supervisor reports to the Radi ological Services Manager and is responsible for radiation exposure reduction and for providing technical support to the Radiation Pr otection organization. This is accomplished, in part, by providing input to the planning and coordination of plant activities. The Radiological Support Supervisor makes recommendations
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-028 12.1-3 for the control/elimination of radiologi cal conditions that increase personnel exposure or the release of radioactivity; and provides additional assurance that regulatory requirements and industry guidance are incorporated into the program to ensure exposures are maintained ALARA.
In addition to the organization structure that has been provided for implementation of the ALARA policy, the following groups perform reviews to further ensure exposures are ALARA. a. The Plant Operations Committee (POC) 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 including nuclear safety. A written administrative procedure describes the re sponsibilities of the POC. The RPM is a member of the POC and has direct input to this group on all radiological matters; b. The Quality organization performs audits, surveillances, and assessments of plant operations. Audits, surveillances, and assessments will include verification of compliance with plant procedures, with regulations involving nuclear safety, and with operating license provisions. Results of these audits and surveillances will be submitted to both plant and corporate management.
Since the system for ALARA 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 Ma nager on radiological safety, including occupational exposure to personnel. Committee memb ership, responsibilities, authorities, 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-022 12.1-8 n. Seal glands of pumps carrying concentrated radwaste or spent filter and demineralizer resins are flushed with condensate water during operation. After operation, the pumps are also flushed 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) and Procedure Program. Procedures are formally reviewed for ALARA considerations as part of the approval process. The guidance provided by Regulatory Guide 8.8 is considered during this review. Additional information regarding ALARA considerations in procedures is provided in Section 12.5.3.2.
In addition to the above process, the 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,
- 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),
- d. Residual heat removal (RHR),
- 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-005 12.2-6 12.2.1.2.2.9 Traveling In-Core Probe System Sources. The prim ary 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 Resulting From Crud Buildup. Fission and corrosion product deposition is complex to describe analytically. In estimating the effects of this source, operating experience from other BWR plants is used (Reference 12.2-4); Section 12.4 discusses this in greater detail.
12.2.1.3 Radwaste Building
The radiation sources present in the radwaste building are discussed in Chapter 11.
12.2.1.4 Byproduct, Source and Special Nuclear Materials
A list of all byproduct, s ource and special nuclear material in sealed sources with an activity greater than 100 mCi is given in Table 12.2-12a. Several areas outside the Reactor, Turbine, and Radwaste Buildings and within the site area have been esta blished for use and storage of radioactive material in the form of activated components, seal ed sources, and contaminated tools, equipment, and protective clothing. Those areas where the total quantity of activity could at times be greater than 100 mCi unde r normal conditions are listed in Table 12.2-12b.
12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES
12.2.2.1 General
Design features that limit the airborne 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, Table 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 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 12.2-7 counting room is located at el. 487 ft 0 in. As seen in Figure 9.4-3 , the incoming ventilation flow into this room consists of fresh air. The outgoing air flow is to the area surrounding the counting room. It is c oncluded that the airborne concentration in the counting room is small.
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 leakage 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 v iv s c ,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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-005 12.2-28 Table 12.2-12a
Special Sources With Strength Greater Than 100 Millicuries
Isotope Identification Form Quantity (mCi) Use/Location 241AmBe 2-81-020 Solid 14,770 Calibration (EOF) 241AmBe 2-82-050 Solid 2,845 Neutron source (plant) 241AmBe 2-86-073 Solid 988 WNP-1 source (plant) 137Cs 2-79-033 Solid 524 Panoramic shepard cal (EOF) 137Cs 2-83-097 Solid 4,673 ARM calibration (plant) 137Cs 2-84-058 Solid 1,474 Shepard series 28 cal (EOF) 137Cs 2-88-002 Solid 133 Victoreen model 878-W cal (plant) 137 Cs 2-93-026 Solid 909 MG calibrator (EOF) 137 Cs 2-99-028 Solid 2,078 Mini-89 calibrator (plant) 137 Cs08-132 Solid 358,600 Hopewell calibrator (EOF) 137 Cs08-133 Solid 422 Hopewell calibrator (EOF) 137 Cs13-230 Solid 12,940 ARM calibration (plant) 238 PuBe 2-84-047 Solid 11,150 WNP-3 startup source (plant) 238 PuBe 2-84-048 Solid 11,260 WNP-3 startup source (plant)
Table as of 9/9/2015.
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015
LDCN-14-005 12.2-28a Table 12.2-12b
Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area Area Location Approximate Size (sq. ft.)
Normal Contents Normal Activity (mCi) LSA Storage
Pad Outside SW side of Radwaste Bldg 1800 Low-activity dry and compacted solid wast
containers 930 Building 13 West portion of Bldg 13 2700 Contaminated protective clothing 300 Warehouse 5 NE portion of Bldg 80 at
Snake River
Warehouse
Complex 4000 Radioactive &
contaminated equipment 590 Building 167 ~0.5 miles E of Plant 6332 Radioactive &
contaminated equipment 1370 Building 167 Storage Yard
~0.5 miles E of Plant 68000 C-van storage 1620 Outage Storage Yard Fenced area W of Bldg 105 2400 Outage radioactive material & C-van storage 114 Kootenai HP Calibration Lab Kootenai (Bldg
- 34) Rms 102 &
102A 600 Calibrators/irradiators, calibration sources, radioactive HP
instruments 377030 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)
Relief valve vents radioactive (VR)
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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-037 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES
12.3.1 FACILITY DESIGN FEATURES
Columbia Generating Station plant incorporates the design objectives 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 12.3-5 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 less 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-037 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 control and traffic patter ns in the plant have been ev aluated to maintain personnel radiation exposures ALARA and to minimize the spread of contamination.
Normal entry into the plant is as follows:
- a. Personnel normally access the plant Protected Area through the security Protected Area Access Point (PAAP).
- b. The main Radiologically Controlled Area (RCA) normally 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 withstand 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 low 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 radiation zones is enclosed within 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 radioactivity, that may be introduced by the
handling of radioactively contaminated systems, as well as the number of man rems due to contact with the same systems, can be reduced by first decontaminating them. Means exist in the present design where radioactively contaminated systems can be decontaminated chemically 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 contaminated 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 mothballing, the above provisions will reduce to low
levels the occupational radiation exposure to personnel. According to Regulatory Guide 1.86, this method involves "putting the facility in a 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 could 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-13-039 12.3-9 a. Receiving and opening shipments as required by 10 CFR 20.1906,
- b. Storage of licensed materials as required by 10 CFR 20.1801 and 20.1802,
- c. Inventory and control of radioactive materials,
- d. Posting of radioactive material storage areas and tagging of source,
- e. Leak tests - sources 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 Radiological Services Manager/
Radiation Protection Manager (RPM) is responsible for the control and monitoring of seal ed and unsealed source and byproduct materials. The Nuclear Mate rial Manager appointed by the Engineering Manager is accountable for speci al nuclear materials (SNM).
The Chemistry Technical Supervisor is responsible for the minimization of radioactive waste and th e preparation, offsite shipment, and disposal of radioactive materials and radwaste.
Monitoring during handling of these materials is provided by Ra diation Protection. Experience and qualifications of Radiation Protection personnel are 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 equipment 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 radiation 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 direct radiation streaming at the site boundary.
The shielding arrangement for the major sources of radiation in the turbine building, including those discussed above, are shown in Figures 12.3-5 through 12.3-10.
12.3.2.3.4 Radwaste Building
The shielding arrangement for the major sources in the radwaste building are shown in Figures 12.3-11 through 12.3-14.
12.3.3 VENTILATION
The plant ventilation systems for the different areas of the plant are designed to meet the requirements of 10 CFR Parts 20 and 50. Gaseous wastes will be released in a controlled manner to environs such that during plant operation, onsite and offsite radioactivity levels are ALARA. The design features which limit and reduce the airborne radioactivity are as follows:
- a. In the reactor, radwaste, and turbine generator 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 which handles radioactivity contaminated material are ventilated at a minimum rate of three volume air changes per hour;
- d. All sinks and chemical laboratory work areas where radioactive samples or materials are handled are provided with exhaust hoods to protect operating personnel from airborne contaminants;
- e. All liquid equipment leaks which are 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 special filtration equipment to limit airborne radioactive contaminants are
- a. Standby gas treatment system (see Section 6.5), b. Control room emergency filtration system (see Sections 9.4 and 6.4), c. Reactor building sump vent exhaust filter system (see Section 9.4), and d. Radwaste building exhaust filtration system (see Section 9.4). In addition, small local absolute particulate filters are used to locally filter the effluent from sample sink hoods and chemical hoods.
These small filter 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 absolute particulate filters downstream of the adsorber section as described in Regulatory Guide 1.52.
The 5-ft spacing between filter frames, as required in Regulatory Guide 1.52, is not provided in these units. These units are of low capacity (1000 cfm), and personnel access into 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-031 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.
10 CFR 50.68(b) requires radiation monitors in storage and associated handling areas when fuel is present to detect excessive radiation levels and to alert personnel to initiate appr opriate safety actions.
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 sufficient 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 12.3-18 radiologically controlled areas, provide protection in areas where area radiation monitors have not been installed.
An additional area radiation monitor is installed on the 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 three installed particulate monitors measure the airborne particulate activ ity levels in the radwaste and reactor build ing ventilation exhaust and furnish recording signals to the main control room.
These units draw approximately 3 cfm air sample through the particulate filter which is monitored by a shie lded 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. External 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 12.3-20 In the turbine building, airborne contamination is most likely to arise from nuclear steam leaks or offgas processing systems piping. Heater bay areas, steam jet air ejector rooms, turbine areas, and steam-driven feedwater pump areas have the potential for airborne contamination.
Regular radiological surveys in the turbine building are not augmented by continuous particulate monitoring of the turbine building ventilation exhaust.
Each of the continuous particulate monitors has an 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 Ci/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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 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 scintillation detectors with count ratemeters. Means for remote recording are provided for in the main control room. Sample collectors consist of shielded, fixed particulate, filter-type air collectors. The cali bration frequency will occur at least annually and after major maintenance. Instrument response checks will be made at least monthly. Monitors will be calib rated 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 r eactor and radwaste buildings. The monitors are located so as to monitor the exhaust air from that building prior to any filtration. In addition, charco al sampling cartridges are installe d in each monitor for laboratory analysis of iodine.
Each of the three channels of the airborne ra dioactivity monitors ha s an independent local visual and audible alarm. Hi gh radioactivity or equipment failure will generate an alarm signal. No automatic system functions are performed by the alarm signals.
12.3.4.5 Annunciators and Alarms The annunciators are of the window type with pushbutton switches for sound acknowledgment and light reset functions. There are no automatic system actions performed by the area radiation monitors. The annunciator alarm windows are located in the main control room.
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 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Chapter 12 RADIATION PROTECTION
TABLE OF CONTENTS
Section Page LDCN-13-039 12-i 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AND RADIATION EXPOSURES TO MEMBERS OF THE PUBLIC ARE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) .............. 12.1-1 12.1.1 POLICY CONS IDERATIONS ...................................................... 12.1-1 12.1.2 DESIGN CONS IDERATIONS ...................................................... 12.1-4 12.1.3 OPERATIONAL CO NSIDERATIONS ............................................ 12.1-8 12.1.3.1 Procedures and Me thods 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 In troduction
...................................................................... 12.2-2 12.2.1.2.2.2 Recirculati on System Sources ................................................ 12.2-2 12.2.1.2.2.3 Reactor Water Cleanup System Sources .................................... 12.2-3 12.2.1.2.2.4 Reactor Core Isolation Cooling System Source ........................... 12.2-3 12.2.1.2.2.5 Residual Heat Re moval System Sources .................................... 12.2-3 12.2.1.2.2.6 Fuel Pool Cooling a nd Cleanup and System Sources ..................... 12.2-4 12.2.1.2.2.7 Main Steam and Re actor Feedwater Systems Sources
.................... 12.2-5 12.2.1.2.2.8 Offgas Sources in th e 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 Resulti ng From Crud Bu ildup .................................... 12.2-6 12.2.1.3 Radwaste Building ................................................................... 12.2-6 12.2.1.4 Byproduct, Source, and Special Nuclear Materials ............................ 12.2-6 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES ........................ 12.2-6 12.2.2.1 General ................................................................................ 12.2-6 12.2.2.2 Model for Computing the Ai rborne Radionuclide Concentration in a Plant Area .......................................................................... 12.2-7 12.2.2.3 Sources of Air borne 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 for Area Radiation Monitors......................................12.3-20 12.3.4.4 Specification for Airborne Radiation Monitors.................................12.3-21 12.3.4.5 Annuciators and Alarms............................................................12.
3-21 12.3.4.6 Power Sources, 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 12 RADIATION PROTECTION
LIST OF TABLES (Continued)
Number Title Page LDCN-14-005 12-iv 12.2-1 Basic Reactor Data for Source Computations .................................. 12.2-17 12.2-2 Neutron Flux at Reactor Core-Reflector Boundary
............................ 12.2-18 12.2-3 Reactor Core Gamma Ray Energy Spectrum During Operation ............ 12.2-19
12.2-4 Reactor Core Gamma Ray Spectrum Immediately After Shutdown ...................................................................... 12.2-20
12.2-5 Fission Product Source in RHR Pi ping and Heat Exchangers 4 Hours After Shutdown ...................................................................... 12.2-21
12.2-6 Gamma Ray Energy Spectrum fo r Spent Fuel Sources ....................... 12.2-22
12.2-7 Nitrogen-16 Source Strength in Main Steam and Reactor Feedwater ...... 12.2-23
12.2-8 Gamma Ray Energy Spectrum and Volumetric Source Strength in the Hotwell .............................................................
12.2-24 12.2-9 Nitrogen -16 Source Strength in Feedwater Heater 6 ......................... 12.2-25
12.2-10 Nitrogen -16 Source Strengths for Piping Associated With the Main Steam and Reactor Feedwater Systems ................................... 12.2-26
12.2-11 Offgas System Sources in the Turbine Generator Building .................. 12.2-27
12.2-12a Special Sources With Strength Greater Than 100 M illicuries ............... 12.2-28 12.2-12b Radioactive Material Use and Storage Areas Outside the Main Radiologically Cont rolled Area
................................................... 12.2-28a 12.2-13 List of Radioactive Piping and System Designations .......................... 12.2-29
12.2-14 Airborne Radionuclide Concentration in Control Rod Drive Pump Area (el. 422 ft. 3 in. reactor building) ................................................ 12.2-30
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 Chapter 12 RADIATION PROTECTION
LIST OF TABLES (Continued)
Number Title Page LDCN-14-005 12-v 12.2-15 Airborne Radionuclide Concentration in Condensa te Pump Area (el. 441 ft. 0 in. turbine generator building) .................................... 12.2-31 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 Processing
...................... 12.4-13
12.4-6 Occupational Dose Estimates During Refueling ............................... 12.4-14 12.4-7 Occupational Dose Estimates During Inservice Inspection
................... 12.4-15 12.4-8 Occupational Dose Estimates During Special Main tenance ..................
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 strumentati on ...................................................
12.5-21 C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 Chapter 12 RADIATION PROTECTION
LIST OF FIGURES
Number Title LDCN-12-037 12-vi 12.3-1 DELETED 12.3-2 DELETED 12.3-3 DELETED
12.3-4 DELETED
12.3-5 Radiation Zones - Turbine Generator Building
12.3-6 Radiation Zones -
Ground Floor Plan - Turbine Generator Building
12.3-7 Radiation Zones - 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-13-061 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 Chemistry and Radiological Safety Manager and is re sponsible for implementing the RPP. This position meets the requirements of the Radiation Protection Manager (RPM) as described in Reg Guide 1.8. This individual provides organi zational leadership and direction to the Radiati on Protection department;
- c. The Radiological Servic es Manager has direct access to the Plant General Manager in all matters rela ting to radiation safety, a nd has the responsibility and authority for ensuring that plant activ ities meet applicable radiation safety regulations and RPP requi rements. Specific res ponsibilities are provided in 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 Support Supervisor reports to the Radi ological Services Manager and is responsible for radiation exposure reduction and for providing technical support to the Radiation Pr otection organization. This is accomplished, in part, by providing input to the planning and coordination of plant activities. The Radiological Support Supervisor makes recommendations
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-15-028 12.1-3 for the control/elimination of radiologi cal conditions that increase personnel exposure or the release of radioactivity; and provides additional assurance that regulatory requirements and industry guidance are incorporated into the program to ensure exposures are maintained ALARA.
In addition to the organization structure that has been provided for implementation of the ALARA policy, the following groups perform reviews to further ensure exposures are ALARA. a. The Plant Operations Committee (POC) 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 including nuclear safety. A written administrative procedure describes the re sponsibilities of the POC. The RPM is a member of the POC and has direct input to this group on all radiological matters; b. The Quality organization performs audits, surveillances, and assessments of plant operations. Audits, surveillances, and assessments will include verification of compliance with plant procedures, with regulations involving nuclear safety, and with operating license provisions. Results of these audits and surveillances will be submitted to both plant and corporate management.
Since the system for ALARA 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 Ma nager on radiological safety, including occupational exposure to personnel. Committee memb ership, responsibilities, authorities, 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-022 12.1-8 n. Seal glands of pumps carrying concentrated radwaste or spent filter and demineralizer resins are flushed with condensate water during operation. After operation, the pumps are also flushed 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) and Procedure Program. Procedures are formally reviewed for ALARA considerations as part of the approval process. The guidance provided by Regulatory Guide 8.8 is considered during this review. Additional information regarding ALARA considerations in procedures is provided in Section 12.5.3.2.
In addition to the above process, the 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,
- 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),
- d. Residual heat removal (RHR),
- 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-005 12.2-6 12.2.1.2.2.9 Traveling In-Core Probe System Sources. The prim ary 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 Resulting From Crud Buildup. Fission and corrosion product deposition is complex to describe analytically. In estimating the effects of this source, operating experience from other BWR plants is used (Reference 12.2-4); Section 12.4 discusses this in greater detail.
12.2.1.3 Radwaste Building
The radiation sources present in the radwaste building are discussed in Chapter 11.
12.2.1.4 Byproduct, Source and Special Nuclear Materials
A list of all byproduct, s ource and special nuclear material in sealed sources with an activity greater than 100 mCi is given in Table 12.2-12a. Several areas outside the Reactor, Turbine, and Radwaste Buildings and within the site area have been esta blished for use and storage of radioactive material in the form of activated components, seal ed sources, and contaminated tools, equipment, and protective clothing. Those areas where the total quantity of activity could at times be greater than 100 mCi unde r normal conditions are listed in Table 12.2-12b.
12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES
12.2.2.1 General
Design features that limit the airborne 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, Table 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 C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015 12.2-7 counting room is located at el. 487 ft 0 in. As seen in Figure 9.4-3 , the incoming ventilation flow into this room consists of fresh air. The outgoing air flow is to the area surrounding the counting room. It is c oncluded that the airborne concentration in the counting room is small.
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 leakage 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 v iv s c ,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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-14-005 12.2-28 Table 12.2-12a
Special Sources With Strength Greater Than 100 Millicuries
Isotope Identification Form Quantity (mCi) Use/Location 241AmBe 2-81-020 Solid 14,770 Calibration (EOF) 241AmBe 2-82-050 Solid 2,845 Neutron source (plant) 241AmBe 2-86-073 Solid 988 WNP-1 source (plant) 137Cs 2-79-033 Solid 524 Panoramic shepard cal (EOF) 137Cs 2-83-097 Solid 4,673 ARM calibration (plant) 137Cs 2-84-058 Solid 1,474 Shepard series 28 cal (EOF) 137Cs 2-88-002 Solid 133 Victoreen model 878-W cal (plant) 137 Cs 2-93-026 Solid 909 MG calibrator (EOF) 137 Cs 2-99-028 Solid 2,078 Mini-89 calibrator (plant) 137 Cs08-132 Solid 358,600 Hopewell calibrator (EOF) 137 Cs08-133 Solid 422 Hopewell calibrator (EOF) 137 Cs13-230 Solid 12,940 ARM calibration (plant) 238 PuBe 2-84-047 Solid 11,150 WNP-3 startup source (plant) 238 PuBe 2-84-048 Solid 11,260 WNP-3 startup source (plant)
Table as of 9/9/2015.
C OLUMBIA G ENERATING S TATION Amendment 63 F INAL S AFETY A NALYSIS R EPORT December 2015
LDCN-14-005 12.2-28a Table 12.2-12b
Radioactive Material Use and Storage Areas Outside the Main Radiologically Controlled Area Area Location Approximate Size (sq. ft.)
Normal Contents Normal Activity (mCi) LSA Storage
Pad Outside SW side of Radwaste Bldg 1800 Low-activity dry and compacted solid wast
containers 930 Building 13 West portion of Bldg 13 2700 Contaminated protective clothing 300 Warehouse 5 NE portion of Bldg 80 at
Snake River
Warehouse
Complex 4000 Radioactive &
contaminated equipment 590 Building 167 ~0.5 miles E of Plant 6332 Radioactive &
contaminated equipment 1370 Building 167 Storage Yard
~0.5 miles E of Plant 68000 C-van storage 1620 Outage Storage Yard Fenced area W of Bldg 105 2400 Outage radioactive material & C-van storage 114 Kootenai HP Calibration Lab Kootenai (Bldg
- 34) Rms 102 &
102A 600 Calibrators/irradiators, calibration sources, radioactive HP
instruments 377030 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)
Relief valve vents radioactive (VR)
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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-037 12.3-1 12.3 RADIATION PROTECTION DESIGN FEATURES
12.3.1 FACILITY DESIGN FEATURES
Columbia Generating Station plant incorporates the design objectives 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 12.3-5 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 less 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-037 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 control and traffic patter ns in the plant have been ev aluated to maintain personnel radiation exposures ALARA and to minimize the spread of contamination.
Normal entry into the plant is as follows:
- a. Personnel normally access the plant Protected Area through the security Protected Area Access Point (PAAP).
- b. The main Radiologically Controlled Area (RCA) normally 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 withstand 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 low 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 radiation zones is enclosed within 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 radioactivity, that may be introduced by the
handling of radioactively contaminated systems, as well as the number of man rems due to contact with the same systems, can be reduced by first decontaminating them. Means exist in the present design where radioactively contaminated systems can be decontaminated chemically 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 contaminated 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 mothballing, the above provisions will reduce to low
levels the occupational radiation exposure to personnel. According to Regulatory Guide 1.86, this method involves "putting the facility in a 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 could 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-13-039 12.3-9 a. Receiving and opening shipments as required by 10 CFR 20.1906,
- b. Storage of licensed materials as required by 10 CFR 20.1801 and 20.1802,
- c. Inventory and control of radioactive materials,
- d. Posting of radioactive material storage areas and tagging of source,
- e. Leak tests - sources 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 Radiological Services Manager/
Radiation Protection Manager (RPM) is responsible for the control and monitoring of seal ed and unsealed source and byproduct materials. The Nuclear Mate rial Manager appointed by the Engineering Manager is accountable for speci al nuclear materials (SNM).
The Chemistry Technical Supervisor is responsible for the minimization of radioactive waste and th e preparation, offsite shipment, and disposal of radioactive materials and radwaste.
Monitoring during handling of these materials is provided by Ra diation Protection. Experience and qualifications of Radiation Protection personnel are 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 equipment 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 radiation 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 direct radiation streaming at the site boundary.
The shielding arrangement for the major sources of radiation in the turbine building, including those discussed above, are shown in Figures 12.3-5 through 12.3-10.
12.3.2.3.4 Radwaste Building
The shielding arrangement for the major sources in the radwaste building are shown in Figures 12.3-11 through 12.3-14.
12.3.3 VENTILATION
The plant ventilation systems for the different areas of the plant are designed to meet the requirements of 10 CFR Parts 20 and 50. Gaseous wastes will be released in a controlled manner to environs such that during plant operation, onsite and offsite radioactivity levels are ALARA. The design features which limit and reduce the airborne radioactivity are as follows:
- a. In the reactor, radwaste, and turbine generator 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 which handles radioactivity contaminated material are ventilated at a minimum rate of three volume air changes per hour;
- d. All sinks and chemical laboratory work areas where radioactive samples or materials are handled are provided with exhaust hoods to protect operating personnel from airborne contaminants;
- e. All liquid equipment leaks which are 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 special filtration equipment to limit airborne radioactive contaminants are
- a. Standby gas treatment system (see Section 6.5), b. Control room emergency filtration system (see Sections 9.4 and 6.4), c. Reactor building sump vent exhaust filter system (see Section 9.4), and d. Radwaste building exhaust filtration system (see Section 9.4). In addition, small local absolute particulate filters are used to locally filter the effluent from sample sink hoods and chemical hoods.
These small filter 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 absolute particulate filters downstream of the adsorber section as described in Regulatory Guide 1.52.
The 5-ft spacing between filter frames, as required in Regulatory Guide 1.52, is not provided in these units. These units are of low capacity (1000 cfm), and personnel access into 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 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-031 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.
10 CFR 50.68(b) requires radiation monitors in storage and associated handling areas when fuel is present to detect excessive radiation levels and to alert personnel to initiate appr opriate safety actions.
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 sufficient 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 12.3-18 radiologically controlled areas, provide protection in areas where area radiation monitors have not been installed.
An additional area radiation monitor is installed on the 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 three installed particulate monitors measure the airborne particulate activ ity levels in the radwaste and reactor build ing ventilation exhaust and furnish recording signals to the main control room.
These units draw approximately 3 cfm air sample through the particulate filter which is monitored by a shie lded 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. External 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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 12.3-20 In the turbine building, airborne contamination is most likely to arise from nuclear steam leaks or offgas processing systems piping. Heater bay areas, steam jet air ejector rooms, turbine areas, and steam-driven feedwater pump areas have the potential for airborne contamination.
Regular radiological surveys in the turbine building are not augmented by continuous particulate monitoring of the turbine building ventilation exhaust.
Each of the continuous particulate monitors has an 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 Ci/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 63 F INAL S AFETY A NALYSIS R EPORT December 2015 LDCN-10-013 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 scintillation detectors with count ratemeters. Means for remote recording are provided for in the main control room. Sample collectors consist of shielded, fixed particulate, filter-type air collectors. The cali bration frequency will occur at least annually and after major maintenance. Instrument response checks will be made at least monthly. Monitors will be calib rated 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 r eactor and radwaste buildings. The monitors are located so as to monitor the exhaust air from that building prior to any filtration. In addition, charco al sampling cartridges are installe d in each monitor for laboratory analysis of iodine.
Each of the three channels of the airborne ra dioactivity monitors ha s an independent local visual and audible alarm. Hi gh radioactivity or equipment failure will generate an alarm signal. No automatic system functions are performed by the alarm signals.
12.3.4.5 Annunciators and Alarms The annunciators are of the window type with pushbutton switches for sound acknowledgment and light reset functions. There are no automatic system actions performed by the area radiation monitors. The annunciator alarm windows are located in the main control room.
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