ML21361A187
| ML21361A187 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/20/2021 |
| From: | Energy Harbor Nuclear Corp |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML21361A235 | List:
|
| References | |
| L-21-287 | |
| Download: ML21361A187 (242) | |
Text
TABLE OF CONTENTS Section Title Page 12.0 RADIATION PROTECTION 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1-1 12.1.1 POLICY CONSIDERATIONS 12.1-1 12.1.2 DESIGN CONSIDERATIONS 12.1-2 12.1.2.1 Design Features 12.1-2 12.1.2.2 Utilization of Experience Gained from Operating Facilities to Ensure that Radiation Doses are ALARA 12.1-4 12.1.2.3 Design Guidance Given to Individual Designers 12.1-5 12.1.2.4 Design Review Procedures 12.1-6 12.1.2.5 Decommissioning Design Considerations 12.1-6 12.1.3 OPERATIONAL CONSIDERATIONS 12.1-7 12.1.3.1 ALARA Training Program 12.1-8 12.1.3.2 Radiation Zoning and Access Control 12.1-11 12.2 RADIATION SOURCES 12.2-1 12.2.1 CONTAINED SOURCES 12.2-1 12.2.1.1 Source Terms 12.2-1 12.2.1.2 Reactor Building 12.2-2 12.2.1.3 Auxiliary Building 12.2-7 12.2.1.4 Intermediate Building 12.2-8 12.2.1.5 Turbine Building 12.2-9 12.2.1.6 Radwaste Building 12.2-9 12.2.1.7 Offgas Building 12.2-10 12.2.1.8 Sources Resulting From Design Basis Accidents 12.2-11 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 12.2-11 12.2.2.1 Reactor Building (Outside of Drywell) 12.2-11 12.2.2.2 Radwaste Building 12.2-13 12.2.2.3 Turbine Building 12.2-13 12.2.2.4 Fuel Handling Area of the Intermediate Building 12.2-13 12.2.2.5 Other Buildings 12.2-14 12.
2.3 REFERENCES
FOR SECTION 12.2 12.2-14 Revision 12 12-i January, 2003
TABLE OF CONTENTS (Continued)
Section Title Page 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 FACILITY DESIGN FEATURES 12.3-1 12.3.1.1 Equipment and Facility Design Features 12.3-1 12.3.1.2 Illustrative Examples of Plant Design Features to Minimize Occupational Doses 12.3-5 12.3.2 SHIELDING 12.3-8 12.3.2.1 Design Objectives 12.3-8 12.3.2.2 Design Description 12.3-8 12.3.3 VENTILATION 12.3-19 12.3.3.1 Design Bases 12.3-19 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION 12.3-24 12.3.4.1 Area Radiation Monitoring 12.3-25 12.3.4.2 Airborne Radioactivity Monitoring 12.3-36 12.3.4.3 Detection of MPC (DAC) Levels of Airborne Radioactivity 12.3-54 12.3.4.4 System Setpoints 12.3-67 12.
3.5 REFERENCES
FOR SECTION 12.3 12.3-68 12.4 DOSE ASSESSMENT 12.4-1 12.4.1 ESTIMATES OF PERSONNEL OCCUPANCY REQUIREMENTS 12.4-1 12.4.2 ESTIMATES OF ANNUAL PERSON-REM DOSES 12.4-2 12.4.3 ESTIMATED INHALATION DOSES 12.4-3 12.4.4 ESTIMATED ANNUAL DOSE OUTSIDE THE NUCLEAR FACILITY AT THE BOUNDARY OF THE RESTRICTED AREA 12.4-5 12.4.4.1 Skyshine and Direct Dose from Turbines 12.4-5 12.4.4.2 Direct Doses From Stored Radwaste 12.4-9 12.4.4.3 Direct Doses From the External Surfaces of Buildings 12.4-9 12.4.4.4 Doses From Gaseous Radioactive Plume 12.4-9 12.
4.5 REFERENCES
FOR SECTION 12.4 12.4-10 Revision 13 12-ii December, 2003
TABLE OF CONTENTS (Continued)
Section Title Page 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 Facilities 12.5-2 12.5.2.2 Access and Egress to Radiation Protection Controlled Areas 12.5-4 12.5.2.3 Health Physics Instrumentation 12.5-5 12.5.3 HEALTH PHYSICS INSTRUCTIONS 12.5-9 12.5.3.1 Radiation and Contamination Surveys 12.5-9 12.5.3.2 Procedures and Methods Ensuring ALARA 12.5-10 12.5.3.3 Controlling Access and Stay Time 12.5-13 12.5.3.4 Contamination Control 12.5-14 12.5.3.5 Training Programs 12.5-15 12.5.3.6 Personnel Dosimetry 12.5-17 12.5.3.7 Evaluation and Control of Potential Airborne Radioactivity 12.5-19 12.5.3.8 Radioactive Material Handling and Storage Methods 12.5-20 12.6 DESIGN REVIEW OF PLANT SHIELDING FOR SPACES/
SYSTEMS WHICH MAY BE USED IN POSTACCIDENT OPERATIONS OUTSIDE CONTAINMENT 12.6-1 12.
6.1 INTRODUCTION
12.6-1 12.6.2 RADIOACTIVE SOURCE RELEASE 12.6-2 12.6.3 RADIOACTIVE SOURCE DISTRIBUTION 12.6-3 12.6.4 SYSTEMS CONTAINING RADIOACTIVE SOURCES 12.6-6 12.6.4.1 LPCS, HPCS, RHR, (LPCI mode), RCIC Systems 12.6-6 12.6.4.2 RHR (Shutdown Cooling Mode) 12.6-7 12.6.4.3 RHR (Suppression Pool Cooling Mode) 12.6-7 12.6.4.4 RWCU 12.6-7 12.6.4.5 Liquid Radwaste System 12.6-7 12.6.4.6 Sampling System 12.6-8 12.6.4.7 Offgas System 12.6-8 12.6.4.8 Annulus Exhaust Gas Treatment System (AEGTS) 12.6-8 12.6.4.9 Feedwater Leakage Control System 12.6-8 12.6.5 SHIELDING REVIEW 12.6-8a Revision 15 12-iii October, 2007
TABLE OF CONTENTS (Continued)
Section Title Page 12.6.6 AREAS REQUIRING PERSONNEL ACCESS 12.6-9 12.6.7 POSTACCIDENT RADIATION ZONE DRAWINGS AND
SUMMARY
12.6-11 12.6.7.1 Radiation Dose Rates as a Function of Time Following an Accident 12.6-11 12.
6.8 REFERENCES
FOR SECTION 12.6 12.6-11 Revision 12 12-iv January, 2003
LIST OF TABLES Table Title Page 12.2-1 Basic Reactor Data 12.2-15 12.2-2 Core Boundary Neutron Fluxes 12.2-18 12.2-3 Gamma Ray Source Energy Spectra 12.2-20 12.2-4 Gamma Ray and Neutron Fluxes Outside the Vessel Wall 12.2-22 12.2-5 Radiation Shielding Source Terms 12.2-24 12.2-6 Material Composition of the TIP Detectors and Cables 12.2-28 12.2-7 Radiation Levels From the TIP Detector and Cables 12.2-29 12.2-8 Traversing Incore Probe Detector Decay Gamma Activities of Materials in the Detector 12.2-30 12.2-9 Decay Gamma Activities of Materials in the Cable 12.2-31 12.2-10 Typical Turbine Component N-16 Inventories 12.2-32 12.2-11 Parameters and Assumptions Used in Calculating Reactor Building Airborne Activity 12.2-34 12.2-12 Reactor Building Airborne Activity 12.2-35 12.2-13 Radwaste Building Airborne Activity 12.2-36 12.2-14 Turbine Building Airborne Activity 12.2-37 12.2-15 Fuel Handling Area Airborne Activity 12.2-38 12.3-1 (Deleted) 12.3-69 12.3-2 Radiation Shield Thicknesses 12.3-70 12.3.3 Comparison of Non-ESF Charcoal Filter Systems to Regulatory Guide 1.140 Criteria 12.3-73 12.3-4 Detectors Associated with Unit 1 12.3-77 12.3-5 (Deleted) 12.3-79 Revision 19 12-v October, 2015
LIST OF TABLES (Continued)
Table Title Page 12.3-6 Detectors Associated with Common Areas 12.3-81 12.3-7 Detectors Associated with Local Channels 12.3-82 12.3-8 Detector Design Requirements 12.3-83 12.3-9 Readout Module Design Requirements 12.3-84 12.3-10 Airborne Radiation Monitor Subgroup Unit 1 and Common 12.3-85 12.3-11 Isokinetic Probes 12.3-87 12.3-12 Reactor Building Subcompartment Ventilation Data M11, M14 Systems 12.3-88 12.3-13 Radwaste Building Subcompartment Ventilation Data M31 System 12.3-89 12.3-14 Auxiliary Building Subcompartment Ventilation Data M38 System 12.3-91 12.3-15 Intermediate Building Subcompartment Ventilation Data M33 System 12.3-92 12.3-16 Fuel Handling Area Subcompartment Ventilation Data M40 System 12.3-93 12.3-17 Heater Bay Subcompartment Ventilation Data M41 System 12.3-94 12.3-18 Turbine Building Subcompartment Ventilation Data M35 System 12.3-95 12.3-19 Offgas Building Ventilation Exhaust Subcompartment Ventilation Data M36 System 12.3-96 12.4-1 Estimated Manpower Needs and Occupancy Requirements, Unit 1 and Unit 2 12.4-12 12.4-2 Person-rem Estimates for Normal Plant Operations, Anticipated Operational Occurrences and Routine Maintenance, Unit 1 and Unit 2 12.4-13 12.4-3 Boiling Water Reactors Percentages of Exposure by Job Function 12.4-14 Revision 19 12-vi October, 2015
LIST OF TABLES (Continued)
Table Title Page 12.4-4 Boiling Water Reactors Percentages of Exposure by Work Function 12.4-15 12.4-5 Yearly Operational Person-rem for Selected BWR Plants 12.4-16 12.4-6 Summary of Total Occupational Radiation Exposure Estimates by Task 12.4-17 12.4-7 Occupational Dose Estimates During Routine Operations and Surveillance 12.4-18 12.4-8 Occupational Dose Estimates During Routine Maintenance 12.4-20 12.4-9 Occupational Dose Estimates During Waste Processing 12.4-21 12.4-10 Occupational Dose Estimates During Refueling 12.4-22 12.4-11 Occupational Dose Estimates During Inservice Inspection 12.4-23 12.4-12 Occupational Dose Estimates During Special Maintenance 12.4-24 12.4-13 Personnel Radiation Doses From Airborne Activity 12.4-25 12.4-14 Safety/Relief Valve Dischargers Dose for Type 2 Event 12.4-26 12.4-15 Estimated Skyshine Doses 12.4-27 12.4-16 Work Force by Occupational Crafts 12.4-28 12.4-17 Summary of Direct Doses 12.4-29 12.4-18 Dose to Non-Plant Personnel 12.4-30 12.4-19 Direct Dose from the External Surfaces of Buildings 12.4-32 12.4-20 Dose from Gaseous Radioactive Plume 12.4-33 12.5-1 Laboratory Equipment 12.5-21 Revision 13 12-vii December, 2003
LIST OF TABLES (Continued)
Table Title Page 12.5-2 Portable Survey Instruments 12.5-22 12.5-3 Personnel Monitoring Instruments 12.5-23 12.5-4 Radiation Protection Equipment 12.5-24 12.6-1 Occupancy and Radiation Design Objectives 12.6-12 12.6-2 Initial Core Inventory 12.6-13 12.6-3 Initial Radioactive Source Terms (Gammas/cc-sec) 12.6-17 12.6-4 Dose Rates 12.6-18 Revision 12 12-viii January, 2003
12.0 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS The management of the Energy Harbor Nuclear Corp. recognizes its responsibility and authority to operate and maintain the Perry Nuclear Power Plant in a manner that provides for the safety of plant personnel and the public. In accordance with the regulations, the company will maintain the policy of keeping radiation exposure as low as reasonably achievable (ALARA).
It is the intent of the ALARA Program to demonstrate that reasonable measures have been taken to maintain the radiation exposure of plant personnel and members of the public as far below the regulatory limits as reasonably obtainable.
The Manager, Site Chemistry/Radiation Protection is responsible for the radiation health and safety of PNPP. To implement his responsibilities, the Manager, Radiation Protection ensures: the necessary supervisory and technical support is available to provide radiation protection program oversight for monitoring radiological work activities during plant operation, maintenance, and refueling; the planning of the radiological work activities is accomplished to minimize worker doses; and the necessary actions are taken to reduce radiation sources within the plant ensuring occupational radiation doses are maintained ALARA during all radiological work activities.
The Director, Site Engineering, has the responsibility to assure that all design work includes appropriate ALARA considerations for installation, operation and maintenance of each design package.
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The General Plant Manager, and the Director, Performance Improvement Department have the complete responsibility for all onsite activities in connection with the safe and efficient operation and maintenance of the PNPP. They are responsible for managing the affairs of the plant to ensure reliable, efficient and safe operation.
The General Plant Manager, the Director, Site Engineering, and the Radiation Protection Manager work in concert to implement the Companys ALARA policy. However, it is the responsibility of all supervision to enforce the requirements for keeping radiation doses ALARA, and the responsibility of each individual to comply with these requirements.
12.1.2 DESIGN CONSIDERATIONS 12.1.2.1 Design Features Ensuring that occupational radiation doses are ALARA was begun during the early design of PNPP. By familiarizing design engineers with ALARA concepts, and by providing review of the design by radiation protection personnel, the operation of the PNPP will result in personnel doses that are ALARA and will fulfill the intent of <Regulatory Guide 8.8>,
<Regulatory Guide 8.10>, and <10 CFR 20>.
The following design criteria have been, and will continue to be, adhered to:
- a. Access labyrinths are provided for rooms housing equipment that contain high radiation sources to preclude a direct radiation path from the equipment to accessible areas.
Revision 20 12.1-2 October, 2017
- b. Piping penetrations, ducts and voids in radiation shield walls are located to preclude the possibility of streaming from a high to low radiation area or otherwise will be adequately shielded.
- c. Shielding discontinuities caused by shield plugs, concrete hatch covers and shield doors to high radiation areas are provided with offsets to reduce radiation streaming.
- d. Radioactive piping is routed through high radiation areas where practicable, or in shielded pipe chases in low radiation areas.
- e. Sufficient work area and clearance space is provided around equipment to permit ease of servicing.
- f. Instruments requiring in situ calibration are not normally located in high radiation areas.
- g. Non-radioactive equipment which requires servicing is not normally located in proximity with potentially radioactive equipment.
- h. Spread of contamination from radioactive spillage is minimized by providing a floor drain system which collects and routes the liquid to the liquid waste processing system for proper handling.
Decontamination of an area is facilitated by use of materials and coatings which lend themselves to cleaning by standard methods.
- i. Natural traps which could be potential pockets for corrosion product activity are minimized in pipe and ducts by avoiding sharp bends, rough finishes and cracks.
- j. Shielding is provided for equipment which is anticipated to be normally radioactive. The dose levels are designed not to exceed
<10 CFR 20> requirements under the worst operating conditions of the plant.
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- k. Temporary shielding, such as lead blankets, is available on the site in case it is ever needed.
- l. Remote handling of radioactive materials is provided wherever it is needed and practicable.
- m. Process piping that may contain radioactive fluids is routed and dimensioned on piping system drawings, thereby minimizing any field run radioactive piping.
- n. Redundant components for the radwaste processing system have been supplied to increase flexibility in plant operations and decrease radiation doses during maintenance.
- o. Most radioactive components can be flushed to decrease radiation levels and subsequent personnel doses from maintenance.
- p. The radwaste handling system is designed to the extent practicable to be remotely operated and have remote-manual overrides in case a failure occurs.
12.1.2.2 Utilization of Experience Gained from Operating Facilities to Ensure that Radiation Doses are ALARA An important aspect in the design of a nuclear facility is the feedback information obtained from plants currently operating. For this design feedback, information has been obtained directly from operating facilities and from governmental and industrial publications. The following areas of information represent the type of feedback useful in plant design:
- a. Operational radiation levels.
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- b. Trends in radiation levels associated with years of operation based on plant type, plant size, power levels, and plant design.
- c. Radiation zones as determined by occupancy requirements and actual radiation levels.
- d. Location of components with respect to plant operability.
- e. Reliability of components.
- f. Adequacy of plant layout in terms of traffic patterns, and space allocation, such as around radioactive components requiring maintenance, inspections and pipe routing.
- g. Number of plant employees associated with different tasks and the resulting person-rem doses.
Feedback information is used in the design of the plant to identify potential problem areas. Such problems are reviewed with regard to current design and, where applicable, modifications made to the design to eliminate or decrease the potential for such problems to arise.
12.1.2.3 Design Guidance Given to Individual Designers The following design guidance has been given to individual designers of systems associated with radiation protection:
- a. Follow guidance given in <Regulatory Guide 8.8> and <Regulatory Guide 8.10>.
- b. Use maximum realistic source strengths in all calculations.
- c. Consider all sources that significantly contribute to the dose at a particular point.
Revision 13 12.1-5 December, 2003
- d. Place instrument readouts and items needing routine surveillance or attention in as low as practicable radiation zone.
- e. Make the design sensitive to the expected procedures of plant personnel under normal operation and anticipated occurrences.
12.1.2.4 Design Review Procedures Independent design reviews are performed by competent specialists to assess the implementation of the design criteria and further ensure that the final design is compatible with maintaining occupational radiation doses ALARA. All designs influencing radiological control in the plant are reviewed by competent professionals in the area of radiation protection.
As indicated in <Section 12.1.2.3>, personnel associated with design aspects influencing radiation protection have been given the basic ALARA principles. After the preliminary design and layout of the system, a shielding engineer analyzes the various radiation sources and specifies shielding as necessary to conform to the appropriate radiation zone requirements. Radiation protection personnel review the system in terms of the total plant operation and specify necessary changes to keep occupational radiation doses ALARA. Radiation protection personnel are part of the overall design team and report their findings directly to the design project management for the PNPP. These findings are considered in conjunction with design requirements from disciplines not directly associated with radiation control to determine what modifications, if any, should be made to promote ALARA radiation doses.
12.1.2.5 Decommissioning Design Considerations The radiation protection aspects of the plant design are described in
<Section 12.1.2> and <Section 12.3.1>. While these features assure that the plant can be operated and maintained with ALARA exposures, they will Revision 12 12.1-6 January, 2003
also aid in the ALARA aspects of decommissioning. These include shielding, accessibility criteria, equipment separation, decontamination features, and system design considerations.
In addition, a detailed study was performed for the plant that designates the required removal paths for major equipment located within structures of the nuclear island and secondary plant. The report contains the following information.
- a. A list of path numbers by building and floor elevation including minimum path size and floor loading (PSF - live load) for each designated path.
- b. Identification of major equipment for each structure on a floor-by-floor basis including total equipment and component weights, as applicable.
- c. Where removal requirements were defined, required route numbers are indicated for each piece of equipment. Where elevators are used in the removal scheme, elevations are indicated for loading and unloading points.
- d. Plant layout drawings are also provided showing the major equipment and removal routes for each floor elevation.
12.1.3 OPERATIONAL CONSIDERATIONS The site ALARA Committee is comprised of plant staff personnel representing the various disciplines, such as Radiation Protection, Operations, Maintenance, and Nuclear Engineering. The site ALARA Committee is responsible for the overall coordination of the Station ALARA Program and for advising site management in matters relating to ALARA in accordance with <Regulatory Guide 8.8>.
Revision 15 12.1-7 October, 2007
Responsibilities of the site ALARA Committee shall include:
- a. Ensure that a program is established with a management philosophy to maintain occupational exposures ALARA, with specific goals and objectives for implementation,
- b. Ensure that an effective measurement system is established and used to determine the degree of success achieved by station operations with regard to the program goals and specific objectives;
- c. Ensure that the measurement system results are reviewed on a periodic basis and that corrective actions are taken when attainment of the specific objectives appears to be jeopardized;
- d. Ensure that the authority for providing procedures and practices by which the specific goals and objectives will be achieved is delegated; and
- e. Ensure that the resources needed to achieve goals and objectives to maintain occupational radiation exposures ALARA are made available.
12.1.3.1 ALARA Training Program The Radiation Worker Training (RWT) at PNPP will help implement the Companys ALARA policy in accordance with <10 CFR 20>. RWT will help Revision 15 12.1-8 October, 2007
workers understand how radiation protection relates to their jobs and all workers will have frequent opportunities to discuss radiation safety with the Radiation Protection Section personnel when the need arises.
A Radiation Work Permit (RWP) will be initiated for work activities based on the radiation, removable contamination and/or airborne activity levels to be encountered and by the area of the plant where the work will take place. An RWP is also required for work which involves open radioactive systems or when required by Radiation Protection.
The Radiation Work Permit will help implement the Companys ALARA policy in accordance with <10 CFR 20> by defining the radiological hazards and requiring specific radiological precautions. The RWP also becomes a record of how various jobs were performed and the radiological problems associated with specific jobs. By reviewing expired RWPs, recommendations can be made to change procedures or equipment that will result in lower radiation exposures in the future.
Training and RWP requirements will help ensure that the Companys ALARA policy is fulfilled. Some techniques that may be used are:
- a. Temporary shielding may be used. Temporary shielding will be used only if total exposure, which includes installation and removal of the shielding, will be effectively reduced.
- b. Prior to performing maintenance work, consideration will be given to flushing and/or chemically decontaminating in order to reduce crud levels and personnel exposure.
- c. Dry run training will be used for jobs with exceptional radiological problems to familiarize personnel with the work they must perform at the job site. These techniques will assist in Revision 12 12.1-9 January, 2003
improving efficiency and minimize the amount of time spent in radiation areas. These efforts will be documented to improve future efforts.
- d. As much as practicable, work will be performed outside radiation areas. This includes items such as reading instruction manuals or procedures, adjusting tools or jigs, repairing valve internals, and prefabricating components.
- e. For long term repair jobs, consideration will be given to establishing remote observation stations to assist supervising personnel in monitoring work progress from a lower radiation area.
- f. On some jobs, special tools or jigs will be used so that work will be performed more efficiently to reduce errors, thus minimizing the time spent in a radiation area. Special tools may be used to increase the distance from a radiation source to the worker, thereby reducing the exposure.
- g. Entry and exit control points will be established in areas with low levels of radiation to limit the exposure of personnel donning protective equipment or generally preparing to work in such areas.
The access control points will be designed to minimize the spread of contamination from the work areas.
- h. Protective clothing and respiratory protection equipment will be selected to minimize the discomfort of workers and to minimize the Total Effective Dose Equivalent (TEDE) of the workers. Efficiency will increase and less time will be spent in radiation areas.
- i. Personnel will be assigned self reading dosimeters to estimate exposure during a work assignment.
Revision 12 12.1-10 January, 2003
- j. On jobs where general area radiation levels are high, radiation protection coverage may be required during the period of work.
- k. On intricate jobs, especially those which involve high radiation levels, preplanning will include estimation of the person-rem needed to complete the job. At the completion of the work, a debriefing session may be held with the personnel that performed the work (when practical) in an effort to determine how the work could have been completed more efficiently and with less radiation exposure.
12.1.3.2 Radiation Zoning and Access Control During normal full power operation, the design maximum whole body dose rates within the plant that might be received by operating personnel, contractors and authorized visitors will depend upon the following zone designations:
Dose Rate Zone Designation (mrem/hr)
I Unlimited Occupancy 0.5 II Normal Continuous Occupancy 2.5 III Controlled, Limited Access (4hr/wk) 25 IV Controlled, Limited Access (1hr/wk) 100 V Controlled, Limited Access (<1 hr/wk) >100 12.1.3.2.1 Zone I Zone I areas can be occupied by station personnel or visitors on an unlimited time basis. Health hazards due to ionizing radiation in Zone I are absolutely minimal. Access control to Zone I areas is due to security considerations.
Revision 13 12.1-11 December, 2003
12.1.3.2.2 Zone II Design allows Zone II areas to have a higher dose rate than Zone I areas but still allows normal, continuous occupancy. Areas where radiation levels exist such that a major portion of the body could receive in any one hour a dose of 0.5 mrem or more.
12.1.3.2.3 Zone III Zone III areas are designed for a maximum dose rate of 25 mrem/hr and for limited access of less than or equal to 4 hr/wk. Areas in which there exists radiation at such levels that a major portion of the whole body could receive in one hour a dose in excess of 5 mrem at 30 cm from the radiation source or from any surface that the radiation penetrates, will be designated as Radiation Areas. Access to a Zone III, Zone IV or a Zone V area will require personnel monitoring and a Radiation Work Permit.
12.1.3.2.4 Zone IV Zone IV areas are designed for limited access of less than or equal to 1 hr/wk and a maximum dose rate of 100 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the radiation source or from any surface that the radiation penetrates.
Zone IV areas will be designated as Radiation Areas as described in
<Section 12.1.3.2.3> above.
12.1.3.2.5 Zone V Zone V areas are designated as High Radiation Areas (greater than 100 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm from the Radiation Source or from any surface that the radiation penetrates). Access shall Revision 17 12.1-12 October, 2011
not occur unless absolutely necessary. High Radiation Areas with general area dose rates greater than 1,000 mrem/hr or Very High Radiation Area where radiation levels could result in an individual receiving an absorbed dose in excess of 500 RADs in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter from a radiation source or from any surface that the radiation penetrates, will be controlled in accordance with Perry Technical Specification 5.7.
12.1.3.2.6 General Guidelines for Maintenance in High Radiation Areas
- a. Work will be performed in accordance with Regulatory Guide 8.8, Information Relevant To Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Achievable and in accordance with the applicable Radiation Protection Nuclear Operating Procedures.
Revision 17 12.1-13 October, 2011
(DELETED)
Revision 17 12.1-14 October, 2011
(DELETED)
Revision 17 12.1-15 October, 2011
12.2 RADIATION SOURCES 12.2.1 CONTAINED SOURCES 12.2.1.1 Source Terms With the exception of the vessel and drywell shields, shielding designs are based on fission product and activation product sources consistent with <Section 11.1>. For shielding, it is conservative to design for fission product sources at peak values rather than an annual average, even though experience supports a lower annual average than the design average (Reference 1). It should be noted that activation products, principally Nitrogen-16, control shielding calculations in most of the primary system. In areas where fission products are significant, conservative allowance is made for transit decay, while at the same time providing for transient increase of the noble gas source, daughter product formation and energy level of emission. Areas where fission products are significant relative to Nitrogen-16 include: the condenser offgas system downstream of the steam jet air ejector, liquid and solid radwaste equipment, portions of the reactor water cleanup system, and portions of the feedwater system downstream of the hotwell including condensate treatment equipment.
For application, the design sources are grouped first by location and then by equipment type (e.g., reactor building, core sources). The following paragraphs represent the source data in various pieces of equipment throughout the plant. General locations of equipment are shown in the general plant arrangement drawings of <Section 1.2> and
<Section 12.3>.
Revision 12 12.2-1 January, 2003
12.2.1.2 Reactor Building 12.2.1.2.1 Radiation Sources from the Reactor Core The information in this section defines a reactor vessel model and the associated gamma and neutron radiation sources. This section is designed to provide the data required for calculations beyond the vessel. The data selected were not chosen for any given program, but were chosen to provide information for any of several shield program types. In addition to the source data, calculated radiation dose levels are provided at locations surrounding the vessel. This data is given as a potential check point for calculations by shield designers.
12.2.1.2.1.1 Physical Data
presents the physical data required to form the model in <Figure 12.2-1>. This model was selected to contain as few separate regions as possible to adequately portray the reactor.- a. Core Spectrum
- b. Post-Operation Gamma Ray Energy Spectrum
- c. Non-Core Gamma Ray Energy Spectra In
- a. Traversing Incore Probe (TIP)
- b. Reactor Startup Sources The reactor startup sources were shipped to the site in a special cask designed for shielding. The sources were transferred under water while in the cask and were then loaded into the reactor while remaining under water. The sources will be removed after the initial operating cycle and will be stored in the fuel storage pools for later handling and removal. The removal of sources will be handled in accordance with approved procedures and specified shielding requirements.
2.3 REFERENCES
FOR SECTION 12.2
- 1. Smith, J. M., Noble Gas Experience in Boiling Water Reactors, Paper No. A-54, presented at Noble Gases Symposium, Las Vegas, Nevada, September 24, 1974.
- 2. General Electric Co., Mark III Containment Dose Reduction Study, 22A5718 Rev. 2, Jan. 29, 1980.
- 3. Johnson, A. B., Behavior of Spent Nuclear Fuel in Water Pool Storage, BNWL-2256, Batelle, Pacific Northwest Laboratories, September 1977.
Revision 12 12.2-14 January, 2003
TABLE 12.2-1 BASIC REACTOR DATA (Data used in the Original Design of the Nuclear Island Shields)
A. Reactor thermal power, MW 3,579 B. Average power density, watts/cm3 54.07 (1)
C. Physical dimensions Radii
_(in)
- 1. Core equivalent 92.58
- 2. Inside shroud 99.90
- 3. Outside shroud 101.90
- 4. Inside vessel (nominal) 119.0
- 5. Outside vessel (nominal) 125.0
- 6. Outside vessel (reinforced - nominal) 125.75
- 7. Shroud head inside 192.0
- 8. Vessel top head inside 119.0
- 9. Vessel bottom head inside 130.19 Distance
_(in)
- 10. Outside of vessel bottom head -7.75
- 11. Inside of vessel bottom head -0.25
- 12. Vessel bottom head tangent 129.94
- 13. Bottom of core support plate 202.56
- 14. Top of core support plate 204.56
- 15. Bottom of active fuel 213.50
- 16. Top of reinforced vessel wall 210.00
- 17. Top of active fuel 363.5
- 18. Bottom of top guide 371.31 Revision 12 12.2-15 January, 2003
TABLE 12.2-1 (Continued)
Distance (in)
- 19. Top of fuel channel 377.87
- 20. Shroud head tangent 424.23
- 21. Inside of shroud head 452.27
- 22. Outside of shroud head 454.27
- 23. Normal vessel water level 566.6
- 24. Top of steam dryer 720.63
- 25. Vessel top head tangent 727.0
- 26. Inside of vessel top head 846.0
- 27. Outside of vessel top head 849.0 D. Material Densities, gm/cm3 Region(2) Coolant UO2 Zircaloy 340 Stainless A 0.740 0.0 0.0 0.178 B 0.338 0.0 0.0 4.349 C 0.318 2.3 0.978 0.056 C-1 0.597 0.0 0.166 1.697 C-2 0.234 0.0 1.099 0.255 D 0.240 0.0 1.004 1.209 E 0.390 0.0 0.0 0.0 F 0.669 0.0 0.0 0.200 G 0.036 0.0 0.0 0.0 H 0.740 0.0 0.0 0.0 I 0.740 0.0 0.0 0.26 Revision 12 12.2-16 January, 2003
TABLE 12.2-1 (Continued)
E. Typical Core Power Distributions Axial Power Distribution Radial Power Distribution (Typical end-of-life)
Percent of Distance(3)
Equivalent Radius Relative Power __(in)_____ Relative Power 0 1.2 -75 0.343 20 1.2 -68 0.755 35 1.19 -60 1.055 50 1.17 -48 1.190 60 1.15 -36 1.200 70 1.12 -24 1.190 80 1.05 -12 1.170 85 0.995 0 1.155 90 0.778 12 1.140 92.5 0.590 24 1.105 95.0 0.430 36 1.055 97.0 0.375 48 0.945 98.0 0.395 60 0.715 99.0 0.432 68 0.462 100.0 0.518 75 0.212 NOTES:
(1)
Relative locations of dimensions are shown in <Figure 12.2-1>.
(2)
Region locations are shown in <Figure 12.2-1>.
(3)
Distances are measured from the mid-plane of the core.
Revision 12 12.2-17 January, 2003
TABLE 12.2-2 CORE BOUNDARY NEUTRON FLUXES (Data used in the Original Design of the Containment Shields)
Neutron Energy Bounds (neutrons/cm2 - sec) _
16.5 MeV 3.9 E+10 10.00 MeV 5.5 E+11 6.065 MeV 2.0 E+12 3.679 MeV 3.8 E+12 2.231 MeV 4.4 E+12 1.353 MeV 3.9 E+12 820.8 KeV 3.8 E+12 497.9 KeV 2.6 E+12 302.0 KeV 2.3 E+12 183.2 KeV 3.2 E+12 67.38 KeV 2.2 E+12 24.79 KeV 2.2 E+12 9.119 KeV 2.0 E+12 3.355 KeV 2.0 E+12 1.234 KeV 1.9 E+12 454.0 eV 2.0 E+12 167.0 eV 1.9 E+12 61.44 eV 1.8 E+12 22.60 eV 8.8 E+11 13.71 eV 8.8 E+11 8.315 eV 8.2 E+11 5.043 eV 8.4 E+11 3.059 eV 8.3 E+11 Revision 12 12.2-18 January, 2003
TABLE 12.2-2 (Continued)
Neutron Energy Bounds (neutrons/cm2 - sec)_
1.855 eV 8.2 E+11 1.125 eV 8.8 E+11 0.616 eV 3.2 E+13 0.00 eV Revision 12 12.2-19 January, 2003
TABLE 12.2-3 GAMMA RAY SOURCE ENERGY SPECTRA (Data used in the Original Design of the Containment Shields)
A. Gamma ray sources in the core during operation Energy Gamma Ray Source Bounds (MeV) (MeV/sec-watt) _
16.5 8.0 E+8 8.0 7.3 E+9 6.0 5.9 E+10 4.0 5.8 E+10 3.0 5.2 E+10 2.6 6.7 E+10 2.2 7.2 E+10 1.8 8.3 E+10 1.4 9.1 E+10 1.0 7.5 E+10 0.75 6.8 E+10 0.5 6.0 E+10 0.25 9.8 E+10 0.003 B. Post-operation gamma sources in core(1) (MeV/sec - watt)
Energy Time After Shutdown Bounds (MeV) 0 Sec. 1 Day 1 Week 1 Month 6.0 8.2 E+9 <1.0 E+6 <1.0 E+6 <1.0 E+6 4.0 1.8 E+10 7.0 E+6 4.6 E+6 <1.0 E+6 3.0 1.1 E+10 5.7 E+6 3.7 E+6 <1.0 E+6 2.6 1.7 E+10 2.9 E+8 1.7 E+8 <2.0 E+7 2.2 2.1 E+10 4.5 E+8 4.0 E+7 4.0 E+7 1.8 3.3 E+10 3.1 E+9 2.1 E+9 6.4 E+8 1.4 3.7 E+10 2.3 E+9 1.6 E+9 1.1 E+9 0.9 5.1 E+10 7.5 E+9 3.8 E+9 2.1 E+9 0.4 1.2 E+10 1.8 E+9 8.7 E+8 3.6 E+8 0.1 Revision 12 12.2-20 January, 2003
TABLE 12.2-3 (Continued)
C. Gamma ray sources in non-core regions during operation (MeV/cm3-sec-watt)
Energy Region H Shroud Bounds (MeV) Inside Outside Inside Outside 16.5 2.8 E-1 3.6 E-2 2.5 E+2 4.1 E+1 8.0 2.5 3.0 E-1 8.2 E+2 1.3 E+2 6.0 4.8 E-3 6.2 E-4 2.4 E+2 3.9 E+1 4.0 2.3 E-2 4.1 E-3 1.1 E+2 1.9 E+1 3.0 1.0 E-3 1.4 E-4 4.3 E+1 1.3 E+1 2.6 2.3 E+2 4.9 E+1 2.4 E+1 4.5 2.2 5.4 E-3 1.0 E-3 2.7 E+1 5.4 1.8 6.3 E-5 8.1 E-6 7.1 E+1 1.3 E+1 1.4 2.6 E-3 5.0 E-4 3.3 E+1 6.5 1.0 7.5 E-3 1.4 E-3 3.9 E+1 9.9 0.75 4.6 E-4 5.8 E-5 2.9 E+1 4.7 0.5 - - 1.2 E+2 1.9 E+1 0.25 - - 9.3 E+1 1.5 E+1 0.003 Energy Region I (Jet Pumps) Vessel Bounds (MeV) Inside Outside Inside Outside 16.5 1.4 2.2 E-2 2.2 E-1 2.1 E-4 8.0 4.6 6.9 E-2 2.1 1.6 E-3 6.0 1.3 2.0 E-2 5.6 E-1 1.6 E-3 4.0 6.3 E-1 9.3 E-3 2.8 E-1 2.0 E-3 3.0 2.6 E-1 3.7 E-3 1.0 E-1 1.2 E-3 2.6 6.1 4.7 E-2 4.7 E-2 1.1 E-3 2.2 1.8 E-1 2.4 E-3 5.3 E-2 1.4 E-3 1.8 4.2 E-1 6.0 E-3 1.8 E-1 1.1 E-3 1.4 2.1 E-1 2.9 E-3 7.5 E-2 9.3 E-4 1.0 3.2 E-1 3.6 E-3 9.1 E-2 5.5 E-3 0.75 1.6 E-1 2.4 E-3 6.4 E-2 4.2 E-5 0.50 6.6 E-1 9.9 E-3 2.5 E-1 2.0 E-4 0.25 5.1 E-1 7.6 E-3 1.9 E-2 1.8 E-4 0.003 NOTE:
(1)
Operating history of 3.2 years.
Revision 12 12.2-21 January, 2003
TABLE 12.2-4 GAMMA RAY AND NEUTRON FLUXES OUTSIDE THE VESSEL WALL (Data used in the Original Design of the Containment Shields)
A. Neutron fluxes Energy Bounds Neutrons/cm2-sec.
16.5 MeV 5.8 E+6 10.00 MeV 2.9 E+7 6.065 MeV 2.2 E+7 3.679 MeV 4.5 E+7 2.231 MeV 7.5 E+7 1.353 MeV 1.1 E+8 820.8 KeV 1.6 E+8 497.9 KeV 1.5 E+8 302.0 KeV 9.1 E+7 183.2 KeV 1.1 E+8 67.38 KeV 1.2 E+7 24.79 KeV 6.7 E+7 9.119 KeV 1.4 E+7 3.355 KeV 8.6 E+6 1.234 KeV 6.4 E+6 454.0 eV 2.9 E+6 167.0 eV 4.2 E+6 61.44 eV 3.9 E+6 22.60 eV 1.9 E+6 13.71 eV 2.0 E+6 8.315 eV 1.8 E+6 5.043 eV 1.6 E+6 3.059 eV 1.5 E+6 Revision 12 12.2-22 January, 2003
TABLE 12.2-4 (Continued)
A. Neutron fluxes (Continued)
Energy Bounds Neutrons/cm2-sec.
1.855 eV 1.4 E+6 1.125 eV 7.9 E+5 0.616 eV 6.0 E+5 0.000 eV B. Gamma ray energy fluxes Energy Bounds
_____(MeV) _ MeV/cm2-sec2 16.5 1.0 E+9 8.0 3.4 E+9 6.0 3.3 E+9 4.0 1.7 E+9 3.0 7.0 E+8 2.6 1.0 E+9 2.2 6.9 E+8 1.8 6.1 E+8 1.4 5.3 E+8 1.0 3.2 E+8 0.75 4.2 E+8 0.50 4.0 E+8 0.25 1.5 E+8 0.003 Revision 12 12.2-23 January, 2003
TABLE 12.2-5 RADIATION SHIELDING SOURCE TERMS Source Shielding Sources (/cc-sec)
Volume Equipment Identification (cc) 0.2 MeV 0.6 MeV 1.0 MeV 1.6 MeV 2.4 MeV 3.4 MeV 5.0 MeV 6.1 MeV 7.1 MeV A. AUXILIARY BUILDING E12B001 RHR hx shutdown mode 8.4+6 1.4+4 1.2+4 1.1+4 8.2+3 1.0+3 4.9+1 2.2-1 E12C002 RHR pump 1.4+5 1.4+4 1.2+4 1.1+4 8.2+3 1.0+3 4.9+1 2.2-1 E12C001 LPCS pump 1.4+5 1.4+4 1.2+4 1.1+4 8.2+3 1.0+3 4.9+1 2.2-1 E51C002 RCIC pump turbine 2.8+5 3.4+3 1.6+3 1.2+3 2.1+3 2.3+2 1.2+2 5.3+1 4.6+4 3.3+3 G33C001 RWCU pump 1.1+4 1.4+5 6.6+4 6.3+4 7.9+4 8.8+3 3.4+3 1.5+3 1.2+5 8.5+3 G40B0005 ADHR Heat Exchanger 9.2+6 5.0+1 4.1+3 1.1+4 1.5+5 B. REACTOR BUILDING G33B002 RWCU hx 2.5+5 1.4+5 6.6+4 5.3+4 7.9+4 8.8+3 3.4+3 1.5+3 2.2+4 1.6+3 G36A003 RWCU F/D bkwsh.
rec. tk 1.8+6 5.3+6 2.6+6 1.4+6 1.1+6 6.7+4 1.3+3 1.7+1 G36C001 RWCU F/D holding pump 1.1+4 1.4+5 6.6+4 5.3+4 7.9+4 8.8+3 3.4+3 1.5+3 2.2+4 1.6+3 G36D001 RWCU F/D 1.4+6 1.9+7 9.0+6 4.7+6 4.0+6 2.4+5 7.4+3 1.6+3 2.2+4 1.6+3 G50C012 RWCU bkwsh.
trans. pump 1.4+4 5.3+6 2.6+6 1.4+6 1.1+6 6.7+4 1.3+3 1.7+1 C. INTERMEDIATE BUILDING G41A002 Fuel pool surge tk. 2.8+7 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41A003 Fuel trans. tube drn. tk. 6.1+6 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41B001 Fuel pool HX 1.9+6 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41C001 Fuel pool F/D holding pump 7.0+3 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41C003 Fuel pool circ. pump 7.7+4 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41C004 Cask pool drn. pump 1.3+4 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41C005 Fuel trans. tube drn. pump 7.0+3 1.2+4 4.2+3 1.7+1 3.0+0 3.0+0 1.8-1 G41D001 Fuel pool F/D 9.5+6 4.5+5 3.3+5 1.9+5 1.5+5 6.6+3 G50A022 Fuel pool F/D bkwsh.
rec. tk. 2.6+7 4.5+5 3.3+5 1.9+5 1.5+5 6.6+3 G50C027 Fuel pool F/D bkwsh.
trans. pump 2.8+4 4.5+5 3.3+5 1.9+5 1.5+5 6.6+3 Revision 18 12.2-24 October, 2013
TABLE 12.2-5 (Continued)
Source Shielding Sources (/cc-sec)
Volume Equipment Identification (cc) 0.2 MeV 0.6 MeV 1.0 MeV 1.6 MeV 2.4 MeV 3.4 MeV 5.0 MeV 6.1 MeV 7.1 MeV D. RADWASTE BUILDING G50A001 Liquid waste coll. tk. 1.3+8 4.4+4 2.9+4 3.3+4 2.4+4 2.9+ 3 2.2+2 2.7+0 G50A002 Liquid waste sample tk. 1.3+8 4.4+4 2.9+4 3.3+4 2.4+4 2.9+ 3 2.2+2 2.7+0 G50A003 Floor drns. coll. tk. 1.3+8 4.4+4 2.9+4 3.3+4 2.4+4 2.9+ 3 2.2+2 2.7+0 G50A004 Floor drns. sample tk. 1.3+8 4.4+4 2.9+4 3.3+4 2.4+4 2.9+ 3 2.2+2 2.7+0 G50A005 Chemical waste tk. 8.8+7 5.5+4 4.9+4 2.4+4 2.0+4 7.5+ 2 4.5+1 6.7+1 G50A006 Concentrated waste tk. 1.9+7 2.3+6 2.0+6 1.0+6 8.4+3 3.1+ 1 1.9+3 2.8+1 G50A007 Chem. waste dist. tk. 7.5+7 5.6+0 5.0+0 2.5+0 2.1+0 G50A009 Spent resin tk. 3.3+7 3.9+6 1.9+6 2.2+6 9.3+5 4.1+ 4 9.6+2 6.6+0 G50A011 Cnds. F/D settling tk. 3.6+7 4.4+6 4.3+6 1.9+6 1.7+6 6.6+ 4 4.1+3 6.1+1 G50A013 RWCU settling tk. 5.9+6 1.9+7 9.0+6 4.7+6 4.0+6 2.4+ 5 7.9+3 1.6+3 G50A014 Waste sludge settling tk. 8.0+7 G50A024 Waste coll. filtrate tk. 1.5+6 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50A025 Floor drains, filtrate tk. 1.5+6 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50Z001 Waste evap. condenser 5.7+7 2.3+6 2.0+6 1.0+6 8.4+5 3.1+4 1.9+3 2.8+1 G50C001 Waste collector trans. pump 1.2+4 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C002 Waste sample pump 7.3+3 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C003 Floor drns. coll.
pump 1.2+4 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C004 Floor drns. sample pump 7.0+3 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C005 Chemical waste pump 7.0+5 5.5+4 4.9+4 2.4+4 2.0+4 7.5+2 4.5+1 6.7+1 G50C006 Chemical waste dist.
pump 1.2+4 5.6+0 5.0+0 2.5+0 2.1+0 G50C008 Spent resin pump 1.2+4 8.2+6 3.5+6 2.6+6 1.6+6 9.9+4 2.4+3 1.6+1 G50C010 Cond. sludge disch.
mix pump 1.2+4 1.0+6 9.6+5 4.3+5 3.9+5 1.5+4 9.3+2 1.4+1 G50C011 Cond. sludge decant pump 1.2+4 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C013 RWCU sludge disc.
mix pump 1.2+4 7.3+6 3.5+6 1.8+6 1.5+6 9.3+4 3.1+3 6.1+2 G50C014 RWCU sludge decant pump 2.9+3 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C015 Waste sludge disc.
mix pump 1.2+4 7.3+6 3.5+6 1.8+6 1.5+6 9.3+4 3.1+3 6.1+2 Revision 12 12.2-25 January, 2003
TABLE 12.2-5 (Continued)
Source Shielding Sources (/cc-sec)
Volume Equipment Identification (cc) 0.2 MeV 0.6 MeV 1.0 MeV 1.6 MeV 2.4 MeV 3.4 MeV 5.0 MeV 6.1 MeV 7.1 MeV D. RADWASTE BUILDING (Continued)
G50C016 Waste sludge decant pump 1.2+4 4.4+4 2.4+4 3.3+4 2.4+4 2.9+3 2.2+2 2.7+0 G50C017 Waste coll. filtrate pump 4.7+3 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 2.2+2 C50C018 Floor drns. filtrate pump 4.7+3 4.4+4 2.9+4 3.3+4 2.4+4 2.9+3 G50C026 Conc. waste trans.
pump 1.2+4 2.3+6 2.0+6 1.0+6 8.4+5 3.1+4 1.9+3 2.8+1 G50D001 Waste collector filter 7.1+5 2.0+7 8.4+6 6.4+6 3.4+6 2.4+5 5.7+3 3.9+1 G50D002 Floor drains filter 7.1+5 2.0+7 8.4+6 6.4+6 3.9+6 3.9+6 5.7+3 3.9+1 G50D003 Waste demin. 2.3+6 2.0+7 8.4+6 6.4+6 3.9+6 3.9+6 5.7+3 3.9+1 G50D004 Floor drns. demin. 2.3+6 2.0+7 8.4+6 6.4+6 3.9+6 3.9+6 5.7+3 3.9+1 E. TURBINE POWER COMPLEX N23D001 Condensate filter 5.0+5 4.4+6 4.3+6 1.9+6 1.7+6 6.6+4 4.1+3 6.1+1 N24A001 Conds. demin. cation regen. tk. 7.4+6 3.4+5 3.0+5 1.5+5 1.3+5 4.6+3 2.8+2 4.1+3 N24A002 Conds. demin. anion regen. tk. 3.7+6 3.4+5 3.0+5 1.5+5 1.3+5 4.6+3 2.8+2 4.1+3 N24A003 Conds. demin. mix &
hold tk. 7.4+6 7.6+1 3.3+2 5.2+2 1.5+2 1.3+1 3.5+0 5.4+2 N24A004 Conds. demin. bkwsh.
rec. tk. 1.9+7 3.4+5 3.0+5 1.5+5 1.3+5 4.6+3 2.8+2 4.1+3 N24A005 Conds. demin. regen.
chem. rec. tk. 4.5+7 5.5+4 4.9+4 2.4+4 2.0+4 7.5+2 4.5+1 6.7+1 N24C001 Waste transfer pump 7.3+3 5.5+4 4.9+4 2.4+4 2.0+4 7.5+2 4.5+1 6.7+1 N24D001 Condensate demin. 7.4+6 3.4+5 3.0+5 1.5+5 1.3+5 4.6+3 2.8+2 4.1+0 G50A010 Cnds. F/D bkwsh.
rec. tk. 2.0+7 1.1+5 1.1+5 4.7+4 4.3+4 1.6+3 1.0+2 4.1+0 G50C009 Cond. bkwsh. trans.
pump 2.8+4 1.1+5 1.1+5 4.7+4 4.3+4 1.6+3 1.0+2 1.5+0 F. TURBINE BUILDING N64B001 Offgas preheater 5.1+5 1.7+5 6.3+4 4.0+4 3.2+4 1.5+4 5.4+3 2.8+2 2.4+5 N64B002 Offgas condenser 9.6+5 3.8+6 1.9+6 1.1+6 8.7+5 4.5+5 9.9+4 5.8+4 1.4+6 N64D005 Offgas catalytic recombiner 2.4+6 4.1+4 1.6+4 9.8+3 7.8+3 3.8+3 1.3+3 8.2+1 5.3+4 Revision 12 12.2-26 January, 2003
TABLE 12.2-5 (Continued)
Source Shielding Sources (/cc-sec)
Volume Equipment Identification (cc) 0.2 MeV 0.6 MeV 1.0 MeV 1.6 MeV 2.4 MeV 3.4 MeV 5.0 MeV 6.1 MeV 7.1 MeV G. OFFGAS BUILDING N64B010 Offgas Cooler 1.2E+5 7.8+6 4.0+6 2.0+6 3.0+6 2.2+6 1.9+5 1.3+5 Condenser N64D011 Offgas Prefilter 2.7E+5 9.40E+5 1.06E+6 1.09E+6 1.19E+6 5.78E+5 1.19E+5 N64D012 Offgas Charcoal 7.2E+6 4.03E+6 1.39E+6 1.64E+5 7.7E+5 1.26E+6 5.50E+4 Absorber N64D016 Offgas After-filter 2.7E+5 6.40E+4 9.43E+3 9.90E-2 2.15E+3 9.55E+3 3.26E+0 N64D003 Gas Dryer 7.5E+5 1.71E+6 7.28E+5 1.49E+5 5.46E+5 4.43E+5 3.03E+4 NOTE:
(1)
Source model geometry for all calculations is cylindrical in shape.
Revision 12 12.2-27 January, 2003
TABLE 12.2-6 MATERIAL COMPOSITION OF THE TIP DETECTORS AND CABLES A. Detector Region Material Quantity 304L stainless (Co 0.014%) 4.92gm Titanium 0.662gm Alumina 0.885gm Nickel-iron alloy 0.248gm Copper 0.021gm B. Cable Region Material Quantity 304L stainless (Co 0.014%) 0.43gm/in.
AISI 1070 steel 2.16gm/in.
Magnesium oxide (Insulation) 0.0798gm/in.
Revision 12 12.2-28 January, 2003
TABLE 12.2-7 RADIATION LEVELS FROM THE TIP DETECTOR AND CABLES(1)
Decay Time _ Dose Rate, R/hr _
_ Days _ Detector(2) Cable(3) 0.0014 5.6 54.
0.0035 4.7 53.
0.021 3.7 47.
0.042 3.2 41.
0.083 2.4 32.
0.17 1.4 18.
0.50 0.17 2.2 1.00 0.013 0.10 2.00 0.0038 0.018 NOTES:
(1)
Based on three years of simulated use consisting of one hour of detector exposure in core semi-monthly.
(2)
At one meter.
(3)
At one meter from the midpoint of the 12 foot length of irradiated cable adjacent to the detector.
Revision 12 12.2-29 January, 2003
TABLE 12.2-8 TRAVERSING INCORE PROBE DETECTOR DECAY GAMMA ACTIVITIES OF MATERIALS IN THE DETECTOR(1)
Decay Time = 0 Seconds Activation Time = 102 Seconds Activated Activity (Isotope) __(Ci)_
Fe-59 1.1 + 1 Mn-56 1.7 + 5 Cr-51 7.0 + 1 Mn-54 2.1 + 0 Co-58M 3.5 + 3 Co-58 2.2 - 2 Ni-57 1.1 - 1 Co-57 6.0 - 7 Ni-65 4.0 + 2 Co-60M 7.6 + 3 Co-60 1.8 - 3 Co-61 9.6 + 0 Si-31 2.9 + 1 NOTE:
(1)
Excluding U-235.
Revision 12 12.2-30 January, 2003
TABLE 12.2-9 DECAY GAMMA ACTIVITIES OF MATERIALS IN THE CABLE Decay Time = 0 Seconds Activation Time = 102 Seconds Activated Activity
_Isotope_ (Ci/in)
Fe-59 8.2 + 0 Mn-56 7.4 + 4 Cr-51 3.7 + 0 Mn-54 1.6 + 0 Co-58M 1.0 + 2 Co-58 6.5 - 4 Ni-57 3.3 - 3 Co-57 1.8 - 8 Ni-65 1.2 + 1 Co-60M 2.2 + 2 Co-60 5.1 - 5 Co-61 2.8 - 1 Si-31 8.7 - 1 Revision 12 12.2-31 January, 2003
TABLE 12.2-10 TYPICAL TURBINE COMPONENT N-16 INVENTORIES Inventory System/Components _(Curies)
- 1. Main steam line and header system 263
- 2. High pressure turbine 6.4
- 3. Low pressure turbines (6 flow machine) 9.8
- 4. Moisture separator shell-side steam 53
- 5. Moisture separator shell-side liquid 41
- 6. Moisture separator drain system 56 (1)
- 7. First stage reheat system 33
- 8. Second stage reheat system(1) 32 (2)
- 9. First stage reheat drain system 1.4
- 10. Second stage reheat drain system(2) 1.1
- 11. Crossover pipe system 59
- 12. Crossaround pipe system 17
- 13. Feedwater heater and extraction system First Stage(3) 26 (3)
Second Stage 23 Third Stage(3) 27 (3)
Fourth Stage 15 Fifth Stage(4) 0.6 (5)
Sixth Stage 42 Condenser(6) 286 (7)
- 14. Hotwell 18
- 15. SJAE first stage system(8) 0.6
- 16. Recombiner system 0.4
- 17. Separate steam system(9) 0.9 (9)
- 18. Feedwater turbine system ___8.8_
Total 1,021 Revision 12 12.2-32 January, 2003
TABLE 12.2-10 (Continued)
NOTES:
(1)
Includes inventory in liquid and steam in reheat tubes and in steam supply line.
(2)
Includes total inventory beyond reheater outlet.
(3)
Includes total inventory beyond extraction point. Distribution of this will depend on equipment arrangement and sizing.
(4)
Excludes moisture separator drain system activity listed in Item 6.
(5)
Excludes first and second stage reheat drain system activities listed in Items 7, 8.
(6)
Excludes residual activity returned from feedwater turbine.
(7)
Excludes residual activity returned from feedwater turbine.
(8)
Includes inventory in steam supply system.
(9)
Includes total inventory beyond inlet at steam supply line.
Revision 12 12.2-33 January, 2003
TABLE 12.2-11 PARAMETERS AND ASSUMPTIONS USED IN CALCULATING REACTOR BUILDING AIRBORNE ACTIVITY Initial iodine partition factor .00086 in suppression pool Iodine halving time in 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> suppression pool Initial noble gas partition 0.5 factor in suppression pool Noble gas halving time in 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> suppression pool Suppression pool cleanup demineralizer 100 decontamination factor for iodine Revision 12 12.2-34 January, 2003
TABLE 12.2-12 REACTOR BUILDING AIRBORNE ACTIVITY Nuclide Concentration (Ci/cc)
Kr-85m 2.6-7 Kr-85 5.7-9 Kr-87 2.6-7 Kr-88 5.7-7 Kr-89 7.0-8 Xe-131m 3.5-9 Xe-133m 4.4-8 Xe-133 1.7-6 Xe-135m 7.0-8 Xe-135 1.6-6 Xe-137 9.8-8 Xe-138 2.2-7 I-131 1.1-9 I-132 4.8-10 I-133 3.8-9 I-134 2.6-10 I-135 2.0-9 Revision 12 12.2-35 January, 2003
TABLE 12.2-13 RADWASTE BUILDING AIRBORNE ACTIVITY Nuclide Concentration (Ci/cc)
I-131 1.4-11 I-132 1.5-10 I-133 1.0-10 I-134 2.2-10 I-135 1.5-10 Revision 12 12.2-36 January, 2003
TABLE 12.2-14 TURBINE BUILDING AIRBORNE ACTIVITY Nuclide Concentration (Ci/cc)
Kr-83m 3.9-9 Kr-85m 7.6-9 Kr-85 2.8-11 Kr-87 2.2-8 Kr-88 2.4-8 Kr-89 3.4-8 Xe-131m 1.9-11 Xe-133m 3.5-10 Xe-133 1.1-8 Xe-135m 1.9-8 Xe-135 2.8-8 Xe-137 4.7-8 Xe-138 6.7-8 I-131 1.0-11 I-132 1.2-10 I-133 7.5-11 I-134 2.0-11 I-135 1.2-10 Revision 12 12.2-37 January, 2003
TABLE 12.2-15 FUEL HANDLING AREA AIRBORNE ACTIVITY Nuclide Concentration (Ci/cc)
I-131 3.2-12 I-133 1.4-11 I-135 1.1-11 Revision 12 12.2-38 January, 2003
12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES The PNPP has been designed to attain as low as is reasonably achievable radiation doses to plant personnel as well as personnel located around the facility. The guidance of <Regulatory Guide 8.8> has been used in designing the facility to result in radiation doses that are only a small fraction of the limits given in <10 CFR 20>. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the
<10 CFR 20> regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20>
dated October 4, 1993.)
12.3.1.1 Equipment and Facility Design Features The plant restricted area is established in accordance with <10 CFR 20>
and is separated into five controlled access zones to aid in the design of radiation protection features and plant operation. An access zone designation is assigned to each area of the plant for each of two operating conditions: normal operation and shutdown/refueling.
<Figure 1.2-1> lists the Perry equipment names and numbers.
<Figure 12.3-1>, <Figure 12.3-2>, <Figure 12.3-3>, <Figure 12.3-4>,
<Figure 12.3-5>, <Figure 12.3-6>, <Figure 12.3-7>, <Figure 12.3-8>,
<Figure 12.3-9>, <Figure 12.3-10>, and <Figure 12.3-11> show the plant layout. These maps furnish design guidance for normal operating and shutdown plant dose rates. They provide the basis for decision making for locating and designing shielding and equipment in accordance with ALARA design principles. Design reviews include a review of actual in-plant conditions to verify that the intended design and installation Revision 14 12.3-1 October, 2005
is in keeping with ALARA principles and the design guidance provided in these figures. <Figure 12.3-1>, <Figure 12.3-2>, <Figure 12.3-3>,
<Figure 12.3-4>, <Figure 12.3-5>, <Figure 12.3-6>, <Figure 12.3-7>,
<Figure 12.3-8>, <Figure 12.3-9>, <Figure 12.3-10>, and <Figure 12.3-11>
include:
- a. Locations of the sources described in <Section 12.2> and
<Chapter 11>.
- b. Shield wall thicknesses.
- c. Design radiation zones for normal operation and refueling.
- d. Personnel and equipment decontamination areas.
- e. Locations of health physics facilities.
- f. Control panels for radioactive waste treatment equipment.
- g. Onsite laboratories for analysis of chemical and radioactive samples.
- h. Counting room.
These figures also illustrate locations of airborne radioactivity and area monitors.
The counting room is located so that the background radiation levels will be low enough to allow for continuous occupancy and to provide an accurate analytical environment under normal operating conditions and anticipated operational occurrences. The counting room is sized to provide adequate space for the required instrumentation. See
<Section 12.5> for a discussion of instrumentation in the counting room.
Revision 12 12.3-2 January, 2003
Nonradioactive equipment that may require maintenance is located, when possible, in either Zone I or Zone II. Adjacent areas containing potentially radioactive systems are designed to maintain a radiation level less than the Zone IV maximum (100 mrem/hr) during required maintenance.
Equipment located in Zone IV or Zone V is designed to minimize required maintenance and to be operated remotely. Shield wall penetrations for remote operating devices, electrical equipment, pipes, and ventilation ducts are designed and located at positions that prevent a direct line of sight to any significant source, thereby minimizing radiation streaming.
The primary defense against corrosion product buildup and associated neutron activation in the reactor vessel followed by crud transport is to minimize the input of impurities (i.e., iron, cobalt) in the feedwater. The Perry Plant design includes both full flow condensate filters and deep bed demineralizers. This design provides maximum removal of both suspended and dissolved impurities. In addition, an extensive condenser sampling and analysis system is provided to ensure prompt detection of small condenser leaks. Condensate demineralizers are provided to measure water quality in the bed effluent as described in <Section 10.4.6.2>.
The following design considerations have been given to reduce radiocobalt production and crud buildup in normally radioactive systems:
- a. System materials are specified for low corrosion and erosion rates and for low neutron activation source characteristics. Hardfacing materials which have high cobalt content, such as Stellite, are used only where substitute materials cannot satisfy performance requirements.
Revision 13 12.3-3 December, 2003
- b. Packless valves are specified for systems which normally handle radioactive fluids. Leakage of contaminated coolant from the primary system can be reduced by using live-loaded valve packings and/or bellow seals. Where packed valves are specified, as applicable, they are provided with positive backseats, lantern ring leakoffs to the liquid waste management system and special close tolerance graphoil packing in lieu of conventional packing to minimize leakage from the valve.
- c. System design considers decontamination of components. Isolation, vent and drain valves are provided in suitable locations to facilitate local decontamination of system components.
- d. Piping systems are of all welded construction with minimum use of flanged and socket weld connections.
Design practices will allow, whenever practicable, the separation of radioactive piping from nonradioactive piping, electrical equipment and personnel passageways.
The following examples illustrate specific design features that aid in minimizing exposure levels:
- a. Components containing radioactive materials will be separated, when practicable, to reduce radiation doses associated with maintenance.
- b. Cubicles are sized to provide adequate space around components for anticipated maintenance operations and for ease of entry and exit.
- c. The service mode of operation for the waste filters proceeds automatically after operator initiation from the radwaste building control room. Backwashing and precoating are done from a local control panel outside the filter cubicles.
Revision 22 12.3-4 October, 2021
- d. Both regenerable and nonregenerable demineralizers have provisions for remote removal of radioactive contents (spent resins or regenerative solutions) to the waste management systems.
Revision 22 12.3-4a October, 2021
- e. Activated carbon adsorber media can be removed from the filter plenums by a portable vacuum removal system. Adsorber media can be removed without entering the filter plenum.
- f. Particulate filters can be bagged during removal and sealed as the filters are removed from the plenums. Filters are covered with plastic during the entire change.
- g. Tanks containing potentially radioactive fluids are vented and can be drained to the waste management systems. Tanks containing fluids at atmospheric pressure are designed to withstand a pressure equivalent to a full tank of water. Static heads will be somewhat less due to overflow lines near the top of tanks.
- h. Evaporators have provisions for removal of noncondensibles to the waste management systems. Flush and rinse features permit decontamination before maintenance.
12.3.1.2 Illustrative Examples of Plant Design Features to Minimize Occupational Doses Plant design features represent a comprehensive effort to achieve minimization of radiation exposure. These features include:
- a. Radiation shielding of individual items of equipment.
- b. Accumulation of associated items of nonsafety-related equipment within contiguous areas of plant structures.
- c. Shielded chases for pipe runs between equipment cells and elsewhere about the plant.
- d. Other structural design relative to minimizing radiation exposure to operating and maintenance personnel.
Revision 12 12.3-5 January, 2003
Individual shielding means that the person approaching the location of a radioactive component is shielded from direct and most scattered radiation from both the item of equipment he is approaching (until he enters the equipment area) and all other items of radioactive equipment in the path to, and in the near vicinity of, the equipment being approached.
A semi-automated solid radwaste packaging and handling system has been included in the PNPP design to minimize the radiation doses associated with this routine operation. This system is discussed thoroughly in
<Section 11.4>. As mentioned in <Section 12.4>, the anticipated person-rem dose from waste processing for PNPP is only a small fraction of that being experienced in the industry at the time of submittal of the original FSAR.
The Mark III containment design includes a water filled suppression pool that provides the following functions:
- a. A heat sink for safety/relief valve (SRV) discharges,
- b. A heat sink for hot standby operation,
- c. A means to condense steam released to the drywell during a LOCA, and
- d. A continuing long term source of water for the emergency core cooling system.
The surface of the suppression pool is open to the containment so that some fractions of radionuclides discharged to the suppression pool from safety/relief valve operation and other sources can evolve into the containment atmosphere. Previous studies of the radiological consequences have concluded that the expected exposures of operation personnel are within the limits of <10 CFR 20>. (Radiological Revision 13 12.3-6 December, 2003
assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the
<10 CFR 20> regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised <10 CFR 20>
dated October 4, 1993.) However, there is a need to achieve exposures as low as reasonably achievable (ALARA) during normal operation and following normally expected transients.
There are occasional plant upset events that result in steam release from the SRVs. One such event is a complete depressurization of the reactor to the suppression pool following a power isolation transient.
During such a transient there may be sufficient increase in radioactivity within the containment to require egress of all personnel.
In an effort to reduce these operator exposures at the PNPP, the following change was made during the design process:
- a. Addition of a suppression pool cleanup system This system uses a mixed bed non-regenerative demineralizer (1,000 gpm). The main benefit of the system is to significantly reduce operator exposure to radioiodides which would evolve relatively slowly from the pool after the transient. In addition, the system will improve plant availability by allowing earlier operator re-entry without the use of respiratory equipment following a power isolation event.
Revision 17 12.3-7 October, 2011
12.3.2 SHIELDING 12.3.2.1 Design Objectives The design objectives of the plant radiation shielding are:
- 1. To ensure that during normal operation, including anticipated operational occurrences, the radiation dose to plant personnel and authorized site visitors is as low as reasonably achievable and within the limits set forth in <10 CFR 20>.
- 2. To provide the necessary protection for plant operating personnel following a reactor accident to maintain habitability of the control room as specified in <10 CFR 50.67>.
- 3. To limit offsite exposures to the general public to meet the dose requirements of <10 CFR 50.67> for postulated accident conditions and to maintain exposures as low as reasonably achievable, a small fraction of the <10 CFR 20> limits during normal operation.
- 4. To protect certain components from excessive radiation damage or activation.
12.3.2.2 Design Description 12.3.2.2.1 Plant Shielding Description Detailed layout drawings showing all plant structures are shown in
<Figure 1.2-3>, <Figure 1.2-4>, <Figure 1.2-5>, <Figure 1.2-6>,
<Figure 1.2-7>, <Figure 1.2-8>, <Figure 1.2-9>, <Figure 1.2-10>,
Revision 19 12.3-8 October, 2015
<Figure 1.2-11>, <Figure 1.2-12>, <Figure 1.2-13>, <Figure 1.2-14>,
<Figure 1.2-15>, <Figure 1.2-16>, and <Figure 1.2-17>. A general description of the major shielding in the buildings housing radioactive process equipment and fluids is outlined as follows:
- a. Reactor building complex The reactor building complex shielding includes the biological shield wall, drywell shield walls and the shield building wall.
The purpose of the biological shield is to minimize gamma heating in the drywell shield wall, to provide access to the drywell during shutdown and to reduce activation of drywell equipment and materials. The design dose rate used in sizing this shield is to maintain a radiation level in the drywell below 100 rem/hr at full power operation.
The drywell shield wall maintains the major area outside the drywell at Zone II level except for some individual cubicles housing radioactive process equipment and piping, such as cubicles for the reactor water cleanup system and the chase for the main steamline pipes. The shielding for these is sized to maintain a Zone II level outside of each respective cubicle.
Areas in containment routinely visited during power operation include the following systems: SLC(C41), RWCU(G33), CRD(C11), and TIP(C51). The expected occupancy requirements to these areas and the average personnel exposures is provided in- b. Turbine room and heater bay The major source of radiation in the turbine room and heater bay are the N-16 gammas. Shielding is provided around the radioactive equipment in the following systems to ensure that the dose levels are consistent with the access requirements:
- 1. Condensate and feedwater
- 2. Condensate filtration and demineralizer
- 3. High pressure heater drains and vents
- 4. Turbine Revision 12 12.3-10 January, 2003
- 5. Steam seal
- 6. Condenser and auxiliaries
- 7. Offgas Areas within the shield enclosures will normally be restricted access.
- c. Turbine skyshine The 016 (n, p) N16 reaction is of interest in the boiling water reactor because of the coolant activation induced by the high energy end of the fission neutron spectrum. The N16 present in the steam of a direct cycle boiling water reactor is carried with the steam into the turbine, moisture separators and associated equipment of the secondary cycle. Although the N16 decays with a half life of 7.35 seconds, the gamma emission can present a radiation dose problem to the site boundary as a result of the high energy gamma scatter from structures and atmosphere. Relative to this, turbine building gamma ray air scattering (skyshine) analyses with the Morse (Reference 1) Monte-Carlo code and the G-3 (Reference 2) code were made to evaluate the site boundary dose contribution from the N16 radiation. The Morse Code results, as presented in (Reference 3), were normalized in the G-3 code. The G-3 point kernel procedure which was then used to perform the site boundary dose rate calculations is based on application of the Klein Nishina scattering formula to the uncollided flux in a predetermined scattering grid. Normal scattering in air was approximated by use of water buildup factors on the scattered leg from each point in the scattered grid. This approach as well as previous work cited in literature has shown this method yields Revision 12 12.3-11 January, 2003
- 1. A three foot thick outer turbine building wall extending above the moisture separator reheaters.
- 2. A steel shield plate at the ends of the high and low pressure turbines with a labyrinth entry into the turbine area.
- 3. A 6-inch steel shield plate inboard of the moisture separator reheaters with labyrinth entry into the moisture separator reheater area.
- 4. A one foot thick concrete floor slab extending between the turbine and the 6-inch steel shield plate.
- d. Offgas building, auxiliary building, radwaste building, intermediate building, and fuel handling building.
- e. Control room The control room shielding is designed to maintain the dose requirement of 5 rem whole body or its equivalent to any organ for the duration of the accident as specified in <10 CFR 50, Appendix A>, Criterion 19 (for the design basis LOCA analysis the licensing basis limit is 5 rem TEDE). The analysis of the operator dose is presented in <Section 15.6.5>.
- a. Access labyrinths are provided for rooms housing equipment that contains high radiation sources to preclude a direct radiation path from the equipment to accessible areas.
- b. Piping penetrations, ducts and voids in radiation shield walls are located to preclude the possibility of streaming from a high to low radiation area, or otherwise will be adequately shielded.
- c. Shielding discontinuities caused by shield plugs, concrete hatch covers and shield doors to high radiation areas are provided with offsets to reduce radiation streaming.
- d. Radioactive piping is routed through high radiation areas where practicable, or in shielded pipe chases in low radiation areas.
- e. Sufficient work area and clearance space is provided around equipment to permit ease of servicing.
- f. Instruments requiring in situ calibration will not normally be located in high radiation areas.
- g. Non-radioactive equipment which requires servicing will not normally be located in proximity with potentially radioactive equipment.
- h. Spread of contamination from radioactive spillage is minimized by providing a floor drain system which collects and routes the liquid to the liquid waste processing system for proper handling.
- i. Natural traps which could be potential pockets of corrosion product activity are minimized in pipe and ducts by avoiding sharp bends, rough finishes, cracks, etc.
- j. Shielding is provided for all equipment which is anticipated to be normally radioactive. The dose levels are designed not to exceed
- k. Temporary shielding such as lead blankets, will be available on the site in the event it is ever needed.
- l. Remote handling of radioactive materials is provided wherever it is needed and practicable.
- m. The guidance of <Regulatory Guide 1.69> has been used with regard to design of concrete radiation shields.
- a. SDC (Reference 4)
- b. QAD6G (Reference 5)
- c. G-3B (Reference 2)
- d. ANISIN (Reference 6)
- e. Microshield (Reference 7)
- f. Microskyshine (Reference 8)
- g. Monte Carlo N-Particle Version 5 (Reference 11)
- a. Maintain the required ambient air temperature to prevent extreme thermal environmental conditions for operating personnel and equipment.
- b. Protect the operating personnel against possible airborne radioactive contamination in areas where this may occur.
- c. Ensure that maximum airborne radioactivity levels for normal and emergency operations, including anticipated operational occurrences, are within the limits of <10 CFR 20, Appendix B>, for areas within plant structures and the restricted areas on the plant site where construction workers and visitors are permitted.
- d. Provide a suitable environment for continuous personnel occupancy in the control room under normal and postaccident conditions in accordance with <10 CFR 50, Appendix A>, Criterion 19.
- a. Air movement patterns are generally from areas of lesser radioactive contamination to areas of progressively greater radioactive contamination prior to final exhaust.
- b. Slightly negative pressures are maintained in specific areas such as the annulus between the containment and shield building to prevent uncontrolled release of contamination. The control room is maintained at slightly positive pressure during normal operation to prevent infiltration of potential contaminants.
- c. Valves and equipment are as leak tight as practicable to prevent leakage of radioactive fluids and subsequent airborne contamination.
- d. Individual air supplies are provided for each building to keep potentially contaminated air flows separate from noncontaminated air.
- e. In general, potentially radioactive air is exhausted through filter trains consisting of roughing, HEPA and charcoal filters to reduce onsite and offsite radiation levels. Filtered and monitored exhausts are provided in all buildings that could potentially Revision 12 12.3-20 January, 2003
- f. Roughing, HEPA and charcoal filters are used for filtration of the recirculated air of the control room and associated offices during accident and other abnormal conditions.
- g. Radiation exposures will be kept as low as practicable while servicing ventilation equipment by the following provisions (incorporated in the plant design):
- 1. The ventilation equipment is not located in normally high radioactive areas.
- 2. Suitable access doors and service aisles are provided to permit ease in servicing and maintenance.
- 3. The roughing and HEPA filters in the ESF filter trains and the HEPA filters in the non-ESF filter trains are serviced on the downstream side of the filter to minimize personnel exposure.
- 4. The activated charcoal adsorber is bulk loaded into the permanently installed, seal welded and gasketless adsorber section with the exception of the Intermediate Building Sub-Exhaust plenum which utilizes tray type adsorber cells.
- h. Control of airborne contaminants during maintenance operations is accomplished by the ventilation system, in accordance with the requirements of <10 CFR 20> as follows:
- 1. Equipment redundancy is provided where practicable and idle equipment is isolated by dampers so that these components can be serviced without disrupting the operation of the system.
- 2. Ventilation is accomplished in potentially contaminated areas by supplying air to the clean areas (corridors) and drawing it into the rooms through doorways and/or wall openings so that the air flow is normally from clean to contaminated areas.
- i. Ventilation ducts enter potentially radioactive areas at least 7 feet above the floor where practical. This minimizes direct exposure to personnel from radiation streaming. Penetration criteria are discussed in <Section 12.3.2>.
- j. Ventilation equipment and ductwork are located in low exposure areas where practical. Therefore, automatic dampers, manual dampers and fire dampers can be maintained with minimum exposure to radiation. In potentially high radiation areas, only manual balancing dampers on the inlet and outlet registers require attention. These were adjusted and set in position prior to plant operation and require little further maintenance or adjustment.
- k. Interior surfaces of ducts are designed to minimize the buildup of dust. Shop made duct joints are welded and field joints are gasketed with bolted connections. Ductwork joints, therefore, do not have gaps where dust could settle.
- l. Access panels allow cleaning and inspection of ductwork. When ductwork needs cleaning, vacuum cleaning should usually be adequate. Where practical, ductwork is accessible for service and maintenance.
- m. Exhaust air from vacuum cleaning or special ventilation can be ducted to an inlet connection in the building exhaust air system in most cases. In radioactive waste areas and the charcoal cleaning plenum areas, inlet connections are close to the expected cleaning equipment location. This minimizes extensive lengths of temporary ducting.
- a. Provide plant personnel with a system that will indicate that the radiation levels are below those requiring special monitoring equipment.
- b. Provide a system which can aid in minimizing personnel exposure to radiation and maintain occupational radiation exposure ALARA.
- c. Provide instrumentation for the reactor operator to monitor selected plant area gamma radiation levels following an incident, thereby enhancing his ability to determine the nature and extent of the incident.
- d. Augment and supplement other monitoring systems, such as the leak detection system and the airborne radiation monitoring system, in the detection of incidents involving release of radioactive material.
- e. Provide alarms (alert and high radiation) to warn personnel when the gamma radiation level of selected areas increases substantially.
- f. Provide a record of radiation levels as a function of time at key strategic areas within the plant.
- g. Provide the reactor operator with alert, high radiation and circuit failure alarms for each channel.
- h. Aid the operator in personnel deployment decisions following an incident involving high radiation.
- i. Assist in the detection of unauthorized or inadvertent movement of radioactive material in the plant.
- j. Warn of excessive gamma radiation levels in areas where special nuclear material is stored or handled.
- k. Warn personnel of high radiation in areas prior to entry.
- a. Detectors The detector is a gamma sensitive device housed in a sealed container separated from the local alarm indicator unit. A halogen quenched Geiger-Mueller tube is coupled with a preamplifier to convert the incident gamma radiation into an electrical signal which is transmitted to the readout module. Design information for the detector is listed in
- b. Local Alarm Indicator Unit All channels, except the control room channel (1D21N400) and the local channels (1D21N340, D21N380, and Revision 19 12.3-31 October, 2015
- 1. A readout meter.
- 2. A high radiation alarm light visible on a 180 degree horizontal azimuth from the wall and protected by a watertight red glass cover and a metal cage. The light receives 120-volt ac, 60-Hertz, electrical power from the readout module.
- 3. An alert alarm light visible on a 180 degree horizontal azimuth from the wall and protected by a watertight amber glass cover and a metal cage. The light receives 120-volt ac, 60-Hertz, electrical power from the readout module.
- 4. A klaxon horn providing an audible alarm in conjunction with the high radiation alarm. The horn receives 120-volt ac, 60-Hertz, electrical power from the readout module. The audible alarm is capable of being silenced at the associated readout module in the control room.
- c. Remote Warning Unit Channels 1D21N030, 1D21N060 and 1D21N080, have remote warning units in addition to alarm indicator units. The remote warning units are identical to the alarm indicator unit except that no audible alarm is required.
- d. Readout Module The readout module contains most of the electronic circuitry for system operation. The module consists of compact, solid state circuitry, a modular design which allows up to 3 modules to be arranged side-by-side in a rack mounted chassis located in the area radiation monitoring panel. The module contains alarm circuitry, a functional control switch and signal processing amplifiers for dose rate indication. Each module contains an independently fused regulated power supply. The modular compact design allows removal of the module from the chassis for replacement or repair. Circuit alignment can be accomplished while the system is energized.
- 1. Three and one-half inch meter with meter range of 0.1 mR/hr to 104 mR/hr.
- 2. High level alarm lamp.
- 3. Alert alarm lamp.
- 4. Failure alarm lamp.
- 5. Switching capability with the following functions:
- e. Recorders Multi-point strip chart recorders, located on the area radiation monitoring recorder panels, provide a permanent record of the Revision 12 12.3-34 January, 2003
- f. Power Supply The area radiation monitoring system channels utilizing control from the centralized control room panels receive electrical power from two sources. The 120-volt ac power to the readout module electronics is supplied from non-Class 1E ac instrument bus which is emergency diesel generator backed through a transfer switch (except during a LOCA). The second power source is used to supply 120-volt ac power for all horns and alarm lamps. This power is supplied from a miscellaneous 120-volt ac distribution panel and is not diesel backed.
- g. System Setpoints The alarm setpoints which are not controlled by other licensing requirements are adjustable and are set and revised as necessary based on operational experience gained throughout plant maturation.
- h. Calibration Area radiation monitors are to be calibrated on a routine basis and after any major maintenance work is performed on the detector or its associated ratemeter. Detector calibration is obtained by Revision 19 12.3-35 October, 2015
- a. Furnish quantitative information (based on representative sampling) to the reactor operator and to operations personnel on the level of airborne radioactivity in plant ventilation systems and selected areas of the plant.
- b. Provide a system which can aid in minimizing personnel exposure to airborne radioactivity and maintain occupational radiation exposure ALARA.
- c. Furnish information to substantiate radiation surveys as required by <10 CFR 20>, and provide supporting documentation of working environments.
- d. Provide instrumentation for the reactor operator to monitor plant ventilation systems, and selected areas of the plant for level of radioactivity during and following an incident, thereby enhancing the ability to determine the nature and extent of the incident.
- e. Supplement the leak detection system in detecting leakage from the reactor coolant pressure boundary.
- f. Provide overall plant monitoring of airborne radioactivity and reasonable assurance that the ambient airborne radiation levels are below those requiring special monitoring equipment.
- g. Supplement other monitoring systems, such as the area radiation monitoring system, in the detection of incidents involving release of radioactive material.
- h. Aid in the protection of the plant personnel from exposure to airborne radioactive materials in excess of the levels allowed by
- i. Provide the reactor operator with alarms for each channel (alert, high radiation or channel failure) and alarms for each subsystem (sample flow low).
- j. Provide the operations staff with a hard copy record of radioactivity levels in the monitored systems.
- k. Continuously monitor the plant ventilation systems for airborne radioactivity in order to permit an assessment to be made of the radiological hazards to be encountered within various regions of plant buildings, and to call attention to equipment malfunction or component failure resulting in the release of radioactivity.
- l. Provide instrumentation for use as the basis for initiating actions related to the plant radiation emergency plan.
- m. Provide instruments of sufficient range so as to monitor the radioactivity levels postulated for accident events.
- a. Equipment located outside the control room is housed in NEMA Type 12 ventilated enclosures.
- b. Setpoint adjustment devices are protected to prevent inadvertent operation.
- c. Readout modules are accessible for test, alignment, changing setpoints, and calibration or inspection without interrupting power to the module.
- d. The detectors used in the airborne radiation monitoring system are scintillation detectors. The entire detector assembly is built into a housing which serves as a shield against changes due to light photons or electrostatic or magnetic fields. The housing extends entirely over the base of the assembly, making a single unit. The detector assembly is covered by 4-Pi shielding. The Revision 12 12.3-40 January, 2003
- e. The detector preamplifier is contained within the detector housing.
- f. High and alert radiation level alarm trip setpoints are adjustable over the entire range at the readout module.
- g. Equipment is designed to be capable of withstanding an integrated gamma dose of 104 rads.
- h. For analog readout modules remote actuated check sources provided with the detector assembly are Cesium-137 for beta scintillation detectors, and Barium-133 for gamma scintillation detectors. For digital readout modules remote LED pulsers are provided with the detector assembly for both beta and gamma scintillation detectors.
- a. Sensitivity of particulate channels equipped with fixed filter:
- b. Sensitivity of particulate channels equipped with moving filter:
- c. Sensitivity of iodine channels:
- d. Sensitivity of gas channels:
- a. Particulate monitoring channel:
- b. Iodine monitoring channel:
- c. Gas monitoring channel:
- a. The readout module contains most of the electronic circuitry for system operation. The module consists of compact, solid state circuitry and modular design which provides for modules to be arranged side by side in a standard 19-inch rack mounted chassis.
- b. The analog readout module has a time constant which is inversely proportional to the count rate with the probable statistical error E less than 15 percent:
- c. Each readout module contains its own independent fused regulated power supply suitable for 120-volt ac, 60-Hertz power input.
- d. Each readout module has provision for determining voltages essential for proper channel operation.
- e. The readout modules have the following features (front panel):
- 1. Range: 10 to 106 cpm logarithmic 10 to 107 cpm digital, for digital readout modules
- 2. Meter Size: 4-1/2; 2% full scale accuracy approximately 3 wide; 1% full scale accuracy (1 digit), for digital readout modules Revision 19 12.3-44 October, 2015
- 3. Meter Scales: 10 to 106 cpm High voltage 500 to 2,500 volts dc; Calibration check point.
- 4. Alarm Lamps:
- 5. Switching capability with the following functions:
- 6. Iodine channel readout modules have a single channel analyzer circuit to provide energy discrimination for selectively monitoring the 364 KeV I131 photopeak. These modules are used in conjunction with the iodine sampling subassemblies.
- f. The readout module has the following features:
- 1. Independent regulated power supply. (No independent regulated power supply provided within digital readout modules)
- 2. High level alarm contacts (nonlatching).
- 3. Alert alarm contacts (nonlatching).
- 4. Failure alarm contacts (nonlatching).
- 5. Buffered output for computer (0 to 10 volts dc isolated, positive signal).
- 6. Output for recorder (0 to 10 mV dc).
- 7. Output for remote readout meter(s).
- 8. Fixed circuit failure alarm setpoint.
- 9. Alert level alarm setpoint adjustment (variable over full range).
- 10. High level alarm setpoint adjustment (variable over full range).
- 11. Test points (MCA and signal generator jacks). (No MCA jacks provided on digital readout modules)
- g. Special features:
- 1. Both the high level and the alert level alarm lamps flash in an alarm condition. These lamps change from the flashing state to a steady on state when the module alarm acknowledge button is depressed.
- 2. Switching capability for alarm setpoint trip adjustment is provided such that, when operated, the meter will indicate the alarm setpoint and the alarm trip circuitry will actuate.
- 3. Switching capability for check source operation is provided such that, when the pushbutton is depressed, a radioactive source is actuated at the detector assembly to provide a response check of the channel. For digital readout modules, on LED pulser is actuated at the detectors photomultiplier tube to generate a signal for the response check of the channel. In addition, the circuitry incorporates an alarm defeat provision such that when the check source is actuated, the alert and high level alarm will not actuate. The check source returns to the retracted position upon loss of power.
- 4. The failure alarm will actuate upon loss of detector voltage, loss of line voltage, loss of amplifier output signal, or loss of detector input signal.
- 5. The readout module alert level alarm circuit and high level alarm circuit will be manually reset when the initiating signal returns to normal. The failure alarm circuit will automatically reset when the initiating signal returns to normal.
- 6. Alarm outputs:
- 7. An accessible test point or connector is provided for input of a signal generator to the circuit for calibrating the alarm trip setpoint circuit. Capability is provided for disconnection of the detector input signal during pulse generator test. In addition, for the analog readout modules, a test point or connector is provided for scaler readout at a point after the discriminator circuit.
- 8. Each analog readout module has two separate alert and two separate high level setpoint adjustment circuits, either setpoint of which may be inserted into the alarm circuit.
- 9. Analog readout modules are provided with four connectors for termination of wiring on the module. One is provided for termination of electrical power supplied to the module, one provided for termination of field wiring (excluding high voltage), one for termination of high voltage, and one is used for termination of module output circuits (which includes alarm outputs, recorder and computer signals). Digital readout modules are provided with six connectors for termination of wiring on the module.
- 10. Power supply:
- 11. Calibration:
- a. Reactor Building:
- 1. Drywell Atmospheric Radiation Monitor (1D17K670)
- 2. Containment Atmospheric Radiation Monitor (1D17K680)
- 3. (Deleted)
- 4. Containment Vessel and Drywell Purge Exhaust Radiation Monitor (1D17K660)
- b. Radwaste Building:
- c. Auxiliary Building:
- d. Control Complex:
- e. Intermediate Building:
- f. Fuel Handling Building:
- g. (Deleted)
- h. Offgas Building:
- a. Assume a constant one MPC (or DAC equivalent) level of either Cs137, I131 or Kr85 in any subcompartment during normal operation.
- b. Exhaust flowrate from the subcompartment.
- c. Air dilution factor relative to the airborne radiation monitor sampling point is:
- d. Typical sensitivity of airborne radiation monitor as referenced in
- e. Airborne radioactivity concentration at radiation monitor sampling point.
- a. Reactor Building:
- 1. The containment vessel and drywell purge radiation monitor can detect 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) level of Cs137 in any reactor building subcompartment listed in
- 2. The containment vessel and drywell purge monitor can detect during purge a noble gas concentration of 1 MPC (0.1 DAC) in the drywell (drywell at purge); containment pool area (refueling operation); containment free space, and the RWCU HX area (normal operation).
- 3. In the following locations, a noble gas concentration of 10 MPC (1 DAC) can be detected by the containment vessel and drywell purge exhaust radiation monitor:
- 4. With the use of the containment atmospheric radiation monitor (D17K680), sufficient radiological surveillance is available for the information of personnel in the reactor building.
- 5. During refueling operations when the reactor is opened, radioactive substances from the reactor coolant may locally contaminate the air and not be detectable for some time on the exhaust monitor. Radiation Protection performs localized surveys as deemed necessary to ensure airborne radioactivity is controlled per <10 CFR 20>.
- b. Radwaste Building:
- 1. The radwaste building ventilation exhaust radiation monitor can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 in any radwaste building subcompartment listed in
- 2. The radwaste building ventilation exhaust radiation monitor can detect a 10 MPC (1 DAC) level of noble gas in any radwaste building subcompartment listed in
- c. Auxiliary Building:
- 1. The auxiliary building ventilation exhaust radiation monitor can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 in any building subcompartment listed in
- 2. The auxiliary building ventilation exhaust radiation monitor can detect a 1 MPC (0.1 DAC) level of noble gas in building subcompartments as listed in
- d. Intermediate Building:
- e. Fuel Handling Area:
- 1. The fuel handling area ventilation exhaust radiation monitoring system can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 in any subcompartment listed in
- 2. The fuel handling area ventilation exhaust radiation monitors can detect a 1 MPC (0.1 DAC) level of noble gas in the cask storage pool, spent fuel pool and fuel transfer pool area as listed in
- f. Heater Bay:
- 1. The heater bay ventilation exhaust radiation monitors can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 on any floor level in the building
- 3. A 1 MPC (0.1 DAC) level of noble gas at Elevation 620-6 can be detected in summer and less than 15 MPC (1.5 DAC) in winter operation due to differences in the dilution factors. Less than 15 MPC (1.5 DAC) can be detected in the FDW Lube Oil purifier room in summer or winter operation.
- 4. A 25 MPC (2.5 DAC) level of noble gas at Elevation 647-6 hallway can be detected in winter operation and less than 2 MPC (0.2 DAC) can be detected in summer operation.
- g. Turbine Building:
- 1. The turbine building ventilation exhaust radiation monitor can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 in all three areas of interest in
- 2. A noble gas concentration of 1 MPC (0.1 DAC) can be detected by the exhaust radiation monitors in all subcompartments except in the condenser vacuum pump and sample extraction areas where less than 25 MPC (2.5 DAC) can be detected.
- h. Offgas Building:
- 1. The offgas building ventilation exhaust radiation monitor can detect a 1 MPC (0.45 DAC) level of I131 or 1 MPC (1 DAC) of Cs137 in any subcompartment listed in
- a. High radiation areas, where whole body dose levels may exceed 100 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 cm, but less than 1,000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area.
- b. Air flow patterns are consistent with the basic ventilation design criteria of the plant. Clean, filtered outside air is supplied to continuous occupancy areas (corridors, clean areas); these areas are exhausted into rooms and areas of successively higher potential for airborne contamination. Air flow is such that reversal or exfiltration from potentially contaminated areas is normally precluded. This ventilation arrangement essentially eliminates the possibility of personnel exposure to airborne radioactivity in continuous occupancy areas <Section 12.3.3>.
- c. Administrative and physical controls are provided for access to areas in which potential sources of hazardous levels of airborne radioactivity from piping and equipment are located. Access shall require a Radiation Work Permit should airborne activity levels warrant one.
- d. Maintenance, in a radiologically restricted area, requires a RWP.
- e. Radiation protection programs are discussed in <Section 12.5>.
- a. Auxiliary Building Ventilation Exhaust:
- b. Radwaste Building Ventilation Exhaust:
- c. Reactor Building Purge Exhaust:
- d. Fuel Handling Building Exhaust:
3.5 REFERENCES
FOR SECTION 12.3
- 1. E.A. Straker, et al., The Morse Code, A Multigroup Neutron and Gamma Rate Monte Carlo Transport Code, ORNL-4585, Oak Ridge National Laboratory (1970).
- 2. G-3, General Purpose Gamma Ray Scattering Code, Rev. 1A, GAI Code N105.
- 3. W. A. Woolson, et al., Calculation of the Dose at Site Boundaries from N-16 Radiation in Plant Components, JRB 72 5076J, December 8, 1972.
- 4. Arnold, E. D. and Maskewitz, B. F., SDC - A Shield Design Calculation Code for Fuel Handling Facilities, ORNL-3041, March 1966.
- 5. Malenfont, R. E., QAD, A Series of Point-Kernel General Purpose Shielding Programs, Los Alamos Scientific Laboratory Report No. 3573, April 5, 1967.
- 6. Boling, M. A. and W. A. Rhoads, ANISIN/DT FII, A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering.
AL-66-MEMD-171.
- 7. Grove Engineering, Inc., Microshield Users Manual, Version 2.0.
- 8. Grove Engineering, Inc., Microskyshine Users Manual, Version 1.10.
- 9. NRC Safety Evaluation Report dated June 28, 1994 accompanying Amendment 62 to the Technical Specification.
- 10. <10 CFR 50.68>
Revision 12 12.3-68 January, 2003
- 11. Los Alamos National Laboratory, Monte Carlo N-Particle Version 5 (MCNP5), Version 1.60 Revision 19 12.3-68a October, 2015
TABLE 12.3-1 RADIATION ZONE DESIGNATIONS AND CODE (DELETED)
Revision 14 12.3-69 October, 2005
TABLE 12.3-2 RADIATION SHIELD THICKNESSES Shield Thickness Equipment Identification (ft of concrete)
- a. AUXILIARY BUILDING E12B001 RHR heat exchanger 2 E12C002 RHR pump 2 E21C001 LPCS pump 2 E22C001 HPCS pump 2 E51C001 RCIC pump 2 G33C001 RWCU pump 3 G40B005 ADHR heat exchanger 1-1/2(1)
- b. REACTOR BUILDING G33B001 RWCU heat exchanger (regen.) 4 G33B002 RWCU heat exchanger (non-regen.) 4 G36A003 RWCU fill and drain backwash receiving tank 3 G36C001 RWCU fill and drain holding pump 2 G36D001 RWCU fill and drain 3-1/2 G50C012 RWCU backwash transfer pump 2
- c. INTERMEDIATE BUILDING G41A002 Fuel pool surge tank 1 G41A003 Fuel transfer tube drain tank 2 G41B001 Fuel pool heat exchanger 2 G41C003 Fuel pool circulating pump 2 G41C005 Fuel transfer tube drain pump 2 G41D001 Fuel pool fill and drain 2-1/2 G50A022 Fuel pool fill and drain backwash receiving tank 2-1/2 G50C027 Fuel pool fill and drain transfer pump 2-1/2
- d. RADWASTE BUILDING G50A001 Liquid waste collection tank 2 G50A002 Liquid waste sample tank 2 G50A003 Floor drains collection tank 2 G50A004 Floor drains sample tank 2 G50A005 Chemical waste tank 2 G50A006 Concentrated waste tank 2-1/2 G50A007 Chemical waste distribution tank 1 G50A009 Spent resin tank 3 G50A011 Condensate fill and drain settling tank 3 G50A013 RWCU settling tank 3 NOTE:
(1)
The ADHR heat exchanger is shielded by a partial height concrete shield wall supplemented by 5-6 layers of 15 psf lead blankets above the concrete wall.
Revision 18 12.3-70 October, 2013
TABLE 12.3-2 (Continued)
Shield Thickness Equipment Identification (ft of concrete)
- d. RADWASTE BUILDING (Continued)
G50A014 Waste sludge settling tank 2 G50A024 Waste collection filtrate tank 2 G50A025 Floor drains filtrate tank 2 G50C001 Waste collector transfer pump 1 G50C002 Waste sample pump 1 G50C003 Floor drains collector pump 1 G50C004 Floor drains sample pump 1 G50C005 Chemical waste pump 1 G50C006 Chemical waste distribution pump 1 G50C008 Spent resin pump 2 G50C010 Condensate sludge discharge mix pump 1 G50C011 Condensate sludge decant pump 1 G50C013 RWCU sludge discharge mix pump 2 G50C014 RWCU sludge decant pump 1 G50C015 Waste sludge discharge mix pump 1 G50C016 Waste sludge decant pump 1 G50C017 Waste collection filtrate pump 1 G50C018 Floor drains filtrate pump 1 G50C026 Conc. waste transfer pump 2 G50D001 Waste collector filter 2 G50D002 Floor drains filter 2 G50D003 Waste demineralizer 2-1/2 G50D004 Floor drains demineralizer 2-1/2 G50Z001 Waste evaporator cond. 2-1/2
- e. TURBINE POWER COMPLEX N23D001 Condensate filter 2-1/2 N24A001 Condensate demineralizer cation regen. tank 2-1/2 N24A002 Condensate demineralizer anion regen. tank 2-1/2 N24A003 Condensate demineralizer mix and hold tank 2-1/2 N24D001 Condensate demineralizer 2-1/2
- f. OFFGAS BUILDING N64B010 Offgas cooler condenser 3 N64D011 Offgas prefilter 3 N64D012 Offgas charcoal absorber 3 N64D016 Offgas after-filter 1 N64D030 Offgas desiccant dryer 3
- g. TURBINE ROOM N25B001 Moisture separator reheater 3 N31C001 Main turbine 3 N64B001 Offgas preheater 4 N64B002 Offgas condenser 4 N64D005 Offgas catalytic recombiner 4 Revision 18 12.3-71 October, 2013
TABLE 12.3-2 (Continued)
Shield Thickness Equipment Identification (ft of concrete)
- h. HEATER BAY N21B003 Compressure feedwater heater no. 3 2 N21B004 Direct contact heater 2-1/2 N27B001 Intermediate pressure feedwater heater no. 5 2 N27B002 High pressure feedwater heater no. 6 2 N27B003 Heater no. 5 drain cooler 1-1/2 N27C003 Main feedwater pump drive turbine 2(1)
N33B002 Steam seal evaporator 2 NOTE:
(1)
Except at knockout opening in north wall of the A feedpump turbine room due to the distance from sources in the room. No shielding is required at this opening to meet radiation zoning criteria on
<Figure 12.3-4>. This exception does not apply to the B feedpump turbine room.
Revision 12 12.3-72 January, 2003
TABLE 12.3-3 COMPARISON OF NON-ESF CHARCOAL FILTER SYSTEMS TO
<REGULATORY GUIDE 1.140> CRITERIA Regulatory Position System Design Feature 1.a The design conforms with this position.
1.b The design conforms with this position.
1.c The design conforms with this position.
1.d The design conforms with this position.
2.a All non-ESF charcoal plenums are Seismic Category I and consist of the following components in sequence: prefilters, HEPA charcoal and HEPA.
Fans, ducts, dampers, and related instrumentation are also provided.
2.b The design conforms with this position except that HEPA filter arrangements are not always 3 high by 10 wide.
2.c The design conforms with the intent of Section 5.6 of ERDA 76-21.
2.d The design conforms with this position.
2.e The design conforms with this position.
2.f The design conforms with this position. Duct and housing leak tests will be performed in accordance with Section 6 of ANSI N510-1980 instead of ANSI N510-1975.
3.a The relative humidity of exhaust air for the non-ESF charcoal filter systems is not expected to exceed 70 percent.
3.b The design conforms with this position.
3.c The design conforms with this position.
3.d The design conforms with the intent of Section 4.4 of ERDA 76-21.
3.e The design conforms with this position.
Revision 12 12.3-73 January, 2003
TABLE 12.3-3 (Continued)
Regulatory Position System Design Feature 3.f The design conforms with this position.
3.g The design conforms with this position. The original or replacement batch of impregnated activated carbon shall meet the qualification and batch test results summarized in Table 5-1 of ANSI N509-1980, which meets or exceeds the requirements of Table 5-1 of this Regulatory Guide.
3.h The design conforms with this position. The adsorbent shall meet the requirements of Table 5-1 of ANSI N509-1980, which meets or exceeds the requirements of Table 5-1 of ANSI N509-1976.
3.i The design conforms with this position.
3.j The design conforms with this position.
3.k The design conforms with this position.
3.l The design conforms with this position.
4.a The design conforms with the intent of the recommendations of Section 2.3.8 of ERDA 76-21 and Section 4.7 of ANSI N509-1976.
4.b The maximum length of component plus 2 ft 6 inches is provided due to space limitations imposed by equipment room size. It was determined that this is adequate for the replacement of the prefilters and HEPA filters, and is consistent with the manufacturers recommendation.
4.c The design conforms with this position.
4.d Preoperational Phase Testing meets the intent of this position. The testing will be performed while active construction is in progress on the project, but sufficiently complete to ensure that the installed HEPA filters and charcoal are not subjected to airflow that would invalidate inplace testing.
Revision 12 12.3-74 January, 2003
TABLE 12.3-3 (Continued)
Regulatory Position System Design Feature 5.a Testing procedures will meet the intent of this position. Visual inspection will be performed in accordance with the provisions of Section 5 of ANSI N510-1980 instead of ANSI N510-1975.
5.b Testing procedures will meet the intent of this position. The airflow distribution testing will be performed in accordance with the provisions of Section 8.3.2 of ANSI N510-1980 instead of ANSI N510-1975.
5.c Testing procedures will meet the intent of this position. The inplace leak test on upstream HEPA filter banks will be performed in accordance with the provisions of Section 10 of ANSI N510-1980 instead of ANSI N510 1975. In addition Section 10 of ANSI N510-1980 requires Sections 8 and 9 to be performed as prerequisites to the inplace leak test on the HEPA filter bank and the Section 8 and 9 testing will be performed in accordance with the provisions of Sections 8 and 9 of ANSI N510-1980.
Inplace leak testing on downstream HEPA filter banks will not be performed. Testing frequency will meet the intent of the provision but may be based upon refueling outage intervals for systems M14 and M38.
5.d Testing procedures will meet the intent of this position. The inplace leak test on the charcoal adsorber stage will be performed in accordance with the provisions of Section 12 of ANSI N510-1980 instead of ANSI N510-1975. In addition Section 12 of ANSI N510-1980 requires Sections 8 and 9 be performed as prerequisites to the inplace leak test on the charcoal adsorber stage and the Section 8 and 9 testing will be performed in accordance with the provisions of Section 8 and 9 of ANSI N510-1980.
Testing frequency will meet the intent of the provision but may be based upon refueling outage intervals for systems M14 and M38.
6.a(1) Testing procedures will meet the intent of this position per 5.d above.
Revision 12 12.3-75 January, 2003
TABLE 12.3-3 (Continued)
Regulatory Position System Design Feature 6.a(2) Initially installed charcoal will conform with the requirements of this position. New activated carbon meets the requirements of Table 5-1 of ANSI N509-1980, which meets or exceeds the requirements of Table 1, <Regulatory Guide 1.140>, March, 1978.
6.a(3) Plant operating procedures will conform with the requirements of this position. Laboratory testing for non-ESF adsorbers will be conducted in accordance with the specification for testing of ESF adsorbers. Non-ESF adsorbers are tested with no test parameter exceptions. Testing frequency will meet the intent of the provision but may be based upon refueling outage intervals for systems M14 and M38.
6.b The design conforms to this position. The preoperational testing procedures conform to this position. The plant operating procedures conform to this position, with the exception that utilization of adsorbent samples removed from the subject bed may be used to refill the samplers. The new unused activated carbon used to replace a bed on failure to meet the applicable tests of Table 2 will meet the requirements of Table 5-1 of ANSI N509-1980, which meets or exceeds the requirements of Table 1 of
<Regulatory Guide 1.140>, March 1978. Testing frequency will meet the intent of the provision but may be based upon refueling outage intervals for systems M14 and M38.
Revision 12 12.3-76 January, 2003
TABLE 12.3-4 DETECTORS ASSOCIATED WITH UNIT 1(1)
Detector(4) Channel(2) Location(5) 1D21N030(3) Personnel air lock Containment at Elev. 600-6 1D21N040 CRD HCU west Containment at Elev. 620-6 1D21N050 RWCU fill and drain Containment at Elev. 642-0 receiver tank area east 1D21N060(3) TIP drive area Containment at Elev. 600-6 1D21N070 RWCU fill and drain area Containment at Elev. 664-7 1D21N080(3)(6) Upper pool area Containment at Elev. 689-6 1D21N110 Auxiliary building, Auxiliary building at 574 east Elev. 574-10 east 1D21N120 Auxiliary building, Auxiliary building at 574 west Elev. 574-10 west 1D21N130 Turbine room east Turbine room at Elev. 647-6 east 1D21N140 CRD HCU east Containment at Elev. 620-6 1D21N160(3) Turbine room west Turbine room at Elev. 647-6 west 1D21N170 Turbine building, 605 Turbine building at Elev. 605-6 1D21N180 Hotwell pump area Turbine building at Elev. 577-6 1D21N190 Turbine building sump Turbine power complex at area Elev. 548-6 1D21N200 Offgas building, 584 Offgas building at Elev. 584-0 1D21N210 Condensate filter pump Turbine power complex at area Elev. 568-6 1D21N220 Offgas after-filter area Offgas building at Elev. 602-6 Revision 12 12.3-77 January, 2003
TABLE 12.3-4 (Continued)
Detector(4) Channel(2) Location(5) 1D21N230 High pressure feedwater Heater bay at Elev. 600-6 heater area 1D21N240 Feedpump area Heater bay at Elev. 647-6 1D21N400 Control room Control complex at Elev. 654-6 1D21N410 Offgas holdup area Turbine building at Elev. 577-6 NOTES:
(1)
Readout modules located on Panel 1H13-P803 and recorded on Panel 1H13-P600.
(2)
All channels have a local alarm and indicating unit except 1D21N400.
(3)
In addition to a local alarm and indicating unit, Channels 1D21N030, 1D21N060 and 1D21N080 have remote warning units, and Channel 1D21N160 has a remote warning light. Channel 1D21N060 also has a remote meter mounted on the TIP drive control panel and uses flashing (stroboscopic) blue high radiation alarm lights.
(4)
Range of detectors is 0.1 to 104 mR/hr.
(5)
Detector and local alarm and indicating units located inside containment are housed in unpainted aluminum NEMA Type 4 enclosures.
All other equipment is housed in NEMA Type 12 enclosures.
(6)
(Reference 9) and (Reference 10) should be reviewed prior to making any modification that would impact this instruments ability to be used as an area radiation monitor.
Revision 12 12.3-78 January, 2003
TABLE 12.3-5 (DELETED)
Revision 19 12.3-79 October, 2015
TABLE 12.3-5 (Continued)
(DELETED)
Revision 19 12.3-80 October, 2015
TABLE 12.3-6 DETECTORS ASSOCIATED WITH COMMON AREAS(1)
Detector(4) Channel(2) Location D21N250(3) Radwaste 574 west Radwaste building at Elev. 574-10 west D21N260(3) Radwaste 574 east Radwaste building at Elev. 574-10 east D21N270(3) Radwaste 602 Radwaste building at Elev. 602-0 D21N280 Process sample room Radwaste building at Elev. 623-6 D21N290(3) Radwaste evaporator area Radwaste building at Elev. 623-6 D21N310 Fuel pool cleanup fill Intermediate building at and drain area Elev. 599-0 D21N320(5) Fuel preparation pool Intermediate building at Elev. 620-6 D21N330(5) Spent fuel storage pool Intermediate building at Elev. 620-6 D21N420 Fuel pool cooling Intermediate building at circulating pump area Elev. 574-10 NOTES:
(1) Readout modules located on Panel H13-P906 and recorded on Panel H13-P907.
(2) All channels have an alarm and indicating unit.
(3) In addition to an alarm and indicating unit, Channels D21N250, D21N260, D21N270, and D21N290 have a remote meter and remote alarm in the radwaste building control room.
(4) Range of detectors is 0.1 to 104 mR/hr.
(5) (Reference 9) and (Reference 10) should be reviewed prior to making any modification that would affect these instruments ability to be used as area radiation monitors.
Revision 19 12.3-81 October, 2015
TABLE 12.3-7 DETECTORS ASSOCIATED WITH LOCAL CHANNELS Detector(3) Channel Location(4) 1D21N340 Unit 1 drywell Containment at Elev. 655-0 (portable) (Azimuth approximately 160)
D21N380(1) Waste compactor area Radwaste building at Elev. 623-6 D21N370(1) Solid radwaste drumming Radwaste building at area Elev. 623-6 NOTES:
(1) Channels D21N380 and D21N370 are locally mounted channels and have a remote meter and a remote alarm in the radwaste building control room.
(2) (Deleted)
(3) Range of detectors is 0.1 to 104 mR/hr.
(4) Detector and alarm and indicator units located inside containment are housed in unpainted aluminum NEMA Type 4 enclosures. All other equipment is housed in NEMA Type 12 enclosures.
Revision 19 12.3-82 October, 2015
TABLE 12.3-8 DETECTOR DESIGN REQUIREMENTS Type G-M tube Range, mR/hr 0.1 to 104 Energy dependence 15% (80 KeV to 1.5 MeV)
Circuitry Solid state preamplifier Mounting Wall bracket Remote Capability Up to 1,500 feet Exposure Capability of withstanding a total integrated dose of 105 rads Enclosure NEMA 12 (NEMA 4 in containment)
Dead Time 20 sec.
Revision 15 12.3-83 October, 2007
TABLE 12.3-9 READOUT MODULE DESIGN REQUIREMENTS Response Time Meter response is approximately 2.5 seconds for full scale deflection.
Time constant of 60, 6, .06 seconds at 0.01, 0.1, 1 mR/hr respectively.
Susceptibility The input signal is shielded to prevent gross fluctuations and false trips due to normal electromagnetic interference caused by electric motors, circuit breaker closure, welding.
Stability Drift is less than 3% of the measured point over a period of 30 days at environmental design center.
The system is capable of operation on 120 volt ac, 60 hertz and will operate within specifications under voltage or frequency changes of 10%.
For the operating temperature range the shift due to temperature will be less than 0.5% per C.
Accuracy The overall accuracy of the system will be the actual reading relative to the true reading within 25% of any decade at a reference energy in the range of 0.1 to 2.5 MeV.
Precision The precision will be 10% of any single measurement level at the environmental design center.
Revision 12 12.3-84 January, 2003
TABLE 12.3-10 AIRBORNE RADIATION MONITOR SUBGROUP UNIT 1 AND COMMON Radiation Monitor Instrument Subsystem(2) Sample Point Channels(1)(3) Function of Subsystem Location D17K720 Radwaste Building Ventilation Ventilation ductwork GSP Local, control room and radwaste Radwaste Bldg.
Exhaust Radiation Monitor upstream of filter HSP control room indication and 623-6 East trains PSP alarms. Ventilation supply fan trip on high radiation.
1D17K700 Auxiliary Building Ventilation Ventilation ductwork GSP Local and control room alarms and Aux. Bldg.
Exhaust Radiation Monitor upstream of filter HSP indication. Ventilation supply 620-6 West trains PSP fan trip on high radiation.
D17K730 Intermediate Building Ventilation Ventilation ductwork GSP Local and control room alarms and Intermediate Bldg.
Exhaust Radiation Monitor downstream of exhaust H&P Filters indication. Supply fan trip on 682-6 S.W.
fan high radiation.
D17K710 Fuel Handling Area Ventilation Ventilation ductwork GSP Local and control alarms and Intermediate Bldg.
Exhaust Radiation Monitor upstream of the HSP indication. Supply fan trip on 682-6 N.W.
exhaust filters PSP high radiation. Fuel handling area evac. alarm on high rad.
1D17K760 Offgas Building Ventilation Ventilation ductwork GSP Local and control room alarms Offgas Bldg.
Exhaust Radiation Monitor upstream of exhaust HSP and indication. 635-0 filter trains PSP 1D17K660 Containment Vessel and Drywell Ventilation ductwork GSP Local and control room alarms Intermediate Bldg.
Purge Exhaust Radiation Monitor outside containment HSP and indication. Drywell and 654-6 upstream of exhaust PSP containment evac. alarm.
filter trains D17K770 Control Room Airborne Radiation Ventilation supply GSP Local and control room alarms and Control Complex Monitor duct downstream of HSP indication. High radiation on 679-6 common supply plenum PSP gas channel shifts ventilation into emergency recirculation mode.
Revision 19 12.3-85 October, 2015
TABLE 12.3-10 (Continued)
Radiation Monitor Instrument Subsystem(2) Sample Point Channels(1)(3) Function of Subsystem Location 1D17K670 Drywell Atmospheric Monitor Drywell, 617-3 Elev. GSP Local and control room indication Fuel Handling Bldg.
HSP and alarms and drywell evacuation 620-6 PSP alarm. Isolation of hydrogen purge Valves M51F090 and M51F110 upon noble gas high alarm.
1D17K680 Containment Atmospheric Recirculated contain- GSP Local and control room indication Intermediate Bldg.
Monitor ment air, 674 Elev. HSP and alarms and containment and 665-0 PSP drywell evac. alarm.
NOTES:
(1)
Analog signals are recorded.
(2)
Tag Numbers prefixed by 1D17 are components associated with Unit 1.
Tag Numbers prefixed by D17 are components associated with the common areas of the plant.
(3)
GSP = Gas chamber scintillator-photomultiplier HSP = Halogen cartridge scintillator-photomultiplier PSP = Particulate filter scintillator-photomultiplier H = Halogen P = Particulate Revision 19 12.3-86 October, 2015
TABLE 12.3-11 ISOKINETIC PROBES UNIT 1 Monitor Isokinetic Duct Flow Ventl. System No.(1) Probe No.(1) (inches) CFM M14 Containment Vessel and Drywell Purge 1D17K660 1D17N661A 48 x 48 5,000 1D17N661B 48 x 48 25,000 M15 Reactor Bldg. Annulus Exhaust Gas Treatment 1D17K690A 1D17N691A 14 x 16 400min(2) 1D17K690B 1D17N691B 14 x 16 400min(2)
M36 Offgas Bldg. Vent. System 1D17K760 1D17N761 30 x 46 16,700(3)
M38 Auxiliary Bldg. Vent. System 1D17K700 1D17N701 30 x 90 29,325 COMMON M25 Control Room HVAC and Emerg. Recirc. System D17K770 D17N771 38 x 32 11,540 M31 Radwaste Bldg. Vent. System D17K720 D17K721A 32 x 90 30,000 D17K721B 32 x 90 30,000 M33 Intermediate Bldg. Vent. System D17K730 D17N731 46 x 46 27,400 M40 Fuel Handling Area Vent. System D17K710 D17N711 40 x 60 30,000 NOTES:
(1) Unit 1 has 1 preceding the number, i.e., 1D17K---.
Common areas have no prefix to the number, i.e., D17K---.
(2) With no recycle to the annulus space, 2,000 CFM is possible.
(3) Based on total Offgas Vent Stack flow.
Revision 19 12.3-87 October, 2015
TABLE 12.3-12 REACTOR BUILDING SUBCOMPARTMENT VENTILATION DATA M11, M14 SYSTEMS Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Reactor Operation: 5,000 -
642 level, RWCU F/D backwash rec. 500(1) 0.10 max.
tank area (radiation Zone 5 area) 654 level east RWCU valve area 200(1) 0.04 max.
(radiation Zone 5 area) 664 level, RWCU F/D valve & holding 750(1) 0.15 max.
pump room (radiation Zone 5 area) 664 level, RWCU F/D room (2) 120(1) (each) 0.024 max.
(radiation Zone 5 area) 664 level north, RWCU HX 2,300(1) 0.46 max.
(radiation Zone 5 area)
Reactor Shutdown Mode of Operation: 25,000 -
Drywell area 16,500 0.66 Containment free space 8,500 0.34 642 level, RWCU F/D backwash rec. 500(1) 0.017 max.
tank area (radiation Zone 5 area) 664 level north, RWCU HX 2,300(1) 0.077 max.
(radiation Zone 5 area) 654 level east RWCU F/D valve nest 200(1) 0.007 max.
area (radiation Zone 5 area) 664 level, RWCU F/D valve & holding 750(1) 0.025 max.
pump room (radiation Zone 5 area) 664 level, RWCU F/D room 120(1) (each) 0.004 max.
(radiation Zone 5 area)
NOTE:
(1)
Exhaust flow rate is a result of M11 Supply Air.
Revision 12 12.3-88 January, 2003
TABLE 12.3-13 RADWASTE BUILDING SUBCOMPARTMENT VENTILATION DATA M31 SYSTEM Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Operation 30,000 -
574-0 Level:
Floor drain collection pump room A and B 400 (each) 0.013 Fuel pool sludge decant pump room 200 0.0066 Equipment drain sump pump 200 0.0066 RWCU sludge decant pump room A 100 0.0033 RWCU sludge decant pump room B 200 0.0066 Floor drain sump room 200 0.0066 Condensate sludge decant pump room 400 0.013 Waste sample pump room 600 0.02 Floor drain sample pump room 600 0.02 Chemical waste pump room A and B 600 (each) 0.02 Corridor 1,700 0.0567 Waste collector transfer pump room A and B 400 (each) 0.03 Radiation Zone 5 Areas:
Fuel pool sludge discharge mixing pump area A and B 400 (each) 0.013 RWCU sludge discharge mixing pump area A and B 400 (each) 0.013 Condensate sludge discharge mix pump area A and B 400 (each) 0.013 Fuel pool F/D backwash settling tank room area A and B 400 (each) 0.013 RWCU F/D backwash settling tank A and B 400 (each) 0.013 Condensate filter backwash settling tank A and B 400 (each) 0.013 602 Level:
Chemical waste distillate tank area 900 0.03 Detergent drain tank & pump area 300 0.01 Corridor 1,800 0.06 Radiation Zone 5 Areas:
Spent resin pump area A and B 400 (each) 0.013 Concentrated waste transfer pump area A and B 400 (each) 0.013 Floor drain collection tank area A and B 600 (each) 0.02 Revision 12 12.3-89 January, 2003
TABLE 12.3-13 (Continued)
Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Radiation Zone 5 Areas: (Continued)
Waste collection tank area A and B 700 (each) 0.023 Spent resin tank area A and B 400 (each) 0.013 Concentrated waste tank area A and B 400 (each) 0.013 Chemical waste tank area A and B 700 (each) 0.023 Floor drain sample tank area A 300 0.01 Floor drain sample tank area B 600 0.02 Waste sample tank area A 600 0.02 Waste sample tank area B 300 0.01 623 Level:
Process sample room 2,250 0.075 Radwaste control panel area Air Supplied Only -
Dry Active Waste Handling Area 3,500 0.1167 Corridor 6,600 0.2200 Loading and drum storage area 3,610 0.1203 Reverse osmosis unit room 300 0.01 Chemical treatment room 300 0.01 Ventilation exhaust equipment room 1,200 0.04 Exhaust system area 1,000 0.033 Valve station 130 0.0043 Radiation Zone 5 Area:
Chemical waste evaporator rooms A and B 900 (each) 0.0333 Waste demineralizer area 140 0.0047 Floor drain demineralizer area 140 0.0047 Waste cement mixing pump room A and B(1) 400 (each) 0.0133 Waste mixing dewatering tank room A and B 400 (each) 0.0133 646 Level:
Filter, precoat pump and tank room 1,540 0.0513 Waste collector filtrate pump room 200 0.0067 Waste collector filtrate room 570 0.0190 Floor drain filtrate pump room 200 0.0067 Floor drain filtrate room 570 0.0190 616 Level:
Full drum storage area 2,000 0.0667 Decontamination area 500 0.0167 NOTE:
(1)
Either portions or all of the equipment located in these areas are abandoned. However, the M31 System is operational. Therefore, this information is being retained in this table.
Revision 13 12.3-90 December, 2003
TABLE 12.3-14 AUXILIARY BUILDING SUBCOMPARTMENT VENTILATION DATA M38 SYSTEM Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Operation 29,325 -
574 level - Corridor 1,800 0.061 Radiation Zone 5 Areas 568 level: LPCS pump room 2,280 0.078 RHR-A pump room 5,415 0.185 RCIC pump room 2,105 0.072 RHR-C pump room 1,545 0.053 RHR-B pump room 5,420 0.185 HPCS pump room 2,280 0.078 599 level - Corridor and containment vessel/turbine building water chiller area/ADHR heat exchanger room 9,885 0.337 599 level - RWCU pump room A 3,175 0.108 RWCU pump room B 915 0.031 614 level - Steam tunnel 4,000 0.136 620 level - Northwest corridor 3,615 0.123 Northeast corridor and ventilation supply equipment area 8,240 0.281
- pipe chase access room 910 0.031 Revision 18 12.3-91 October, 2013
TABLE 12.3-15 INTERMEDIATE BUILDING SUBCOMPARTMENT VENTILATION DATA M33 SYSTEM Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Operation 27,400 -
574 level-General area 900 0.033 Tool Decon Area 1,100 0.040 Tool Storage Area 900 0.033 599 level-General area 5,400 0.197 Control complex controlled access entry 100 0.004 Electrical equipment room 1,200 0.044 620 level 8,300 0.303 Annulus exhaust gas treatment rooms (4) 500 (each) 0.018 654 level-General area & recombiner area 4,800 0.175 Containment vessel & drywell purge (2) 1,000 (each) 0.036 Containment vessel & drywell purge (2) 1,200 (each) 0.044 682 level-General area 6,000 0.218 Fuel handling area exhaust filter rooms (3) 800 (each) 0.029 Revision 12 12.3-92 January, 2003
TABLE 12.3-16 FUEL HANDLING AREA SUBCOMPARTMENT VENTILATION DATA M40 SYSTEM Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Operation 30,000 -
574 level Corridor, north and south 1,000 (each) 0.033 Radiation Zone 5 Areas:
Control rod drive pump room, 2,500 (each) 0.083 north and south Fuel pool cooling and cleanup circulating pump 300 0.01 Fuel pool cooling and cleanup F/D transfer pump room 200 0.0067 Fuel pool cooling and cleanup F/D backwash rec. tank room 250 0.0083 Intermediate building floor and equipment drain sump pump room 250 0.0083 Postaccident sample room 300 0.01 599 level Corridor, north and south 1,000 (each) 0.033 Control rod drive maintenance area 8,100 0.27 Radiation Zone 5 Areas:
Pool Area:
(Cask storage pool, spent fuel pool, Fuel transfer pool) 15,300 0.51 Fuel pool cooling and cleanup heat exchanger 400 0.0133 Fuel pool cooling and cleanup F/D room A, B 200 0.0067 Fuel pool cooling and cleanup F/D room C, D 200 0.0067 Hot I&C repair room 500 0.0167 Revision 12 12.3-93 January, 2003
TABLE 12.3-17 HEATER BAY SUBCOMPARTMENT VENTILATION DATA M41 SYSTEM SUMMER OPERATION(1) WINTER OPERATION(2)
Exhaust Dilution Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Flow Rate (CFM) Factor Normal Operation 360,000 - 198,000 -
560 level (Heater area) 107,000 0.30 53,500 0.27 DC feedwater heater, MCV access area 580 level (Hot water 97,000 0.269 48,500 0.245 heating equipment)
Rad. Zone 5 areas:
DC feedwater heaters and auxiliary boiler evaporator 120,000 0.333 60,000 0.303 600 level (Hallway) 28,500 0.0792 14,250 0.072 Rad. Zone 5 areas:
intermediate feedwater heater, HP feedwater heater 158,500 0.440 79,250 0.400 620 level (Hallway) 21,500 0.0597 750 0.004 FDW Lube Oil Purifier Room 1,500 0.004 750 0.004 Rad. Zone 5 area:
Auxiliary Condenser 166,000 0.461 79,600 0.402 647 level (Hallway) 14,000 0.0389 400 0.002 Rad. Zone 5 areas:
Steam seal evaporator 83,000 0.23 39,800 0.20 Feedwater pump area (2) 41,500 0.11 19,900 0.10 (each) (each)
Feedwater heater area 60,000 0.17 22,400 0.11 NOTES:
(1)
With louvers open.
(2)
With louvers closed.
Revision 12 12.3-94 January, 2003
TABLE 12.3-18 TURBINE BUILDING SUBCOMPARTMENT VENTILATION DATA M35 SYSTEM SUMMER OPERATION(1) WINTER OPERATION(2)
Exhaust Dilution Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Flow Rate (CFM) Factor Normal Operation 360,000 - 198,000 -
Condenser bay area 96,000 0.27 96,000 0.48 (radiation Zone 5)
Turbine operating floor 180,000 0.50 118,000 0.60 Hot well pump & L.P.
heater area 25,500 0.07 25,500 0.129 Condenser vacuum pump area 300 0.001 300 0.002 Sample extraction area 600 0.002 600 0.003 Steam seal exhaust area 24,700 0.07 24,700 0.125 NOTES:
(1)
With louvers open (2)
With louvers closed Revision 12 12.3-95 January, 2003
TABLE 12.3-19 OFFGAS BUILDING VENTILATION EXHAUST SUBCOMPARTMENT VENTILATION DATA M36 SYSTEM Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Normal Operation 15,400 -
(Based on total Offgas Vent Stack flow) 16,700 548 level Radiation Zone 5 areas:
Turbine power complex
- a. Condensate demineralizer backwash rec. tank 600 0.035
- b. Condensate filter backwash rec.
tank area 700 0.041 568 level Turbine power complex, corridor 2,200 0.131 Turbine power complex, condensate filter pump area 1,500 0.089 Condensate demin. cubicles (6) 500 (total) 0.029 Condensate filter cubicles (8) 500 (total) 0.029 Caustic & acid storage tank room 2,000 0.119 Offgas exhaust plenum area 600 0.035 Radiation Zone 5 area:
Turbine power complex, Cation regen. tank area 500 0.029 577 level (Radiation Zone 5 area)
Turbine building, holdup pipe area 1,000 0.059 584 level Offgas building, corridor 700 0.041 Revision 14 12.3-96 October, 2005
TABLE 12.3-19 (Continued)
Exhaust Dilution Subcompartment Flow Rate (CFM) Factor Radiation Zone 5 areas:
Offgas cooler condenser room 200 0.011 Offgas regenerator room A and B 250 0.014 (each)
Charcoal absorber rooms 1,000 0.059 (Total) 602 level Radiation Zone 5 areas:
Offgas building, corridor 1,500 0.089 Desiccant dryer area 450 0.026 After filter and prefilter rooms (3) 200 0.011 (each) 605 level Radiation Zone 5 areas:
Steam jet air ejector room A and B 800 0.047 (each)
Preheater area 800 0.047 620 level Offgas building, floor area 2,500 0.149 Offgas sample panels 400 0.023 (Total) 624 level Turbine building, lab. hoods 2,000 0.119 (Total)
Turbine building, hydrogen oxygen analyzer area (A and B) 300 0.017 (each)
Revision 14 12.3-97 October, 2005
12.4 DOSE ASSESSMENT The estimates of exposure use both design radiation zones and the associated occupancy times described in <Section 12.3.1>, along with operational data gathered from similar BWR plants. Design radiation levels used have been developed from conservative assumptions which indicate maximum radiation levels and not those anticipated for normal plant operation.
Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20> regulations prior to October 4, 1993. These assessments, including any reference to Unit 2, are considered historical.
Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised
<10 CFR 20> dated October 4, 1993. During normal plant operation, the site ALARA Committee has the responsibility to determine the annual person-rem exposure for the Perry Nuclear Power Plant. This will be done based on anticipated routine and outage conditions, work loads and number of personnel on site.
12.4.1 ESTIMATES OF PERSONNEL OCCUPANCY REQUIREMENTS Occupancy requirements throughout the plant are based on operating experience and project manpower needs, and were considered in the establishment of the design radiation zones described in
<Section 12.3.1>. To estimate occupancy requirements, plant personnel are categorized into five groups according to work function. Feedback information from operating facilities indicates that contract workers will be called upon to do some tasks as indicated in- a. Skyshine and direct dose from turbines
- b. Direct dose from stored radwaste
- c. Direct dose from the external surfaces of buildings
- d. Dose from the gaseous radioactive plume 12.4.4.1 Skyshine and Direct Dose from Turbines The dose analyses for normal operation of both units were based on an 80 percent load factor, and 50 and 24 percent occupancy factors for offsite and onsite exposures, respectively. For distances beyond 300 feet, a single lumped source was assumed in the turbine building; for distances less than 300 feet, all major sources were considered separately. The resultant doses at selected locations are given in
- a. The activities scheduled during this period on Unit 2 can be divided into three basic categories: completion of construction, startup/testing and site engineering. The total duration of these activities is scheduled to be 74 months between commercial operation of Unit 1 and Unit 2.
- b. The relative orientation of the structures is as shown on
- c. The startup/testing activities on Unit 2 consist of such items as fuel loading, preoperational testing and startup and power testing.
- d. The occupational composition and size for the startup/testing work for a single shift is conservatively assumed as follows:
- 1. Engineers - 25 required
- 2. Electricians - 20 required
- 3. Pipe fitters - 20 required
- 4. Crane operators - 4 required Revision 19 12.4-6 October, 2015
- 5. Operating engineers - 4 required
- 6. Utility personnel - 20 required
- 7. Laborers - 10 required
- 8. Technicians - 50 required
- e. There will be three 9-hour shifts per day.
- f. During any shift, the following time schedules are assumed:
- 1. A portion of the work force (25 percent) spends 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on the turbine building operating floor, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> outside all plant structures and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inside plant structures. It is conservatively assumed that while inside a plant structure, the individuals are shielded by a minimum of two feet of concrete and are located at a distance of 300 feet from the Unit 1 turbine building.
- 2. The remainder of the work force spends 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> outside all plant structures and 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> inside plant structures.
- g. The completion of construction activities on Unit 2 consists of installation of equipment and piping, electrical wiring, closing of concrete construction openings, completion of painting and architectural items and completion of final roadways and landscaping in the area around Unit 2.
- h. The size of the work force needed for the completion of Unit 2 construction, based on present schedule resource loading, is given on <Figure 12.4-1>.
- i. The work force is assumed to be composed of the occupational crafts listed in
- j. During a working day, the following time schedules are assumed:
- 1. A portion of the work force (40 percent) spends 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> outside all plant structures and 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> inside plant structures. It is conservatively assumed that while inside a plant structure, the individuals are shielded by a minimum of two feet of concrete and are located at a distance of 300 feet from Unit 1 turbine building.
- 2. The remaining 60 percent of the work force is conservatively assumed to spend all 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> of the work shift outside plant structures.
- k. The site organization is a field engineering group housed onsite to complete portions of the final design and handle field construction problems. The estimate of the work force does not include operations, radiation protection, maintenance, and I&C personnel since their occupational exposures are discussed in
- l. The size of the site organization work force is shown on
- m. During the working day, the following schedule is assumed: the average worker spends 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> inside the warehouse/office, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> outside all plant structures and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inside plant structures.
4.5 REFERENCES
FOR SECTION 12.4
- 1. U.S. Nuclear Regulatory Commission, <NUREG-0713>, Occupational Radiation Exposure at Commercial Nuclear Power Reactors, 1982, December 1983.
- 2. U.S. Nuclear Regulatory Commission, <NUREG-0322>, Ninth Annual Occupational Radiation Exposure Report, 1976, October 1977.
- 3. U.S. Nuclear Regulatory Commission, Occupational Radiation Dose Assessment in Light-Water Reactor Power Plants, Design Stage Man-rem Estimates, <Regulatory Guide 8.19>, May 1978.
- 4. Vance, J., Weaver, C. L., Lepper, E. M., A Preliminary Assessment of the Potential Impacts on Operating Nuclear Power Plants of a 500 mrem/hr Occupational Exposure Limit, Report to the Nuclear Regulatory Staff by the Atomic Industrial Forum, April 1978.
- 5. Verna, B. J., Radioactive Maintenance, Parts 1-4, Nuclear News, September 1978, November 1975, January 1976, March 1976.
- 6. General Electric Information Document, Mark III Containment Dose Reduction Study, 22A5718, 1/29/80.
- 7. Murphy, T. D. et al., Occupational Exposure at Light Water Cooled Power Reactors, 1969-1975, U.S. Nuclear Regulatory Commission,
<NUREG-0109>, August 1976.
Revision 14 12.4-10 October, 2005
- 8. Pelletier, C. A., et al., Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plant, September 1974.
- 9. Endres, G. W. R., Garcia, W. T., and Shipler, D. B., BNWL-SA-6103, Dose to Construction Workers at Operating Reactor Sites, presented at ACRS Meeting on Radiological and Site Evaluation, Washington, D.C., July 11, 1978.
Revision 12 12.4-11 January, 2003
TABLE 12.4-1 ESTIMATED MANPOWER NEEDS AND OCCUPANCY REQUIREMENTS, UNIT 1 AND UNIT 2 Zone I Zone II Zone III Group Number _(%) _ _ (%) _ _ (%) _
Administrative 100 97 3 0 Radiation Protection 90 80 18.5 1.5 Technical Section 200 93 6 1 Operations Section 200 86 13 1 Maintenance Section 200 70 28.5 1.5 Contractors 300 75 24 1 Total 1,090 Revision 12 12.4-12 January, 2003
TABLE 12.4-2 PERSON-REM ESTIMATES FOR NORMAL PLANT OPERATIONS, ANTICIPATED OPERATIONAL OCCURRENCES AND ROUTINE MAINTENANCE, UNIT 1 AND UNIT 2 Zone I Zone II Zone III Group (0.5 mrem/hr) (2.5 mrem/hr) (25 mrem/hr)
Administrative 48.5 0 0 Radiation Protection 36.0 41.6 33.8 Technical Section 93.0 30.0 50.0 Operations Section 86.0 65.0 50.0 Maintenance Section 70.0 142.5 75.0 Contractors 112.6 180.0 75.0 Subtotals: 828.9 person-rems (station personnel) 367.6 person-rems (contract personnel)
Total: 1,196.5 person-rems Revision 13 12.4-13 December, 2003
TABLE 12.4-3 BOILING WATER REACTORS PERCENTAGES OF EXPOSURE BY JOB FUNCTION(1)
Job Function Utility Contractor Total Maintenance 22.6 55.8 78.4 Operations 6.4 0.9 7.3 Radiation Protection 2.9 3.1 6.0 Supervisory 1.6 0.4 2.0 Engineering _2.6 _3.7 _ 6.2 Totals 36.1 63.9 100 NOTE:
(1)
<NUREG-0713> (Reference 1).
Revision 12 12.4-14 January, 2003
TABLE 12.4-4 BOILING WATER REACTORS PERCENTAGES OF EXPOSURE BY WORK FUNCTION(1)
Work Function Utility Contractor Total Reactor Operations 8.5 2.5 11 Routine Maintenance 4.0 5.0 9 Inservice Inspection 2.8 7.2 10 Special Maintenance 0.4 1.6 2 Waste Processing 0.5 0.5 1 Refueling 34.7 32.3 __67 Totals 50.9 49.1 100 NOTE:
(1)
<NUREG-0713> (Reference 1).
Revision 12 12.4-15 January, 2003
TABLE 12.4-5 YEARLY OPERATIONAL PERSON-REM FOR SELECTED BWR PLANTS(1) 1980 1981 1982
- 1. Dresden 1, 2 & 3 2,105 2,802 2,923
- 2. Monticello 531 1,004 993
- 3. Nine Mile Point 591 1,592 1,264
- 4. Peach Bottom 2 & 3 2,302 2,506 1,977
- 5. Quad Cities 1 & 2 4,838 3,146 3,757
- 6. Vermont Yankee 1,338 731 205 Ave. Person-rem/Unit 1,170 1,178 1,112 NOTE:
(1)
<NUREG-0713> (Reference 1).
Revision 13 12.4-16 December, 2003
TABLE 12.4-6
SUMMARY
OF TOTAL OCCUPATIONAL RADIATION EXPOSURE ESTIMATES BY TASK Dose Percentage of Total Function (person-rems/yr-unit) person-rem dose___
Routine operation 43 11 Routine maintenance 38 10 Waste processing 3 1 Refueling 251 66 Inservice inspection 39 10 Special maintenance 6 2 Total person-rems/yr-unit 380 Revision 16 12.4-17 October, 2009
TABLE 12.4-7 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE OPERATIONS AND SURVEILLANCE Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Control room 0.1 6,000 2 - 1.2 Walking and checking:
Turbine and feedwater 100 0.1 1/shift 10.0 heat exchanger 1.0 1.0 1 1/shift 1.0 Containment cooling system 1.0 1.0 1 1/day 0.36 Standby liquid control system 1.0 1.0 1 1/day 0.36 ECCS and process equip 1.0 1.0 1 1/shift 1.0 C&I panels and equip in containment 1.0 1.0 1 1/shift 1.0 Fuel pool system 1.0 0.4 1 1/day 0.1 RWCU 1.0 0.5 1 1/shift 0.6 CRD system 1.0 0.5 1 1/shift 0.6 Recirc flow control 1.0 0.6 1 1/day 0.22 Misc auxiliary building 1.0 1.0 1 1/shift 1.0 100 0.1 1 1/shift 10.0 Traversing incore probe system 10 0.1 1 1/shift 1.2 Misc. in containment 1.0 1.0 1 1/day 0.4 Instrument calibration in containment 1.0 0.6 1 1/week 0.03 Revision 13 12.4-18 December, 2003
TABLE 12.4-7 (Continued)
Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Radiochemistry 1.0 1.0 2 1/day 0.73 Radiation protection surveys 1.0 4.0 1 1/day 1.46 15 1.0 1 1/day 5.48 100 0.5 1 1/week 2.6 Sample stations in reactor building 5.0 0.5 1 1/shift 2.7 Other local samples 5.0 0.1 1 1/day 0.15 Remote sampling 1.0 0.3 1 1/day 0.1 Containment personnel 1.0 0.05 3 1/shift 0.16 lock Total 43 Revision 13 12.4-19 December, 2003
TABLE 12.4-8 OCCUPATIONAL DOSE ESTIMATES DURING ROUTINE MAINTENANCE Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
RWCU filter precoat 1.0 2.0 1 1/week 0.1 RWCU pump and valve 150 1.5 2 1/week 23.4 TIP system 10 2.0 2 1/month 0.5 CRD 2.5 10.0 1 1/week 1.3 HVAC systems 1.0 8.0 1 1/week 0.4 Sample stations 5.0 1.5 1 1/day 2.7 Misc. auxiliary 2.0 4.0 2 1/week 0.8 building pumps 25.0 0.2 2 1/week 0.5 (LPCS, HPCS, RCIC, etc.)
Feedwater and 25.0 0.5 2 1/week 1.3 condensate pumps 1.0 1.5 2 2/weeks 0.3 valves and heat exchangers Condensate 1.0 1.5 2 1/day 1.1 demineralizer and heat exchangers filters Total 32.4 Revision 17 12.4-20 October, 2011
TABLE 12.4-9 OCCUPATIONAL DOSE ESTIMATES DURING WASTE PROCESSING Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Operation of liquid 0.2 112 1 1/week 1.1 waste processing system Operation of solid 1.0 30 1 1/week 1.5 waste processing system Total 3 Revision 13 12.4-21 December, 2003
TABLE 12.4-10 OCCUPATIONAL DOSE ESTIMATES DURING REFUELING Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Reactor pressure 60 40 10 1/year 27 vessels head and internals - removal and installation Refueling operations 10 100 15 1/year 18.0 Fuel sipping 2.0 100 2 1/year 0.4 CRD replacement 260 35 5 1/year 54 CRD repair 15 200 6 1/year 20 Low Power range 95 20 4 1/year 9 monitor MSIV repair 75 100 6 1/year 52 RHR system 50 27 8 1/year 11 RWCU pump 180 35 3 1/year 22 RWCU valve and 110 45 6 1/year 36 heat exchanger Turbine inspection 3 80 10 1/3 years 0.8 Turbine overhaul 3 250 20 1/20 years 0.75 Total 251 Revision 13 12.4-22 December, 2003
TABLE 12.4-11 OCCUPATIONAL DOSE ESTIMATES DURING INSERVICE INSPECTION Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Providing access: 200 10 4 1/year 8 installation of platforms, ladders, etc., removal of thermal insulation Drywell 100 50 6 1/year 30 Reactor building 5 50 2 2/year 1 Total 39 Revision 13 12.4-23 December, 2003
TABLE 12.4-12 OCCUPATIONAL DOSE ESTIMATES DURING SPECIAL MAINTENANCE Average Exposure Number Dose Dose Rate Time of (person-rem/
Activity (mrem/hr) __(hr)__ Workers Frequency year)____
Feedwater sparger 800 60 244 1/40 years 6 Special cleaning evolutions or See Note(1) one-time plant modifications of radiological significance as defined by the Perry staff.
Total 6 Note:
(1)
No dose is specified for activity. Dose will be estimated on a case-by-case basis based upon the specific radiological conditions, task duration, and manpower requirements to accomplish the activity.
Revision 13 12.4-24 December, 2003
TABLE 12.4-13 PERSONNEL RADIATION DOSES FROM AIRBORNE ACTIVITY Yearly Person-rem Dose Rates Doses (person-rem/
Routine Routine Instrument Main Total (rem/hr) yr-unit)
Surveillance Operation Calibration tenance Refueling ISI man-hrs/yr WB Thyroid WB Thyroid Reactor building 5,000 1,100 70 2,200 --- - - 8,370 1.8 x 10-4 4.3 x 10-3 1.5 36.0 (during operation)
Reactor building --- --- 40 2,600 1,500 440 4,580 Neg(1) 3.8 x 10-4 Neg 1.7 (refueling)
Fuel handling 150 --- 20 300 1,500 - - 1,970 Neg 1.5 x 10-5 Neg 3.0 x 10-2 building Radwaste building 150 7,400 20 300 --- - - 7,870 Neg 1.2 x 10-4 Neg 9.4 x 10-1 Turbine building 1,200 --- 40 5,000 --- - - 6,240 2.5 x 10-4 9.0 x 10-5 1.6 5.6 x 10-1 Total 3.1 37.7 NOTE:
(1)
Negligible.
Revision 13 12.4-25 December, 2003
TABLE 12.4-14 SAFETY/RELIEF VALVE DISCHARGERS DOSE FOR TYPE 2 EVENT Organ Dose Whole body + eye () 150 mrem/event Skin () 440 mrem/event Thyroid .87 mrem/event Revision 12 12.4-26 January, 2003
TABLE 12.4-15 ESTIMATED SKYSHINE DOSES(1)
Distance (ft) Occupancy Factor(2) Dose (mrem/yr) 500 0.19 258 1,500 0.19 19 (40 hrs/wk) 2,900 .4 1.6 (Exclusion boundary) (50% occupancy)
NOTES:
(1)
Two units operating.
(2)
Plant factor (80 percent) times percentage occupancy.
Revision 12 12.4-27 January, 2003
TABLE 12.4-16 WORK FORCE BY OCCUPATIONAL CRAFTS Approximate Percentage Occupational Craft ___of Work Force (%)__
Boilermakers 7 Electricians 18 Ironworkers 9 Pipe fitters 15 Painters 7 Laborers 11 Operating engineers 6 Carpenters 9 All other crafts 18 Total 100 Revision 12 12.4-28 January, 2003
TABLE 12.4-17
SUMMARY
OF DIRECT DOSES(1)
_______________Distance (ft)_______
2,900 Exclusion Source 500 1500 Boundary Skyshine 258 19 1.6 Surfaces of buildings 7.2 x 10-1 6.3 x 10-2 2.6 x 10-3 See Note(2) See Note(2)
Radioactive plume 13.6 7.8 1.1 NOTES:
(1)
Doses in mrem/yr.
(2)
Estimated from- 1. 25% of work force 5 hrs/day in turbine building(1) Negligible Negligible 0.046 0.23 2.0 2 hrs/day inside plant(2) Negligible Negligible 0.0016 0.0032 0.028 2 hrs/day outside plant(3) 0.0008 0.025 0.16 0.37 3.2
- 2. 75% of work force 2 hrs/day outside plant(3) 0.0008 0.025 0.16 0.37 9.6 7 hrs/day inside plant(2) Negligible Negligible 0.0016 0.0112 0.29 Total 15.1 person-rem B. Construction workers(5)
- 1. 40% of work force 2 hrs/day outside plant(4) 0.00026 0.008 0.06 0.14 111.8 7 hrs/day inside plant(2) Negligible Negligible 0.0016 0.0112 8.9 Revision 13 12.4-30 December, 2003
- 2. 60% of work force 9 hrs/day outside(4) 0.00026 0.008 0.06 0.61 731.0 Total 851.7 person-rem C. Site Organization(5) (mrem/hr) (mrem/hr) (mrem/hr) (mrem/day) (person-rem) 2 hrs/day inside plant(2) Negligible Negligible 0.0016 0.0032 2.8 2 hrs/day outside plant(3) 0.0008 .025 .16 .37 327.6 5 hrs/day at warehouse office(6) 0.00026 0.008 .02 .14 124.0 Total 454.4 person-rem NOTES:
- a. Oil Lab
- b. Chemistry Office
- c. (Deleted)
- d. Chemistry Counting Room
- e. Low Level Chemistry Laboratory
- f. High Level Chemistry Laboratory
- g. Respirator Cleaning
- h. (Deleted)
- i. Mens and Womens Locker Rooms and Lavatories
- j. (Deleted)
- k. Personnel Decontamination Room
- l. Lunch Room
- m. (Deleted)
- n. Respirator Issue Room See <Figure 1.2-4> for the arrangement of these facilities.
- a. Health Physics Storage & Frisking Room
- b. Health Physics and Radiation Protection Offices and Support Facilities
- c. Audio/Visual Room
- d. Whole Body Counting Room
- e. RRA Access Control Point See <Figure 1.2-5> for the arrangement of these facilities.
- a. The health protection problems associated with radiation or radioactive materials
- b. The precautions or procedures to minimize exposure
- c. The purpose and use of protective devices
- d. The appropriate response to warning signals made in the event of an unusual occurrence or malfunction at the plant
- e. Applicable sections of the license and Title 10 of the Code of Federal Regulations for the protection of personnel from exposures to radiation or radioactive materials.
- a. A notice describing the following documents including where they may be examined:
- 1. <10 CFR 19>
- 2. <10 CFR 20>
- 3. PNPP licenses
- 4. PNPP Operating Procedures applicable to licensed activities
- b. Any notice of violation involving radiological working conditions, any proposed or actual imposition of civil penalty, and order issued for imposing requirements or for modifying, suspending or revoking a license, and any response from PNPP.
- c. Form NRC-3.
- a. A report of the total effective dose equivalent received by that worker during the previous full year from January 1 through December 31.
- b. A report of the total effective dose equivalent received by a terminated worker while that worker was engaged in activities pursuant to PNPP licenses.
- c. A report of the actual or estimated total effective dose equivalent received by a terminated worker during the final calendar year or portion thereof that the worker was engaged in activities pursuant to PNPP licenses. In addition, a worker shall be provided a report of any exposure reported to the NRC for that specific worker.
6.1 INTRODUCTION
An accident equivalent to that described in <Regulatory Guide 1.3> would release large fractions of the nuclear core fission product inventory.
Once released these fission products could be transferred to various areas in the plant creating high radiation areas and limiting personnel access. This review determines if these postaccident radiation fields unduly limit personnel access to areas necessary for mitigation of an accident. Corrective actions for problems identified are also determined. This design review of plant shielding for spaces and systems required for postaccident operations outside containment is in accordance with TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations <NUREG-0578> Section II.B.2, as clarified by
<NUREG-0660> and <NUREG-0737>.
The review is based on the following guidelines:
(Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the <10 CFR 20> regulations prior to October 4, 1993.
Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised
<10 CFR 20> dated October 4, 1993.)
- a. The postaccident dose rate in areas requiring continuous occupancy should not exceed 15mR/hr (control room and onsite technical support center.)
Revision 12 12.6-1 January, 2003
- b. The postaccident dose rate in areas which do not require continuous occupancy should be such that the dose to an individual during a required access period is less than 5 rem whole body (total effective dose equivalent) or its equivalent (sample stations, panels, motor control centers, etc.)
- c. The integrated dose to safety equipment as a result of the accident should be less than the dose for which the equipment has been qualified to ensure that the capability of the equipment to perform its safety function has not been degraded.
- d. The minimum radioactive source term used in the evaluation should be equivalent to the source term recommended in the <Regulatory Guide 1.3>.
- a. Noble gas (Kr, Xe) 100%
- b. Halogens (I, Br) 50%
- c. Alkali metal (Cs, Rb) 50%
- d. Others 1%
- a. Source A - Liquid in the suppression pool and other systems not isolated from the core at the start of the accident and containing only liquid from a depressurized source is assumed to contain the following percents of core inventory of radioactive fission products:
- 1. Noble gases (Kr, Xe) 0%
- 2. Halogens (I, Br) 50%
- 3. Alkali metals (Cs, Rb) 50%
- 4. Others 1%
- b. Source B - For evaluating vital areas, the drywell atmosphere is assumed to contain the following percents of radioactive fission product core inventory:
- 1. Noble gases (Kr, Xe) 100%
- 2. Halogens (I, Br) 25%
- 3. Others 0%
- c. Source C - For evaluating vital areas, the primary containment atmosphere is assumed to contain the following percents of radioactive fission product core inventory:
- 1. Noble gases (Kr, Xe) 100%
- 2. Halogens (I, Br) 25%
- 3. Others 0%
- d. Source D - Until the reactor vessel is depressurized, gases in the steam lines and any other vapor containing lines not isolated from the reactor coolant system are assumed to contain the following percents of radioactive fission product core inventory:
- 1. Noble gases (Kr, Xe) 100%
- 2. Halogens (I, Br) 25%
- 3. Others 0%
- e. Source E - For equipment qualification inside containment, the larger of the two following source terms are used:
- 1. Source terms:
- 2. Source terms:
- f. Source F - For qualification of equipment inside containment for non-LOCA events which do not depressurize the primary system, the following source terms are used:
- 1. Noble gases (Kr, Xe) 10%
- 2. Halogens (I, Br) 10%
- 3. Others 0%
- a. Unattenuated radiation penetrating adjacent compartments shield walls.
- b. Direct radiation from piping or equipment.
- c. Radiation scattered over or around shield walls.
- a. Normal operating sources which may exist at the time of the accident.
- b. Airborne sources from equipment leakage.
- a. For continuous occupancy - the Control Room and Technical Support Center.
- b. For frequent occupancy - the Remote Shutdown Panel, if required.
- c. For infrequent occupancy - the Sampling Station, Sample Analysis Area, Auxiliary Building elevation 620 east end in area of 1P57F0565B (outboard MSIV accumulator safety-related air isolation valve) and the Remote Shutdown Panel.
SUMMARY
Postaccident radiation zone drawings are given in <Figure 12.6-1>,
<Figure 12.6-2>, <Figure 12.6-3>, <Figure 12.6-4>, <Figure 12.6-5>,
<Figure 12.6-6>, and <Figure 12.6-7>. These radiation zones represent the maximum expected radiation dose rate for each area at the start of the accident. Normal operating dose levels which may exist at the time of the accident are not shown on these drawings. A summary of major locations and accident dose rates is given in6.8 REFERENCES
FOR SECTION 12.6
- 1. Arnold, E. D. and Maskewitz, B. F., SDC-Shield Design Calculation Code for Fuel Handling Facilities, ORNL-3041, March 1966.
- 2. Kamphouse, J. L., SPOT1 Shield Code, Gilbert Associates, Inc.,
October 1979.
- 3. A. Tobia, Data for the Calculation of Gamma Radiation Spectra and Beta Heating from Fission Products, Revision 3, Central Electricity Generating Board, RD/B/M2666, CNDC (73) P4, June 1973.
Revision 19 12.6-11 October, 2015
TABLE 12.6-1 OCCUPANCY AND RADIATION DESIGN OBJECTIVES Required Occupancy Dose Rate Limit Integrated Dose Objective Continuous 15 mR/hr 5 rem for duration Frequent 100 mR/hr 5 rem for all activities Infrequent 500 mR/hr 5 rem per activity Accessway 5000 mR/hr Included in above doses Revision 12 12.6-12 January, 2003
TABLE 12.6-2 INITIAL CORE INVENTORY ISOTOPE Ci/WATT I-131 2.7 + 4 I-132 3.8 + 4 I-133 5.5 + 4 I-134 5.9 + 4 I-135 5.1 + 4 Kr-83m -
Kr-85m 7.2 + 3 Kr-85 2.9 + 2 Kr-87 1.2 + 4 Kr-88 1.8 + 4 Sr-89 2.4 + 4 Sr-90 2.3 + 3 Y-90 2.4 + 3 Sr-91 3.0 + 4 Y-91 3.1 + 4 Sr-92 3.3 + 4 Y-92 3.3 + 4 Sr-93 3.8 + 4 Y-93 3.9 + 4 Y-94 4.1 + 4 Rh-105m 5.1 + 3 Revision 12 12.6-13 January, 2003
TABLE 12.6-2 (Continued)
ISOTOPE Ci/WATT Ru-106 1.6 + 4 Rh-106 1.7 + 4 Rh-107 1.6 + 4 Sb-127 2.8 + 3 Te-127 3.6 + 3 Sb-128 4.1 + 3 Sn-128 2.6 + 3 Sb-129 9.0 + 3 Te-129m 1.6 + 3 Cs-139 4.9 + 4 Ba-139 5.0 + 4 Ba-140 4.8 + 4 La-140 5.0 + 4 Ba-141 4.6 + 4 La-141 4.6 + 4 Ce-141 4.6 + 4 Pr-142 3.5 + 3 Ba-142 3.9 + 4 La-142 4.2 + 4 Xe-131m -
Xe-133m 1.5 + 3 Xe-133 5.5 + 4 Xe-135m 9.7 + 3 Revision 12 12.6-14 January, 2003
TABLE 12.6-2 (Continued)
ISOTOPE Ci/WATT Xe-135 7.4 + 3 Xe-138 4.7 + 4 Br-83 2.7 + 3 Br-84 5.4 + 3 Br-88 1.8 + 4 Rb-88 2.3 + 4 Br-89 2.3 + 4 Y-95 4.4 + 4 Zr-95 4.5 + 4 Nb-95 4.5 + 4 Zr-97 4.6 + 4 Nb-97 4.6 + 4 Mo-99 5.1 + 4 Tc-99m 4.5 + 4 Mo-101 4.6 + 4 Ru-103 4.4 + 4 Rh-103m 4.3 + 4 Tc-103 3.5 + 4 Ru-105 2.4 + 4 Rh-105 2.4 + 4 Te-129 8.6 + 3 Sb-130 1.3 + 4 Sb-131 2.2 + 4 Revision 12 12.6-15 January, 2003
TABLE 12.6-2 (Continued)
ISOTOPE Ci/WATT Te-131 2.4 + 4 Te-131m 4.3 + 3 Te-132 3.8 + 4 Te-133 2.4 + 4 Te-133m 3.2 + 4 Te-134 4.9 + 4 Cs-134 2.0 + 3 Cs-137 3.3 + 3 Ba-137m 3.0 + 3 Cs-138 5.0 + 4 Ce-143 4.1 + 4 Pr-143 4.0 + 4 Ce-144 3.5 + 4 Pr-144 3.5 + 4 Pr-145 2.3 + 4 Pr-146 2.3 + 4 Pr-147 1.7 + 4 Nd-147 1.8 + 4 Pm-148 3.8 + 3 Nd-149 1.0 + 4 Pm-149 1.5 + 4 Pm-151 5.4 + 3 Eu-156 4.8 + 3 Revision 12 12.6-16 January, 2003
TABLE 12.6-3 INITIAL RADIOACTIVE SOURCE TERMS (GAMMAS/CC-SEC)
GAMMA-ENERGY SOURCE SUPPRESSION RWCU STEAM REACTOR BLDG REACTOR COOLANT GROUP (MEV)(1) POOL SYSTEM DOME ATMOSPHERE SYSTEM
%CORE NOBLE GAS 0 10 100 100 100
%CORE HALOGENS 50 10 25 25 50
% CORE SOLIDS 1/50 (Cs-Rb) 0 0 0 1/50 (Cs-Rb) 0.1 - 0.5 1.86 +9 5.46 +9 2.06 +10 1.34 +8 4.69 +10 0.5 - 1.0 6.84 +9 1.33 +10 3.32 +10 2.15 +8 7.17 +10 1.0 - 1.5 3.26 +9 3.26 +9 4.41 +10 2.86 +8 3.48 +10 1.5 - 2.0 6.63 +8 1.48 +9 1.18 +10 7.69 +7 1.09 +10 2.0 - 2.5 4.57 +8 1.16 +9 8.50 +9 5.52 +7 1.43 +10 2.5 - 3.0 9.31 +7 1.94 +8 1.29 +10 8.36 +7 2.72 +9 3.0 - 3.5 1.49 +7 2.39 +6 2.28 +9 1.48 +7 1.49 +8 3.5 - 4.0 2.76 +7 3.87 +7 1.14 +8 7.41 +5 2.76 +8 4.0 - 5.0 7.74 +5 5.80 +5 1.80 +6 1.16 +4 7.74 +6 Note:
(1)
(Reference 3) of <Section 12.6.8>
Revision 12 12.6-17 January, 2003
TABLE 12.6-4 DOSE RATES Time 0 Dose Rate Location (mR/hr)
AUXILIARY BUILDING Steam Tunnel 2.10E+9 RWCU Pump Cubicle 4.83E+8 LPCS Pump Cubicle 3.89E+8 RCIC Pump Cubicle 1.37E+9 RHR HX Cubicle 5.57E+8 Corridor Outside RCIC pump room 568-4 3.50E+4 Outside LPCS pump room 568-4 4.41E+3 Outside RHR HX room 568-4 2.84E+4
@ Elev. 599-0 (see Figure 12.6-2)
North Corridor and East side 1.61E+7 West side 2.84E+4 Revision 19 12.6-18 October, 2015
TABLE 12.6-4 (Continued)
INTERMEDIATE BUILDING Above Elev. 646-0 3.11E+4 Elev. 620-6 1.67E+4 Elev. 599-0 2.72E+4 Elev. 574-10 1.34E+3 DIESEL GENERATOR BUILDING 5.09E+0 CONTROL COMPLEX <15(1)
Control Room <15(1)
CENTRAL ALARM STATION 3.23E+2(1)
GUARD HOUSE 1.11E+2(1)
TECHNICAL SUPPORT CENTER <15(1)
NOTE:
(1) Represents average dose rate for 30 days.
Revision 19 12.6-19 October, 2015