ML20255A253
| ML20255A253 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/25/2018 |
| From: | Entergy Operations |
| To: | Plant Licensing Branch IV |
| Klett A | |
| Shared Package | |
| ML20254A353 | List: |
| References | |
| EPID L-2020-LLA-0164 | |
| Download: ML20255A253 (192) | |
Text
WSES-FSAR-UNIT-3 12-i CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS Section Title Page 12.0 RADIATION PROTECTION 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES 12.1-1 ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS 12.1-1 12.1.2 DESIGN CONSIDERATIONS 12.1-1 12.1.3 OPERATIONAL CONSIDERATIONS 12.1-9 12.1.4 DECOMMISSIONING CONSIDERATIONS 12.1-10 12.2 RADIATION SOURCES 12.2-1 12.2.1 CONTAINED SOURCES 12.2-1 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES 12.2-5
12.2 REFERENCES
12.2-9 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 FACILITY DESIGN FEATURES 12.3-1 12.3.2 SHIELDING 12.3-12 12.3.3 VENTILATION 12.3-19 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY 12.3-21 MONITORING INSTRUMENTATION
12.3 REFERENCES
12.3-32 12.3A TMI SHIELDING STUDY 12.3A-1 12.3A.1 INTRODUCTION 12.3A-1 12.3A.2 SOURCE TERMS 12.3A-1 12.3A.3 ANALYZED SYSTEMS AND AREA DOSE RATES 12.3A-3 12.3A REFERENCES
WSES-FSAR-UNIT-3 12-ii CHAPTER 12 TABLE OF CONTENTS (Cont'd)
Section Title Page 12.4 DOSE ASSESSMENT 12.4-1 12.4.1 ANTICIPATED DOSE RATES 12.4-1 12.4.2 ESTIMATES OF PLANT PERSONNEL EXPOSURES (DIRECT) 12.4-1
12.4 REFERENCES
12.4-6 12.5 HEALTH PHYSICS PROGRAM 12.5-1 12.5.1 ORGANIZATION 12.5-1 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 12.5-3 12.5.3 PROCEDURES 12.5-6
WSES-FSAR-UNIT-3 12-iii Revision 308 (11/14)
CHAPTER 12 LIST OF TABLES Table Title 12.2-1 Neutron Fluxes Outside the Reactor Pressure Vessel 12.2-2 Neutron Leakage Rate from Axial Regions of the Reactor Vessel 12.2-3 Gamma Fluxes at Full Power Operation 12.2-4 Reactor Coolant System Sources (Gamma Spectra) at Shutdown 12.2-5 Reactor Coolant System Sources (Material Activation Spectra) at Shutdown 12.2-6 Pressurizer Steam Section Activity 12.2-7 Maximum Activity Inventory in CVCS Components (Curies) 12.2-8 Maximum Activity Inventory in BMS Components (Curies) 12.2-9 Maximum Activity Inventory in FPS Components 12.2-10 Maximum and Expected Activity Inventory in SIS Components (Curies) 12.2-11 Maximum Activity Inventory in WMS Components (Curies)
(DRN 03-2066, R14) 12.2-12 LOCA Core Inventory (Curies/MWt) 12.2-12a Core Inventory for Steaming Events (Curies)
(DRN 03-2066, R14) 12.2-13 Fission Product Gamma Source in Containment Building (Mev/sec) 12.2-14 Assumptions and Parameters used to Calculate Airborne Concentrations 12.2-15 Average Airborne C/MPC in Reactor Auxiliary Building, Turbine Building, Containment and Fuel Handling Building (DRN 02-110, R12) 12.2-15a Average Airborne C/DAC in Reactor Auxiliary Building, Turbine Building, Containment, and Fuel Handling Building (DRN 02-110, R12) 12.2-16 Reactor Auxiliary Building Room by Room C/MPC and Whole Body Dose Commitment Values (DRN 02-110, R12) 12.2-16a Reactor Auxiliary Building Room by Room C/DAC and Dose Commitment Values (DRN 02-110, R12)
(DRN 03-2066, R14) 12.2-17 18-Group Gamma-Ray Source Strengths per Fuel Assembly 3 Days after Shutdown (DRN 03-2066, R14)
(LBDCR 14-007, R308) 12.2-18 Waterford 3 Original Steam Generator Storage Facility Radioactive Isotope Inventory (LBDCR 14-007, R308) 12.3-1 Allowable Dose Rates 12.3-2 Area Radiation Monitors 12.3-3 Airborne Radiation Monitors 12.3-4 Neutron Streaming Dose Rates in Containment
WSES-FSAR-UNIT-3 12-iv Revision 14 (12/05)
CHAPTER 12 LIST OF TABLES (Cont'd)
Table Title (DRN 03-2066, R14) 12.3A-1 TMI Source Term for Shielding Evaluation (DRN 03-2066, R14) 12.3A-2 SIS Sump Water Total Source Strengths 12.3A-3 SIS Sump Water Noble Gas Source Strength Ratios 12.3A-4 SIS Sump Water Halogen Source Strength Ratios 12.3A-5 SIS Sump Water All Others Source Strength Ratios 12.3A-6 Containment Gaseous Total Source Strengths 12.3A-7 Containment Plateout Total Source Strengths 12.3A-8 System Containing Radioactive Material 12.3A-9 Areas Requiring Accessibility Following An Accident 12.4-1a Estimate of Personnel Exposure (Man-Rem) 12.4-1b Estimate of Personnel Exposure During Maintenance (Man-Rem) 12.4-2 Data from Operating PWR Plants 12.4-3 Yearly Averages and Grand Average for Number of Personnel and Man-Rem Doses for Operating PWR Plants 12.4-4 Distribution Man-Rem Doses for Various Functions 12.5-1 Counting Room Instrumentation 12.5-2 Portable Radiological Survey Instrumentation 12.5-3 Personnel Monitoring Instrumentation 12.5-4 Health Physics Equipment
WSES-FSAR-UNIT-3 12-v CHAPTER 12 RADIATION PROTECTION LIST OF FIGURES Figure Title 12.1-1 Radiation Doses from Various Decommissioning Activities 12.1-2 Radiation Doses from Decommissioning 12.3-1 Radiation Zones Reactor Auxiliary Building Plan EL(+)46.00' 12.3-1a Radiation Zones Reactor Auxiliary Building Plan EL(+)21.00' 12.3-1b Radiation Zones Reactor Auxiliary Building Plan EL(-)4.00' 12.3-2 Radiation Zones Reactor Auxiliary Building Plan EL(-)35.00' 12.3-2a Radiation Zones Reactor Auxiliary Building Section (Sheet 1) 12.3-2b Radiation Zones Reactor Auxiliary Building Section (Sheet 2) 12.3-3 Radiation Zones Reactor Auxiliary Building Section (Sheet 3) 12.3-3a Radiation Zones Reactor Auxiliary Building Plan & Sections (Sheet 4) 12.3-3b Radiation Zones Fuel Handling Building Plans 12.3-4 Radiation Zones Fuel Handling Building Sections 12.3-4a Radiation Zones Reactor Building Plan EL(+)46.00' 12.3-4b Radiation Zones Reactor Building Plan EL(+)21.00' 12.3-5 Radiation Zones Reactor Building Plan EL(-)4.00' 12.3-5a Radiation Zones Reactor Building Plan EL(-)35.00' 12.3-5b Radiation Zones Reactor Building Section (Sheet 1) 12.3-6 Radiation Zones Reactor Building Sections (Sheet 2) 12.3-6a Radiation Zones Reactor Building Section (Sheet 3) 12.3-6b Radiation Zones Turbine Building Operating Floor - Plan 12.3-7 Radiation Zones Turbine Building Mezzanine Floor - Plan 12.3-7a Radiation Zones Turbine Building Ground Floor - Plan 12.3-7b Radiation Zones Turbine Building Section
WSES-FSAR-UNIT-3 12-vi Revision 307 (07/13)
CHAPTER 12 RADIATION PROTECTION LIST OF FIGURES (Cont'd)
Figure Title 12.3-8 Radiation Zones Turbine Building Sections (LBDCR 13-010, R307) 12.3-8a Radiation Zones Original Steam Generator Storage Facility (LBDCR 13-010, R307) 12.3-9 Reactor Auxiliary Building Normal Ventilation System Ductwork Plan View 12.3-10 Reactor Auxiliary Building Normal Ventilation System Ductwork Section View 12.3-11 Radiation Monitoring System Block Diagram 12.3-12 Intentionally Deleted (DRN 99-2362, R11) 12.3-13 Intentionally Deleted (DRN 99-2362, R11) 12.3-14 Main Control Room Radiation Monitors 12.3-15 Containment Atmosphere Particulate, Iodine & Gas Radiation Monitor 12.3A-1 Shielding and Iodine Study Plot Plan 12.3A-2 Shielding and Iodine Reactor Building Plan EL-4.00' and 35.00' 12.3A-3 Shielding and Iodine Study Reactor Building Plan EL 21.00' 12.3A-4 Shielding and Iodine Study Reactor Auxiliary Building Plan EL-35.00' 12.3A-5 Shielding and Iodine Study Reactor Auxiliary Building Plan EL-4.00' 12.3A-6 Shielding and Iodine Study Reactor Auxiliary Building Plan EL 7.00' 12.3A-7 Shielding and Iodine Study Reactor Auxiliary Building Plan EL +21.00' 12.3A-8 Shielding and Iodine Study Reactor Auxiliary Building Plan El +46.00'
WSES-FSAR-UNIT-3 12-vii Revision 14-B (06/06)
UPDATE REFERENCE LIST Chapter 12 Section Cross References Revision 12-B Table 12.3A-9 ER-W3-1999-1044/DRN 03-114 Revision 13 Section 12.1.2 ER-W3-2002-0621-000/DRN 03-1135 Section 12.1.3 Section 12.4.2.1 Section 12.5.3.7.2 Section 12.5.1 Section 12.5.2.1 ER-W3-2002-0621-000/DRN 03-1429 Revision 14 Table Of Contents ER-W3-2001-1149-000/DRN 03-2066 Section 12.1.4 Section 12.2.1.4 Section 12.2.1.5 Section 12.2.1.9 Section 12.3-14 Section 12.3.1.8 Section 12.3.24 Section 12.3.3.1 Section 12.3A.2 Section 12.3A.3.1 Section 12.3A.3.10 Reference 12.3A Table 12.2-12 Table 12.2-12a Table 12.2-17 Table 12.3-1 Table 12.3A-1 Section 12.3A-1 ER-W3-2001-1149-013/DRN 05-1249 Section 12.3.2.1 ER-W3-2004-0276-001/DRN 05-144 Section 12.3A.2 Section 12.3A.3.8 Table 12.3A-9 sh-2 Figure 12.3A-2 Figure 12.3A-3 Figure 12.3A-7 Figure 12.3A-8 Table 12.2-3 ER-W3-2005-0083-000/DRN 05-455 Table 12.2-15a sh 1-2 Table 12.2-16a sh 1-4 Revision 14-B Section 12.3.1.4 ER-W3-2005-0418-000/DRN 06-319
WSES-FSAR-UNIT-3 12-viii Revision 307 (07/13)
UPDATE REFERENCE LIST Chapter 12 Section Cross References Revision 15 Section 12.5.2.2.3 ER-W3-2006-0210-000/DRN 06-625 Section 12.5.3.7.1 Section 12.3.4.2.3.1 ER-W3-2005-0471-000/DRN 06-1029 Revision 301 Section 12.3A.1 EC-5000082374 Section 12.3A.3 Section 12.3A.3.1 Table 12.3A-9 Sh. 1 Revision 303 Section 12.5.2.2.3 EC-14865 Section 12.5.3.7.1 Table 12.5-3 Revision 305 Table 12.2-16 Sheet 3 of 4 EC-4019 Table 12.2.16a Sheet 3 of 4 Section 12.5.1.3 EC-27665 Section 12.5.2 Section 12.5.3 Table 12.5-1 Table 12.5-2 Table 12.5-3 Table 12.5-4 Revision 306 Section 12.3.4.1.3 EC-12329 Section 12.3.1.4 EC-14275 Section 12.3.1.9 Section 12.5.3.2.4 Revision 307 Section 12.3A.3.2 EC-30976 Table 12.3A-9 Sheet 1 of 2 Section 12.5.1.1 LBDCR 13-005 Section 12.2.1.10 LBDCR 13-009 Table 12.2-11 Sheet 1 of 6 thru Table 12.2-11 Sheet 6 of 6 Table 12.3-1 Table of Contents LBDCR 13-010 Section 12.3.2.2 Section 12.2.1.11 Figure 12.3-8a Table 12.3-1
WSES-FSAR-UNIT-3 12-ix Revision 310 (12/17)
UPDATE REFERENCE LIST Chapter 12 Section Cross References Revision 308 Table of Contents LBDCR 14-007 Table 12.2-12 Sh. 1 of 2 Table 12.2-12 Sh. 2 of 2 Section 12.2.1.12 LBDCR 13-020 Section 12.2.2 Revision 309 Section 12.1.3 LBDCR 16-016 Section 12.1.4 Section 12.2.1.8 Section 12.3.1.9 Section 12.5.3.1 Table 12.5-3 Table 12.2-18 Sheet 1 of 2 LBDCR 15-018 Table 12.2-18 Sheet 2 of 2 Revision 310 Section 12.5.3.6 LBDCR 16-062
WSES-FSAR-UNIT-3 12.1-1 Revision 13 (04/04) 12.0 RADIATION PROTECTION This chapter provides information on the radiation protection features of the plant facility and equipment design, methods employed to achieve such protection, and estimated occupational radiation exposures to operating and construction personnel during normal operation and anticipated operational occurrences.
12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS Waterford 3 has committed to a policy of ensuring that occupational radiation exposures be kept as low as reasonably achievable. This commitment has been incorporated into the design of the plant and will continue throughout plant operation.
To ensure that the commitment would be incorporated into the plant design, Waterford 3 instituted an independent review of the shielding design and analysis performed by Ebasco. Additionally, Waterford 3 conducted an ALARA design review to identify those areas in plant layout, system design, etc. that may be of an ALARA concern, and will conduct an ongoing review and correct discrepancies to the extent possible.
Comments and information exchange, together with similar comments and information received from other clients and the industry, have been used by the architect/engineer to formulate a set of guidelines for the design of the Waterford 3 plant to minimize personnel exposures. These guidelines were prepared and applied by Ebasco engineers.
It is significant to note that these guidelines, summarized in Subsection 12.1.2 and cross referenced to the appropriate Regulatory Guide 8.8 section, paraphrase to a large extent Regulatory Guide 8.8, Information Relevant to Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable (Nuclear Power Reactors) March 1977, even though they preceded it. The guidelines have since been updated to meet the intent of Regulatory Guide 8.8. Waterford 3 will follow the guidelines of Regulatory Guide 8.10, Operating Philosophy For Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable, September 1975.
¨(DRN 03-1135, R13)
The Radiation Protection Manager has the specific responsibility and authority for ensuring that the radiation protection program maintains exposures as low as reasonably achievable. He reports to the Technical Services Manager, and has a direct line of communication with the nuclear operations supervisors.
(DRN 03-1135, R13)
The Vice President-Operations, Waterford 3 is ultimately responsible for establishing and implementing the ALARA program.
12.1.2 DESIGN CONSIDERATIONS The following design guidelines have been used in the analysis of the plant to obtain as low as reasonably achievable personnel exposures.
WSES-FSAR-UNIT-3 12.1-2 Shielding walls have been designed so that the highest dose rate on the outside surface of the wall is less than or equal to the maximum permissible dose rate as per 10CFR20, considering occupancy requirements. For instance, in a controlled area of unrestricted occupancy, the design dose rate at the "hot spot" in the shielding wall would be at most 2.5 mrem/hr. The thickness required to achieve this desired dose rate is maintained throughout the extent of the particular wall (i.e., shielding walls are not contoured). (RG 8.8 paragraphs C.2.b.2 and C.2.b.3).
To the extent possible systems and components handling high activity fluids are located in the same general area of the plant, taking into account separation criteria. (RG 8.8 paragraphs C.2.b.2 and C.2.b.3).
Such systems and components are separated from low activity systems and components; in turn, the latter are located away from "clean" systems and components. With this arrangement, heavy shielding walls can be shared by various components, thus minimizing space requirements and costs. (RG 8.8 paragraphs C.2.b.2 and C.2.b.3).
To the extent possible all components and piping which do not normally contain radioactivity, nor can be expected to ever become radioactive, are separated from the radioactive portions of the plant. This simplified division of the plant into controlled and uncontrolled areas, aids in the unimpeded traffic within both portions, reduces the possibility that radioactive piping is run in clean areas, minimizes the need for shielded pipe chases, and helps in controlling contamination spread into clean areas. (RG 8.8 paragraphs C.2.b.2 and C.2.b.3).
Components belonging to a given system handling radioactivity are generally arranged on the same side, but outside of, corridors. This was done to minimize piping runs in corridors, and thereby to minimize the use of shielded pipe chases. (RG 8.8 paragraph C.2.b.6).
Equipment and components which require manual operation, visual inspection or are expected to need servicing are arranged in the lowest possible radiation field. For example, a system handling radioactive fluid consisting of a tank, pump, associated valving, sampling lines, and instrumentation is laid out as follows:
The tank, requiring the least servicing, is normally placed in a separate shielded cubicle. Serviceable valving and piping is excluded from this cubicle to the maximum practical extent. The pump and valves, which require maintenance, are placed in a separate cubicle. If practicable or when required, further compartmentalization is achieved by placing the pumps and valves in their own individual cubicles.
Sampling lines and instrumentation requiring personnel attendance are brought outside the shield walls to low level radiation zones (II or III), with the exception of inside the containment. A description of each radiation zone is provided in Subsection 12.3.2.2. Even within the containment, such instrumentation is brought to the lowest possible radiation area. (RG 8.8 paragraphs C.2.b.1, C.2.C.2 and C.2.i.5)
Where it was impractical to locate items requiring servicing in low radiation areas, such items were designed so that they can be moved to a low radiation area. Any problems which may occur with rapid removal, local flushing, and decontamination for pieces of equipment (i.e., small pumps, large pumps, etc.) will be reviewed during operation. (RG 8.8 paragraphs C.2.b.9 and C.2.i).
WSES-FSAR-UNIT-3 12.1-3 Revision 11 (05/01)
Sufficient clearance is provided within shielding cubicles housing components potentially requiring maintenance and repair (such as valves, pumps heat exchangers, etc.), so that unimpeded and efficient work on the particular component is allowed. Overly restrictive compartments while saving space, require lengthier stays by maintenance or repair personnel, since their work will be hampered and inefficient.
Furthermore, large compartments will ease installation of temporary shielding barriers should they be re-quired. Provisions for a sling, chain, and hoist, etc. are provided for all pieces of equipment with consideration given to their size, expected frequency of maintenance, and handling problems. (RG 8.8 paragraphs C.2.b.8 and C.2.i).
All instrumentation (flow meters, level gauges) is located in the lowest practical radiation area and will be readily accessible to operators. Operators will not normally be required to enter cubicles housing radioactive components to read or activate instruments. The only exception is instrumentation located by necessity within the containment; and even there, all efforts are made to locate it in the lowest possible radiation field. (RG 8.8 paragraph C.2.b.2).
Radioactive piping is either run in shielded pipe chases or within shielded cubicles housing low maintenance equipment, where practical. Radioactive piping is not run in accessways, and the amount of radioactive piping near frequent maintenance equipment has been minimized. (RG 8.8 paragraph C.2.b.6).
To the largest extent possible, but especially for components and piping handling primary coolant, radioactive resins, and concentrates, connections are provided for flushing portions of the system. The portion of the system to be flushed is dictated by the expected frequency of maintenance of the component(s) housed in a shielded cubicle, the size of the cubicle, the number of components housed therein, the geometry of piping, and the valving arrangement. (RG 8.8 paragraphs C.2.f.3 and C.2.h).
The piping diameter utilized for resin and sludge transfer lines has been sized to minimize plugging. Other provisions such as flushing of the lines after resin or sludge transfer also contribute to the minimization of the plugging problem. We do not believe that oversizing of the piping would significantly decrease the potential for plugging, but would adversely affect the processing function, and possibly worsen the plugging situation by virtue of the resulting lower flow velocities.
In general, all components and piping within a single cubicle are flushable. Flushing with demineralized water will be the preferred manner, since other water or solutions may introduce chemicals which may, if activated, result in further radioactive crud. Flexible hoses from the low drain point to the floor drain are considered acceptable for flushing procedures except for resin carrying lines. These are flushed back to their respective tanks (spent resin and dewatering tanks). If hose is used to flush, the flushing connection will be brought out to a relatively accessible area. (RG 8.8 paragraph C.2.h).
Labyrinths and/or shielding doors are used to eliminate radiation streaming through access openings to the shielded cubicles. (RG 8.8 paragraph C.2.b.4).
WSES-FSAR-UNIT-3 12.1-4 Penetrations for piping and ducts in shielding walls are designed so as not to be on a direct line with a major radioactive source. Openings for the penetrations are kept as small as possible. Packing of the opening will be done, when required, to meet exposure criteria. (RG 8.8 paragraph C.2.b.5).
The extent and degree to which the two following guidelines concerning pumps and valves are used are dependent on the expected radiation levels of any given system. More stringent adherence to these guidelines is paid for highly radioactive systems, with progressive relaxation paralleling a lessening in expected activity.
Pumps serving potentially radioactive systems are housed in shielded areas outside cubicles containing radioactive components. Radioactive piping within the shielded pump cubicle is kept to a minimum. (RG 8.8 paragraph C.2.b.1).
Shielded valve stations for systems handling radioactive fluids are employed, wherever it is feasible, in order to perform valve maintenance without drainage of associated equipment. To further minimize personnel exposure remotely operated valves are utilized where practical and necessary. If manual valves are employed, extension rods through a shield wall to an accessible low radiation area are utilized as necessary. In order to greatly decrease the problems of radiation streaming, the reach rod penetrations are generally offset from the major source of radiation (usually a tank), or are provided with an internal offset. (RG 8.8 paragraphs C.2.c.1 and C.2.i.5).
Equipment and piping which handle radioactive fluids are designed, as indicated below, in a manner conducive to reducing the retention of radioactive crud, and making decontamination easier and efficient.
(RG 8.8 paragraphs C.2.e, C.2.f.3 and C.2.h).
a)
Piping runs are sloped wherever possible to prevent accumulation and assist in the removal of radioactive crud deposits.
b)
The number of elbows, tees, Vs, deadlegs, standpipes, etc. are minimized since these act as crud traps, and also render decontamination difficult.
c)
Where elbows are required, large radius elbows are employed if possible with flow moving down through a vertical elbow, rather than up, as the elbow would then act as a crud trap. Flat-bending of pipes is used when possible.
d)
Orifices are installed in vertical runs as opposed to horizontal ones, when there is an option.
e)
Horizontal piping expansion joints are preferentially used.
If vertical expansion joints are required, configurations resembling traps (e.g., filled joints) are avoided.
f)
For pipes carrying resins, a smooth interior finish is specified, or the pipe is lined with a suitable polymer.
g)
Consumable inserts are employed at welds, where this technique is possible. Use of backing rings will be minimized.
WSES-FSAR-UNIT-3 12.1-5 Revision 11 (05/01) h)
Low leakage valves are employed. ALARA design considerations have included use of low leakage valves with back seats wherever possible. The type of valve and valve arrangement generally used requirements of the system to which the valve belongs.
(DRN 99-0815) i)
All valve packing glands have provisions to adjust packing compression to reduce leakage.
Valves in highly radioactive systems such as Waste or Boron Management Systems are packless diaphragm valves or are provided with stem below seals to reduce leakage (RG 8.8 paragraph C.2.i.6).
Use has been made of radiation resistant seals and gaskets when practical.
Preferential use is made of round bottomed tanks and vessels to minimize crud buildup in the tank bottom. Effluent process lines are placed as close as possible to the bottom, preferably at the lowest point in tank. Drain valves are positioned away from the tank bottom, and preferably located in an accessible area. Operators will not normally be forced to crouch below tanks to operate any valve, and to receive high exposures from the crud deposited therein. (RG 8.8 paragraph C.2.i.7).
The plant ventilation system is designed so that air flow is from areas of low to higher potential airborne contamination. (RG 8.8 paragraph C.2.d.1).
The operation of the Solid Waste Management System is discussed in FSAR Section 11.4.
Single traps are not used in the floor drainage system. This subsystem is entirely free of any radioactive noble gases because radioactive halogens are in solution. Hence no transport of radioactive gases from one cubicle to another through the floor drain system is foreseen. A single trap between the drain tanks and the drain header is deemed sufficient to prevent transport of any radioactive volatiles evolving in those tanks. (RG 8.8 paragraph C.2.b.10).
Equipment drains selectively employ a minimum number of traps to prevent transport of radioactive gases from component to component through the equipment drainage system, which could result in noble gas releases when components are maintained or repaired. Typically, the traps are at the equipment sump.
(RG 8.8 paragraphs C.2.b.10 and C.2.d.6).
Open sumps are not used as receivers of equipment drains. (RG 8.8 paragraph C.2.d.6).
The plant areas are divided into two zones, controlled and uncontrolled, separated by access control points. (RG 8.8 paragraph C.2.a).
The controlled zones encompass all areas which either house radioactive equipment, or which can become contaminated during movement of personnel or components. The checkpoint acts as the last point at which contamination is detected. All efforts are made to control contamination at its point of origin by posting the extent of contamination and following the normal health physics procedures. (RG 8.8 paragraph C.2.a).
WSES-FSAR-UNIT-3 12.1-6 Revision 11-B (06/02)
Sufficient space is provided at shielded cubicles exits to accessways, to place bins to collect protective clothing.
Proper provisions to control access to high radiation areas, in compliance with Section 20.1601 of 10CFR20, are provided. (RG 8.8 paragraph C.2.a).
Communication outlets are strategically located in the plant. Typically, they are near cubicles housing radioactive equipment (especially high radiation areas) or equipment which operates intermittently and which is actuated remotely.
Equipment decontamination areas are provided at selected locations in the plant. The choice of these locations is based on:
a) a low radiation area, b) a location central to equipment which is likely to require decontamination, c) a vicinity near the hot machine shop, and/or d) the capability of routing drain and ventilation lines from the decontamination area to processing systems. The equipment decontamination facility is located in the +21 west wing area adjacent to the Hot Machine Shop. Other temporary areas may be established to expedite equipment decon.
Each decontamination area consists of the decontamination area and a storage area, where equipment can be stored prior to decontamination.
Suitable coatings are used on all floor surfaces of areas in the controlled zone. Personnel decontamination facilities and shielded cubicles have surfaces suitably treated for easy decontamination.
The preceding guidelines were derived utilizing experience from prior plant design and from operating plants.
Note: The following is historical information pursuant to NEI-98-03, which is identified by a designation of "Start" and "End."
Start of Historical Information These guidelines were used in the design of the facility. The design was reviewed by the Waterford 3 and Ebasco engineers assigned to the project. These engineers, besides performing shielding and exposure calculations, worked with the designers and other engineers to ensure that the guidelines were followed.
They advised on the most desirable design option for radiation protection when alternate designs were possible in satisfying the process requirements. Typically, this involved a study of the exposures which were likely to be derived from alternate designs and the selection of the option resulting in the lowest exposure.
The lead applied physics engineer of Ebasco was responsible for the radiation protection design and review.
He originally established the general and specific objectives of the shielding, access, and contamination control design of the project.
He established these objectives in concert with Waterford 3 cognizant personnel. He provided other disciplines involved in the plant design with guidelines and directives which they followed and which were intended to result in ALARA exposures. He, and the engineers assigned to him, monitored the implementation of the guidelines and directives.
WSES-FSAR-UNIT-3 12.1-7 Revision 11-B (06/02)
The lead applied physics engineer was responsible for recommendations to the project engineer, who was responsible for the overall design of the power plant. In such recommendations he requested concurrence of the cognizant engineers assigned by Waterford 3 to review his work.
The radiation protection design review was an ongoing review throughout all phases of the design. The radiation protection cognizant personnel worked side by side with the other discipline engineers and designers to ensure that all necessary radiation protection considerations were taken into account. In addition, more formal shielding reviews were conducted at selected intervals during the design of the project. The formal reviews were performed to ensure that the guidelines stated above were not compromised to ensure ALARA dose rates. Waterford 3 had direct input in these reviews. Calculations, which form the bases for such reviews, were made on the basis of the information available at that point in time in the following categories:
a) source calculation and/or verification for each system involved in the review, b) shield thickness calculations, c) activation calculations if the system involves neutron radiation, and/or d) mapping of dose rate calculation of exposure for the particular layout.
Comments from the engineer performing the review or from Waterford 3 were transmitted to the appropriate discipline either by separate memorandum which refered to the particular drawing or specification, if the comments were very extensive, or were made directly on the specifications or drawings. In either case, the drawing or specification transmittal sheet would bear a note that comments were made which required resolution. Both the transmittal sheets and the comments were retained by the discipline affected by the comments. The success of the normal review for radiation protection design was measured by the paucity of instances where a design had to be modified or improved after it has been completed. While certainly these instances have occurred in the Waterford 3 project, the objectives of the review was indeed to minimize them by effecting whatever changes or improvements were deemed necessary during the conceptual phase of the design, prior to the design being released. As a result of this approach, depiction of many illustrative examples of specific design improvements which resulted from the radiation protection design review is difficult, without documentation of the day by day interaction of the radiation protection engineer with the designer, and the evolution of the radiation protection design.
Most of these examples are documented by comparison of areas and systems at different degrees of completion of the design, with the same areas and systems when the design is complete.
Many changes and improvements were made to the Waterford 3 design to minimize personnel exposures.
Since the finished drawings and specifications reflect only the finished product, listed below are just a few of the major changes and improvements:
NOTE: End of Historical Information.
WSES-FSAR-UNIT-3 12.1-8 Revision 13 (04/04) a)
The ion exchanger area was redesigned to provide a shielded valve gallery and a location from which the valves could be operated remotely.
b)
The Solidification System was originally designed in the drumming station. The System has been abandoned in place and the drumming station converted to the Hot Tool Room.
c)
The filter cubicles were redesigned to a blockhouse configuration for simpler handling of the filter cartridges with a concomitant decrease in personnel exposure.
d)
The reactor vessel cavity and ring girder support were redesigned to minimize the air activation (Ar41)in containment, and the neutron streaming problem. With the present design, neutron dose rates calculated at similar locations in the containment are comparable to those which have been predicted to occur after installation of special shields above the reactor flange level. Such shields, however, would require removal for refueling operations with the attendant personnel exposures. The present shield design which utilizes the support ring girder is considered optimum in that the man-rem saved by installation of an additional above-the-flange shield, estimated to be two man-rem/yr, on the basis of a dose rate reduction of approximately 150 mrem/hr at an average containment location requiring access and an occupancy of 15 minutes/week, is comparable to the increase in man-rem necessitated for the removal of the above-the-flange shield, which has been estimated to be between 0.6 and 1.2 man-rem. The cost of an additional shield would be inordinate with respect to the man-rem saved.
e)
Permanent shielding has been installed on the spent resin recirc/discharge line in the drumming station and on the spent resin tank discharge line in the spent resin tank pump room -35 elevation RAB.
12.1.3 OPERATIONAL CONSIDERATIONS To the greatest extent, all efforts are made to factor operational considerations of radiation exposure in the plant layout and system design, by utilizing the guidelines of Subsection 12.1.2 throughout the design effort.
These guidelines incorporate known operational considerations derived from experience. Information from operating plants have continuously been factored in the design as it progressed to reflect new operational considerations, not well known or appreciated at the time the guidelines were initially developed. In this regard, the guidelines of Regulatory Guides 8.8 and 8.10 will continue to be factored in the design.
Administrative procedures will be established in the plant, which along with design shielding, will ensure that the exposures to personnel will be kept as low as reasonably achievable during plant operation and maintenance. These procedures will be completed prior to initial plant startup, and will be updated to reflect Waterford 3s operating exposure experience.
Stationary and portable detectors are provided for monitoring direct, surface and airborne activity in operating and maintenance locations, particularly in those locations susceptible to radioactive contamination.
Detailed written procedures, shall be prepared and approved as specified in Section 12.5.
¨(DRN 03-1135, R13 Implementation of these procedures is the specific responsibility of the Radiation Protection Manager.
(DRN 03-1135, R13
WSES-FSAR-UNIT-3 12.1-9 Revision 309 (06/16)
The Health Physics program and a detailed description of procedures for radiation exposure-related operation is given in Section 12.5. Summarized below, however, are the salient points.
(LBDCR 16-016, R309)
Health physics personnel will periodically observe jobs in progress in the radiation controlled areas and will perform radiation surveys to ensure that exposure to radiation and contamination levels are kept ALARA. Administrative exposure guidelines are designed to evenly distribute each individual's exposure.
When personnel are assigned to a job or a location where there exists the possibility that administrative guidelines may be exceeded, health physics personnel shall investigate the exposure records of the personnel involved and authorize work by such personnel only after the current exposure history, the amount the guidelines might be exceeded, and the alternatives that are available to complete the job under consideration have all been considered. This will call for consultation and pre-job planning between the supervisor in charge of the job and health physics personnel.
(LBDCR 16-016, R309) 12.1.4 DECOMMISSIONING CONSIDERATIONS While decommissioning will occur only after the termination of plant operation (the estimated facility lifetime is 40 years from the date of the construction permit), it is expected that it will be accomplished through the application of one of the presently available alternative methods. The experience gained in the continued application of these methods, and any developing variations, to nuclear plant decommissionings in the interim years will further ensure that ultimate acceptability of the mode of decommissioning.
At present there are three primary methods for decommissioning commercial power reactors -
mothballing, in-place entombment, and removal of radioactive components and dismantling. The major characteristics of each of these methods are described below. A fourth alternative - conversion to a new nuclear system or a fossil fuel system may prove practicable at selected sites given favorable conditions.
This alternative utilizes the existing turbine-generator system with a new steam supply system. The original Nuclear Steam Supply System is separated from the Electric Generating System and disposed of in accordance with one of the three primary methods. This alternative is not treated separately here since with respect Nuclear Steam Supply System decommissioning it is not distinct from the primary methods.
Mothballing of a nuclear reactor facility consists of putting the facility in a state of protective storage. In general, the facility may be left intact except that all fuel assemblies and the radioactive fluids would be removed from the site. Adequate radiation monitoring, environmental surveillance, and appropriate security procedures would be established to ensure that the health and safety of the public is not endangered.
In-place entombment consists of sealing all the remaining radioactive or contaminated components (e.g.,
the pressure vessel and reactor internals) within a structure integral with the biological shield. All fuel assemblies, radioactive fluids and wastes, and certain selected components would be shipped offsite.
The sealing structure must provide integrity over the period of time in which significant quantities of radioactivity remain with the material in the entombment. An appropriate and continuing surveillance program would be required.
WSES-FSAR-UNIT-3 12.1-10 Revision 309 (06/16)
(LBDCR 16-016, R309)
In the removal/dismantling method, all fuel assemblies, radioactive fluids and waste, and other materials having activities above accepted unrestricted activity levels would be removed from the site. The utility would then have unrestricted use of the site and long-term surveillance would not be required. In the extreme application of this method, the utility may desire to dismantle the remainder of the facility and remove or otherwise dispose of all structural material and components.
(LBDCR 16-016, R309)
To date, experience with decommissioning of civilian nuclear power reactors in the United States includes the shutdown or dismantling of six facilities. In these decommissionings each of the three primary methods described above have been employed. The Carolina Virginia Tube Reactor (CVTR) and Pathfinder Reactor decommissionings are examples of the mothballing method, while the Hallam Nuclear Power Facility, Boiling Nuclear Superheater (BONUS) Power Station, and Piqua Reactor decommissioning is most nearly exemplary of application of the removal/dismantling technology.
Although the sizes of the facilities decommissioned to date have been relatively small, the experience reinforces the conclusion that the facility can be decommissioned while ensuring that in-plant exposures are kept ALARA.
Whichever mode of decommissioning is used, there is assurance that the operation could be performed in a manner consistent with the ALARA philosophy. The assurance stems from the fact that the facility has been designed in accordance with the philosophy outlined in Regulatory Guide 8.8. Since operating conditions present much more limiting radiological conditions than shutdown and decommissioning, it is anticipated that all exposures associated with decommissioning will be ALARA.
(DRN 03-2066, R14)
Estimates of the in-plant and offsite exposures for a reference PWR has been performed by Battelle Northwest Institute. These estimates reveal that the total exposures could range from 1300 man rem for immediate dismantlement to about 500 man rem for preparing the facility for safe storage 30 years and then dismantling. Figures 12.1-1 and -2 present the results of the BNW study. These results are based on the Trojan Nuclear Power Plant as the reference plant. Since Waterford 3 is a more recently designed plant and incorporates many of the design features recommended in Regulatory Guide 8.8, it is reasonable to assume that the exposures would be lower than predicted.
(DRN 03-2066, R14)
As described in the above subsections specific designs and policies are provided to keep in-plant exposures ALARA. Subsection 12.1.2 presents a detailed description of the design considerations. The following summarizes the design features which would be effective in reducing in-plant exposures during decommissioning.
a)
Components containing high activity, low activity and no activity are physically separated.
b)
Shield walls are not contoured to the radiation field of a particular component, instead the shield walls for a particular component are of uniform thickness based on the highest dose rate at any point.
c)
Sufficient spacing is provided within cubicles to facilitate inspection and maintenance and to allow the installation of portable shielding.
d)
Radioactive piping is either run in shielded pipe chases or within shielded cubicles.
WSES-FSAR-UNIT-3 12.1.11 e)
Components and piping containing radioactive material, such as primary coolant, resin and concentrates are provided with connections for flushing with demineralized water.
f)
Equipment and piping are designed to minimize crud buildup and facilitate decontamination. For example pipes are sloped, elbows and Ts in pipes are minimized, round bottom tanks are used.
g)
Plant ventilation is designed so that air flow is from low to potentially higher areas of airborne activity.
h)
Several equipment decontamination areas are provided. The ventilation and drainage from these areas can be routed for processing.
i)
Sampling lines and instrumentation requiring personnel attendance are brought outside shield walls.
J)
Surfaces are suitably coated for ease of decontamination.
k)
In order to reduce neutron activation of materials, shielding is specifically provided to reduce neutron streaming.
l)
Filtration is provided on major HVAC exhaust points. This includes filtration on the large containment purge and the exhaust from the Reactor Auxiliary Building. In addition the Airborne Radioactivity Removal System (i.e., a kidney system) is provided inside the containment.
WSES-FSAR-UNIT-3 12.2-1 12.2 RADIATION SOURCES 12.2.1 CONTAINED SOURCES The radiation sources used for the design and analysis of the shielding requirements are based on the design of plant operation including full power operation, shutdown conditions, refueling operations, and for various postulated accidents. They include the neutron and gamma fluxes outside the reactor vessel, the reactor coolant activation, fission and corrosion product activities, deposited corrosion product sources on reactor coolant equipment surfaces, spent fuel handling sources, and postulated core meltdown sources.
In addition, radiation sources for various auxiliary systems are also tabulated.
12.2.1.1 Reactor Coolant Fission and Corrosion Product Activity The activity utilized for shielding calculations is shown in Table 11.1-2.
The basis for the above reactor coolant radioisotope concentrations is specified in Table 11.1-1, and the manner in which the concentrations have been derived is given in Subsection 11.1.1.1.
12.2.1.2 Neutron Fluxes Outside the Reactor Pressure-Vessel The maximum neutron spectra during full power operation, divided into 10 energy groups, for points on the top and bottom of the reactor pressure vessel along the core centerline, and a point on the side of the vessel adjacent to the maximum axial power, are shown in Table 12.2-1. The neutron spectra in this table include the neutrons scattered from the concrete cavity wall and are used to determine the thickness of the primary shield wall.
To determine the neutron flux streaming up the annulus between the reactor vessel and the reactor vessel cavity wall, a better definition of the neutron spatial and angular fluxes emergent from the vessel is developed by the computer program DOT(1).
The calculation utilizes S4 quadrature, 22 neutron energy groups, and 16 axial intervals from the core bottom to the upper guide structure. A vacuum boundary is used at the outer radius.
The DOT generated data, summarized for convenience in the form of neutron leakage rates from axial intervals spaced along the vessel surface, is shown in Table 12.2-2. The equivalent scalar fluxes can be obtained by division of the leakage rates in the Table 12.2-2 by the surface area of the region.
12.2.1.3 Gamma Fluxes at Full Power Operation The maximum gamma spectra during full power operation, divided into 14 energy groups, for points on the top and bottom of the vessel along the core centerline, and a point for the side of the vessel adjacent to the maximum axial power, are shown in Table 12.2-3.
WSES-FSAR-UNIT-3 12.2-2 Revision 14 (12/05) 12.2.1.4 Reactor Coolant N-16 Activity
(DRN 03-2066, R14)
The N-16 activity in the reactor coolant which determines the shielding requirements for the secondary shield wall and portions of the Chemical and Volume Control System (CVCS) is discussed in Subsection 11.1.3.1. The N-16 activity at various points in the primary system, as well as the CVCS, is determined from the given activity at the reactor outlet nozzle, taking into account the decay due to the transit time between the nozzle and the point of interest as calculated on the basis of the maximum flow rate and changes in density caused by cooling.
(DRN 03-2066, R14) 12.2.1.5 Reactor Coolant System Sources at Shutdown Following shutdown, residual radiation from the Reactor Coolant System is due to fission product decay gamma radiation emanating from the core; material activation sources, radioactive corrosion products which have deposited on surfaces, and the fission and corrosion products in the reactor water. The reactor vessel maximum decay gamma spectra, divided into eight energy groups for various times after shutdown, are shown in Table 12.2-4 for a point on the side of the vessel adjacent to the maximum axial peak.
The material activation spectra at the same location and decay time is shown in Table 12.2-5. The material activation sources include contributions from the vessel wall, barrel, shroud, and primary coolant water. Activation of the vessel insulation and supports is considered and evaluated on the basis of the fluxes computed to exist in the annular cavity space during full power operation.
The fission and corrosion product reactor coolant inventory assumed to be present for shutdown is that specified in Table 11.1-2, corrected for decay up to the point in time of interest.
The activity of radioactive crud and its thickness on Reactor Coolant System surfaces have been evaluated using data from six pressurized water reactors. For the circulating crud, the observed activities were compared to activity values recommended for design by the Draft ANSI N237 Source Term Specification, 1976.
(DRN 03-2066, R14)
(DRN 03-2066, R14)
Table 11.1-10 shows the expected maximum specific activities for deposited corrosion products. The residual activity due to deposited corrosion products is evaluated with the assumption of a thickness of 0.16 mg. crud/cm2 for steam generator tubing and a 1.5 mg. crud/cm2 for piping and system crud level.
(DRN 03-2066, R14)
The current FSAR design basis source terms are based on draft ANSI standard N237. Extended Power Uprate used ANSI/ANS-18.1-1999 as the basis for source terms. An evaluation of the two source terms and the change in the flux-to-dose conversion factors between ANSI 6.1.1 1977 and ANSI 6.1.1 1991 indicate that the EPU source terms are bounded by the current FSAR design basis source terms.
(DRN 03-2066, R14) 12.2.1.6 Pressurizer Activity The liquid section of the pressurizer has source terms due to plateout of radioactive crud plus dispersed fission and corrosive product activity in the water. The maximum deposited activity for the liquid section, assuming a crud film thickness of 1.5 mg. crud/cm2, is derived from the values of Table 11.1-10. As an upper limit, the liquid section will have
WSES-FSAR-UNIT-3 12.2-3 Revision 309 (06/16) dispersed fission and corrosion product activity equal to the primary coolant activity with one percent failed fuel specified in Table 11.1-2.
The activity in the steam section is due to the buildup of the gaseous fission products Xe and Kr from one percent failed fuel. This activity is shown in Table 12.2-6.
12.2.1.7 Contained Sources in Other Plant Systems The source intensity in equipment and pipelines handling radioactive fluids is determined from activity in the reactor water by considering the processes that the reactor water has undergone prior to entering equipment and piping (dilution, filtering, demineralization, delay, change of phase, etc.)
In all cases the process or combination of processes leading to the highest activity is considered for conservatism. The maximum inventory of activity in the various components of the Chemical and Volume Control System, Boron Management System, Fuel Pool System, Safety Injection System and Waste Management System are listed in Tables 12.2-7, 12.2-8, 12.2-9, 12.2-10 and 12.2-11, respectively.
12.2.1.8 Spent Fuel and Spent Fuel Pool The maximum and expected fission and corrosion product activities in the spent fuel pool are specified in Table 11.1-17.
The source terms employed to determine the minimum water depth above spent fuel and shielding walls around the spent fuel pool, as well as shielding of the spent fuel transfer tube, are given in Table 12.2-12.
The activities shown in that table are equilibrium core activities based on 105 percent of full power core conditions.
(LBDCR 16-016, R309)
The spent fuel pool contains greater than 8 cores as described in Subsection 9.1.2.
(LBDCR 16-016, R309)
(DRN 99-2362, R11)
To size the shield for the transfer tube, the hottest element is assumed to be transferred. Its activity is that computed from the assumption of a homogenous core inventory (Table 12.2-12) and an overall peaking factor of 1.80.
(DRN 99-2362, R11) 12.2.1.9 Accident Sources (DRN 03-2066, R14)
The accident source terms which are employed to determine shielding requirements for emergency accessways and containment shielding, as well as potential doses to equipment inside containment following a loss-of-coolant accident (LOCA) are shown in Table 12.2-13. Table 12.2-13 assumes a release to containment of the activity stated in TID 14844(2), namely 100 percent noble gasses, 50 percent halogens, and 1 percent remaining fission product inventory.
The accident sources for the main control room are shown in Table 15.6-18.
(DRN 03-2066, R14)
The accident sources for the Safety Injection System are shown in Table 12.2-10.
(LBDCR 13-009, R307) 12.2.1.10 Low Level Radioactive Waste Storage Facility The Low Level Radioactive Waste Storage Facility (LLRWSF) was added as part of LDCR 95-0059 for storage of radioactive waste awaiting disposal.
The radiation protection design of the LLRWSF, in terms of shielding and dose estimates is based upon the dose rates of the waste containers in the facility when the facility is filled to capability. Table 11.4-11 includes anticipated waste volumes by type for a fully loaded facility. The design basis radiation levels for the wastes to be stored in the facility are based on Waterford 3 historical source terms.
(LBDCR 13-009, R307)
WSES-FSAR-UNIT-3 12.2-4 Revision 307 (07/13)
(LBDCR 13-009, R307)
The radiation shielding configuration of the LLRWSF is designed in accordance with the dose rate criteria per 10CFR20, FSAR and site procedures for both the site boundary and for restricted areas. The nearest site boundary to the LLRWSF is River Road which is approximately 980 ft away. The facility is designed so that the maximum offsite dose rate at the site boundary from the waste stored in the LLRWSF will be maintained < 0.05 rem/yr which is 10% of the allowable. The restricted area is the area surrounding the LLRWSF which will be controlled by the Radiation Protection Group for purposes of protection of individuals from exposure to radiation and radioactive materials. The facility is designed so that the maximum dose when the facility is fully loaded at approximately 60 ft. away is 0.3 mrem/hr.
The projected waste containers curie content is listed in Table 12.2-11. The location of the LLRWSF is shown in Figure 1.2-2.
(LBDCR 13-009, R307)
(LBDCR 13-010, R307) 12.2.1.11 Original Steam Generator and Reactor Vessel Head Storage Facility In support of the Waterford 3 (W3) SG/RVCH Replacement Project, the Original Steam Generator Storage Facility (OSGSF) was added as a part of Steam Generators / Reactor Vessel Head replacement project. The Original Steam Generator Storage Facility (OSGSF) is located outside of the Protected Area and within the site Owner Controlled Area. The OSGSF is located well away from any safety-related onsite systems, structures, or components. The OSGSF is designed as a non-safety related structure to be used to provide secure storage until site decommissioning of the two Original Steam Generators (OSGs), the Original Reactor Vessel Closure Head (ORVCH) and Original Control element Drive Mechanisms (OCEDMs) that were removed and replaced during the W3 SG/RVCH Replacement Outage.
The location of the OSGSF is shown on FSAR Figure 1.2-2.
The movements of the OSGs, ORVCH and OCEDMs into the OSGSF were accomplished through an open side to the building facing Plant North towards Warehouse 5B. The closure of these openings has been accomplished by means of pre-cast concrete panels sealed to the OSGSF walls that minimize airflow through the panel joints.
The design of a reinforced concrete floor accounted for supporting the loads anticipated during facility construction and offloading and storage of the OSGs, ORVCH and OCEDMs. The floor slab was elevated two (2) feet above finished grade of 16 so that surface water could drain away from the building.
The OSGSF floor was coated to create a barrier to prevent leaching of contamination into the floor.
The 24 wall concrete thickness has been established to allow unrestricted personnel access to the outside of the building in accordance with the NRC unrestricted access criteria found in 10 CFR 20. The OSGSF roof will not be accessed by members of the public, and thus, the 10 CFR 20 dose limit of 100 mR/yr is not applicable to the OSGSF roof.
The OSGSF is designed to provide adequate clearance between the stored components and wall surfaces to permit personnel to visually inspect these areas if needed.
The perimeter fence is equipped with a lockable gate which is controlled by Radiation protection. The dose rate at the perimeter fence is as depicted on Table 12.3-1. The OSGSF is designed for adequate flood protection for the OSGs, ORVCH, and OCEDMs. The OSGSF is not designed for occupation except for the capability to perform inspections. Accordingly, the interior of the OSGSF is not required to meet emergency egress requirements or to have installed ventilation systems. Similarly for the purpose of long term storage of the OSGs, ORVCH and OCEDMs, which requires no occupancy of the facilities by personnel, the OSGSF is not designed for any water, wastewater, electrical, or telephone services.
The building concrete shielding design meets the radiological requirements of 40 CFR 190, 10 CFR 20 and plant license requirements, and provides adequate shielding to limit the contact dose rate to 0.1 mR/hour on the exterior wall surface of the building.
(LBDCR 13-010, R307)
WSES-FSAR-UNIT-3 12.2-5 Revision 308 (11/14)
(LBDCR 13-010, R307)
The OSGSF roof is equipped with watertight membrane roofing system to preclude moisture instrusion in the storage facility.
The floor sumps with an external access are provided as common collection point for any liquid within the OSGSF. The sampling port will be used to sample any inventory that collects within the sump, as well as to remove such inventory without requiring entry into the OSGSF.
Two ground water sampling wells have been added down gradient from the OSGSF to monitor groundwater impact assuming an unlikely release of contaminated water from the OSGSF. Monitoring will be performed to identify and mitigate radiological contamination that could reach groundwater.
(LBDCR 13-010, R307)
(LBDCR 13-0020, R308) 12.2.1.12 Independent Spent Fuel Storage Installation (ISFSI)
In order to provide adequate spent fuel storage capacity for WF3, Entergy has established an ISFSI at WF3 on a site located south of the four large water storage tanks that are situated at the south end of the WF3 plant area, just west of the switchyard, within the Protected Area. The ISFSI pad is sized to store 72 HI-STORM storage casks, with each cask capable of storing 32 spent fuel assemblies, which is adequate to meet the projected WF3 spent fuel storage needs over the life of the nuclear power plant.
The WF3 ISFSI operates under the conditions of the general license in accordance with 10 CFR Part 72 regulations. The spent fuel storage cask designs that are approved for use under the general license are listed in 10 CFR 72.214. The HI-STORM 100 Cask System has been approved for use, and is listed in 10 CFR 72.214. The design basis for the HI-STORM 100 Cask System is provided in the Final Safety Analysis Report (FSAR) for the HI-STORM 100 Cask System and as supplemented by changes made by Entergy Operations, lnc., the general licensee, from the HI-STORM FSAR under the provisions of 10 CFR 72.48. Amendment No. 5 to the HI-STORM 100 CoC is being used as the licensing basis for the WF3 ISFSI and dry cask storage activities.
ISFSI operations is evaluated under the WF3 ISFSI 10 CFR 72.212 Evaluation Report, which includes the Radiological Evaluation for the ISFSI as required by 10 CFR 72.104.
(LBDCR 13-0020, R308) 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES Equipment cubicles, corridors, and areas normally occupied by operating personnel can contain small amounts of airborne radioactivity as a result of equipment leakage. For the purpose of evaluating the potential exposure to personnel from this activity. This subsection presents a description of the sources of activity and the models and parameters used to evaluate airborne radionuclide levels in the Reactor Auxiliary Building, the Reactor Building, the Fuel Handling Building and the Turbine Building. Table 12.2-14 includes assumptions, parameters and sources of airborne radioactivity used in the analysis. The sources are determined for each area assuming that leakage occurs in that area and that the leaking fluid contains a fraction of the reactor coolant activity. This fraction is determined from process consideration of leaking fluid (amount of filtering, degassing, demineralization etc. prior to leak). The leak rate is based on typical data from operating plants. The equilibrium airborne concentration is then determined by use of the standard equation of build-up and removal, where build-up is caused by leakage and removal both by radioactive decay and ventilation.
The isotopic airborne concentrations as a fraction of maximum permissible concentration in air (10CFR20) were calculated for those areas normally occupied by operating personnel and where a potential for high exposure exists. These values are presented in Table 12.2-15 and indicate that the dose to a critical organ of a worker, adjusted on the basis of weekly occupancy, would be well below the maximum allowable limit. Furthermore, the dose calculated based on these values would be highly conservative because the maximum permissible concentrations (MPCs) given in 10CFR20 are based on infinite cloud assumptions while volumes of the applicable areas are limited.
WSES-FSAR-UNIT-3 12.2-6 Revision 308 (11/14)
(DRN 99-2362, R11)
A more detailed room by room analysis is performed for the Reactor Auxiliary Building. In this case airborne radionuclide concentrations in the form of C/MPC were calculated as well as the whole body dose, in mrem per hour occupancy, resulting from inhalation and external exposure. The inhalation and external whole body dose conversion factors were taken from Table E-7 of Regulatory Guide 1.109, for the adult group Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the purpose of Evaluating Compliance with 10CFR Part 50, Appendix I (Revision 1) and Table D-3 of WASH 1258(3), respectively. The WASH-1258 values for a semi-infinite cloud model result in highly conservative dose values. The assumptions used in the analysis are essentially the same as those listed in Table 12.2-14. The differences are that for gaseous releases the iodine partition factor is assigned a valve of 1.0, and 0.05 for liquid leakage associated with the steam generator blow down heat exchanger, tank and pumps. Table 12.2-16 lists the results of the equipment and equipment leakage rates.
(DRN 99-2362, R11)
A negligible amount of radioactivity is expected to be released due to removal of reactor vessel head, movement of spent fuel or relief valve venting. Therefore, contribution from these sources to airborne activity are not considered.
WSES-FSAR-UNIT-3 12.2-7 Revision 307 (07/13)
The airborne concentration of a radioisotope in an area having a constant leak rate, source strength and exhaust rate, can be calculated by the equation given below. Radionuclide concentrations in other areas such as corridors are calculated assuming the air in the corridor can be contaminated by exhaust from nearby areas.
(DRN 02-110, R12)
FSAR Table 12.2-15a and 12.2-16a provide isotopic airborne concentrations as a function of derived air concentration (DAC) from 10 CFR 20, Appendix B.
(DRN 02-110, R12)
(DRN 99-2362, R11)
i i
t i
i i
MPC V
i e
PF SS a
W MPC t
1
/
i C
(DRN 99-2362, R11)
- where, W
=
leak rate of fluid in cm 3/hr ai
=
concentration of ith isotope in the primary coolant in ci/cm 3
=
source strength defined as fraction of primary coolant present in the leaking liquid PFi
=
partition factor of ith isotope i
=
total removal rate constant for ith isotope in hr.
-1.
=
di + e di
=
decay constant for ith isotope in hr.-1 e
=
removal rate constant due to exhaust in hr.
t
=
time interval in hours V
=
free volume of the area where leak occurs in cm 3
Ci(t)
=
airborne concentration of the ith radioisotope at time t in ci/cm 3
For small rooms and other operating areas where d for most of the radionuclides, the peak or equilibrium activity (Ceq) is given by the following equation:
Ceq / MPC =
W a SS PF V
d + e MPC
=
i i
i i
i i
i i
17 106
. x In order to calculate tritium concentration in the Fuel Handling Building the following equations were used to calculate evaporation rate from the fuel pool:
(DRN 99-2362, R11)
a w
0.425 95 A
wp
y (DRN 99-2362, R11)
WSES-FSAR-UNIT-3 12.2-8 Revision 307 (07/13) and ventilation rate:
CFM =
wp 4.5 Wi - Wo
- where, (DRN 99-2362, R11) wp
=
Evaporation rate of water in lbm/hr.
(DRN 99-2362, R11)
=
Air velocity over surface in ft/min.
y
=
Latent heat at pool surface water temperature in Btu/lb.
DRN 99-2362, R11) pa
=
Saturation pressure of vapor at room air dew point temperature in in. of mercury.
pw
=
Saturation pressure of vapor at the surface water temperature in in. of mercury.
(DRN 99-2362, R11)
A
=
Surface area of pool in ft2 Wo
=
Moisture content of out-door air in lbm/lbm of dry air Wi
=
Moisture content of in-door air in lbm/lbm of dry air CFM
=
Ventilation rate in ft3/min.
WSES-FSAR-UNIT-3 12.2-9 Revision 307 (07/13)
SECTION 12.2: REFERENCES 1
"DOT IIW User's Manual" WANL-TME-1982, December 1969.
2 DiNunnho, Anderson, Bakes and Anderson, "Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Atomic Energy Commission, March 23, 1962.
3 "Numerical Guides for Design Objectives and Limiting Conditions for Operation to meet the Criterion 'As Low As Practicable' for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents", WASH 1258, AEC, July 1973.
(DRN 99-2362, R11) 4 C-CE-4034, February 16, 1977, information utilized in development of Section 12.2.1 Tables.
DRN 99-2362, R11)
WSES-FSAR-UNIT-3 TABLE 12.2-1 NEUTRON FLUXES OUTSIDE THE REACTOR PRESSURE VESSEL Maximum Neutron spectra (n/cm2-sec)
E(mev)
Side Top Bottom 18 1.60(+4)*
1.09(-2) 14 9.33(+5) 8.08(-1) 10 2.42(+7) 1.35(+1) 8 5.36(+7) 1.00(+1) 6 9.98(+7) 1.06(+1) 4 9.50(+7) 1.16(+1) 3 6.68(+7) 1.33(+1) 2 7.00(+7) 1.17(+1) 1 7.98(+7) 1.53(+1) 0.33 9.60(+7) 3.42(+1)
Epithermal 4.83(+10)
Thermal 1.15(+10)
- Denotes power of ten (10
- - denotes insignificant
WSES-FSAR-UNIT-3 TABLE 12.2-2 NEUTRON LEAKAGE RATE FROM AXIAL REGIONS OF THE REACTOR VESSEL*
Axial Region Boundaries Height (cm)
Neutron Leakage Rate (n/sec.)
Upper (cm)
Lower (cm) 341.6 281.6 60.0 1.8 (+12)**
281.6 251.6 30.0 4.4 (+12) 251.6 236.6 15.0 8.9 (+12) 236.6 229.1 7.5 8.0 (+12) 229.1 221.6 7.5 1.1 (+13) 221.6 214.1 7.5 1.5 (+13) 214.1 206.6 7.5 1.8 (+13) 206.6 191.6 15.0 5.0 (+13) 191.6 161.6 30.0 1.51 (+14) 161.6 131.6 30.0 2.10 (+14) 131.6 101.6 30.0 2.55 (+14) 101.6 71.6 30.0 2.90 (+14) 71.6 11.6 60.0 around 6.12 (+14) core centerline 11.6
-48.4 60.0 5.54 (+14)
-48.4
-108.4 60.0 4.73 (+14)
-108.4
-168.4 60.0 2.13 (+14)
Bottom of the core is at elevation - 168.4 cm, and top of the core is at elevation 168.4 cm.
Denotes power of ten (10).
WSES-FSAR-UNIT-3 TABLE 12.2-3 GAMMA FLUXES AT FULL POWER OPERATION Maximum Gamma Spectra ( /cm 2sec.)
E(mev)
Side Top Bottom 10.00 3.05(+7)*
3.92(+2) 7.34(+3) 9.00 2.62(+8) 2.90(+3) 1.08(+7) 8.00 4.97(+8) 3.08(+3) 1.93(+7) 7.00 5.40(+8) 5.18(+3) 3.42(+7) 6.00 5.99(+8) 6.33(+3) 4.91(+7) 5.00 7.40(+8) 8.03(+3) 6.71(+7) 4.00 9.45(+8) 1.05(+4) 8.88(+7) 3.00 1.40(+9) 1.37(+4) 1.25(+8) 2.00 2.51(+9) 2.00(+4) 1.94(+8) 1.38 1.14(+9) 1.21(+4) 1.14(+8) 1.00 8.23(+8) 1.08(+4) 1.07(+8) 0.75 1.96(+9) 1.49(+4) 1.51(+8) 0.50 3.29(+9) 2.42(+4) 2.51(+8) 0.25 3.79(+9) 2.95(+4) 3.21(+8)
- Denotes power of ten (10)
WSES-FSAR-UNIT-3 TABLE 12.2-4 REACTOR COOLANT SYSTEM SOURCES (GAMMA SPECTRA)
AT SHUTDOWN Maximum Decay Gamma Spectra ( /cm2 -sec.)
E(mev) 1hr.
5 hr.
20 hr.
2d 10d 4.00 1.15(+7)*
1.56(+5) 1.08(+4) 8.28(+3) 5.54(+3) 3.00 2.21(+7) 1.98(+6) 1.59(+5) 1.38(+5) 9.63(+4) 2.00 4.40(+7) 8.53(+6) 3.42(+6) 3.14(+6) 1.96(+6) 1.38 3.04(+7) 6.28(+6) 2.72(+6) 2.45(+6) 1.53(+6) 1.00 2.81(+7) 5.98(+6) 2.62(+6) 2.34(+6) 1.47(+6) 0.75 3.90(+7) 8.46(+6) 3.75(+6) 3.34(+6) 2.09(+6) 0.50 6.26(+7) 1.35(+7) 5.99(+6) 5.34(+6) 3.33(+6) 0.25 7.79(+7) 1.66(+7) 7.32(+6) 6.51(+6) 4.04(+6)
- Denotes power of ten (10)
WSES-FSAR-UNIT-3 TABLE 12.2-5 REACTOR COOLANT SYSTEM SOURCES (MATERIAL ACTIVATION SPECTRA) AT SHUTDOWN Maximum Material Activation Spectra ( /cm2 -sec)
E(mev) 1 hr 5 hr 20 hr 2d 10d 2.00 1.33(+7)*
4.49(+6) 7.87(+4) 1.38 2.25(+6) 2.25(+6) 2.25(+6) 2.25(+6) 2.25(+6) 1.00 3.71(+7) 1.52(+7) 3.99(+6) 3.80(+6) 3.80(+6)
- Denotes power of ten (10)
WSES-FSAR-UNIT-3 TABLE 12.2-6 PRESSURIZER STEAM SECTION ACTIVITY Isotope Activity *(m Ci/cc)
Kr - 85 m 4.1 (-1) **
Kr - 85 1.28 (+3)
Kr - 87 6.6 (-2)
Kr - 88 4.5 (-1)
Xe-133 1.73 (+3)
Xe-135 4.09 (+0)
Xe-138 6.00 (-3)
- Assumes that all Xe and Kr isotopes build up for 1 yr. and are supplied to the pressurizer at a continuous primary water spray rate of 1.5 gpm, with complete stripping of the gas from water.
- Denotes power of ten (10)
WSES-FSAR-UNIT-3 TABLE 12.2-7 (Sheet 1 of 2)
MAXIMUM ACTIVITY INVENTORY IN CVCS COMPONENTS (c)
(CURIES)
Nuclide Regenerative Letdown Purification Purification Heat Exchanger Heat Exchanger Filter Ion-Exchanger Volume (gallons) 5.6/41.4(a) 45.9 2.2 cu.ft.
36 cu.ft.
N-16 8.8E-01(b) 1.2E+00 4.4E-02 1.4E-02 Br-84 1.1E-03 6.9E-03 0.
2.5E-01 Kr-85M 3.6E-01 4.1E-01 0.
0.
Kr-85 9.1E-01 8.9E-01 0.
0.
Kr-87 1.4E-01 2.2E-01 0.
0.
Kr-88 5.7E-01 7.1E-01 0.
0.
Rb-88 1.7E-01 7.1E-01 0.
9.3E+00 Sr-89 3.8E-04 1.9E-03 0.
1.5E+02 Sr-90 2.1E-05 1.0E-04 0.
3.4E+01 Y-90 5.2E-05 2.5E-04 0.
1.1E+00 Y-91 1.8E-03 8.4E-03 0.
8.0E+02 Y-91M 1.1E-04 6.7E-04 0.
3.9E-02 Sr-91 2.2E-04 1.1E-03 0.
7.2E-01 Mo-99 8.2E-02 3.9E-01 0.
1.8E+03 Ru-103 3.2E-04 1.5E-03 0.
9.8E+01 Ru-106 1.9E-05 8.9E-05 0.
2.3E+01 Ti-129 1.5E-03 8.7E-03 0.
6.9E-01 I-131 1.7E-01 8.1E-01 0.
1.1E+04 I-132 4.3E-02 2.3E-01 0.
3.6E+01 Te-132 2.4E-02 1.1E-01 0.
6.1E+02 I-133 2.1E-01 1.0E+00 0.
1.5E+03 I-134 1.7E-02 1.0E-01 0.
6.0E+00 Cs-134 1.5E-02 2.7E-02 0.
5.0E+03 I-135 9.0E-02 4.5E-01 0.
2.1E+02 Cs-136 2.6E-03 4.6E-03 0.
6.4E+01 Cs-137 6.2E-02 1.1E-01 0.
2.3E+04 Xe-131M 4.2E-01 4.2E-01 0.
0.
Xe-133 5.9E+01 5.8E+01 0.
0.
Xe-135M 6.3E-02 1.9E-01 0.
0.
Xe-135 1.5E+00 1.6E+00 0.
0.
Xe-138 3.1E-02 9.7E-02 0.
0.
Ba-140 4.7E-04 2.2E-03 0.
4.7E+01 La-140 4.5E-04 2.1E-03 0.
6.0E+00 Pr-143 4.2E-04 2.0E-03 0.
4.4E+01 Ce-144 2.9E-04 1.4E-03 0.
3.3E+02 Zr-95 4.2E-04 2.0E-03 1.8E+00 2.1E+02 Cr-51 1.2E-04 5.8E-04 2.6E+01 0.
Mn-54 2.0E-05 9.5E-05 2.3E+01 0.
Co-58 1.0E-03 4.9E-03 5.4E+02 0.
Fe-59 6.5E-05 3.1E-04 2.2E+01 0.
Co-60 1.3E-04 6.2E-04 1.9E+02 0.
WSES-FSAR-UNIT-3 TABLE 12.2-7 (Sheet 2 of 2)
Nuclide Deborating Volume Control Boric Acid Charging Ion-Exchanger Tank Makeup Tank Pump Volume (gallons) 36 cu.ft.
4780 11 800.
10.
N-16 0.0 0.0 0.
0.
Br-84 2.5E-02 1.9E-02 5.4E-07 5.7E-05 Kr-85M 0.
2.6E+02 0.
7.4E-02 Kr-85 0.
6.7E+02 0.
1.9E-01 Kr-87 0.
9.8E+01 0.
2.8E-02 Kr-88 0.
4.1E+02 0.
1.2E-01 Rb-88 0.
6.4E+00 6.6E-06 2.0E-02 Sr-89 0.
1.3E-02 1.1E-01 4.0E-05 Sr-90 0.
7.2E-04 2.2E-02 2.2E-06 Y-90 0.
1.7E-03 1.8E-04 5.3E-06 Y-91 0.
6.0E-02 5.6E-01 1.8E-04 Y-91M 0.
2.4E-03 1.3E-07 7.2E-06 Sr-91 0.
7.0E-03 1.9E-05 2.1E-05 Mo-99 0.
2.8E+00 3.1E-01 8.4E-03 Ru-103 0.
1.1E-02 6.7E-02 3.2E-05 Ru-106 0.
6.4E-04 1.6E-02 1.9E-06 Ti-129 6.8E-02 3.5E-02 3.0E-06 1.1E-04 I-131 5.1E+02 5.8E+00 4.4E+00 1.8E-02 I-132 3.6E+00 1.2E+00 2.8E-04 3.6E-03 Te-132 4.8E+01 8.0E-01 1.2E-01 2.4E-03 I-133 1.4E+02 7.0E+00 8.0E-02 2.1E-02 I-134 5.9E-01 3.6E-01 2.0E-05 1.1E-03 Cs-134 0.
9.6E-01 1.6E-00 2.9E-03 I-135 2.0E+01 2.9E+00 4.1E-03 8.6E-03 Cs-136 0.
1.6E-01 1.6E-02 5.0E-04 Cs-137 0.
3.9E+00 7.3E+00 1.2E-02 Xe-131M 0.
1.8E+02 0.
9.0E-02 Xe-133 0.
2.5E+04 0.
1.3E+01 Xe-135M 0.
1.9E+01 0.
9.7E-03 Xe-135 0.
6.4E+02 0.
3.2E-01 Xe-138 0.
9.1E+00 0.
4.5E-03 Ba-140 0.
1.6E-02 2.4E-02 4.8E-05 La-140 0.
1.5E-02 6.2E-04 4.6E-05 Pr-143 0.
1.4E-02 2.3E-02 4.3E-05 Ce-144 0.
9.8E-03 2.2E-01 3.0E-05 Zr-95 0.
1.4E-02 1.5E-01 4.3E-05 Cr-51 0.
4.2E-03 1.7E-02 1.3E-05 Mn-54 0.
6.8E-04 1.6E-02 2.1E-06 Co-58 0.
3.5E-02 3.9E-01 1.1E-04 Fe-59 0.
2.2E-03 1.6E-02 6.7E-06 Co-60 0.
4.4E-03 1.3E-01 1.3E-05 (a) Tube volume/shell volume (b) E-01 denotes 10-1 (c) Inventories < 1 x 10-15 Curies are considered to be insignificant.
WSES-FSAR-UNIT-3 TABLE 12.2-8 (Sheet 1 of 2)
Revision 11-B (06/02)
MAXIMUM ACTIVITY INVENTORY IN BMS COMPONENTS (CURIES)
Nuclide Reactor Drain Equipment Flash Boric Acid Tank Drain Tank Tank (1)
Condensate Tank
Volume (gallons) 1600.
4000.
400.
17200.
Br-84 1.1E-01(a) 3.8E-02 6.0E-03 2.6E-09 Kr-85M 6.9E+00 2.3E+00 3.6E+00 5.4E-04 Kr-85 1.0E+03 3.0E+01 7.8E+00 1.4E+00 Kr-87 3.5E+00 1.2E+00 1.9E+00 2.9E-05 Kr-88 1.2E+01 3.9E+00 6.2E+00 4.0E-04 Rb-88 1.1E+01 3.9E+00 3.1E+00 3.2E-08 Sr-89 5.8E-02 1.4E-02 1.6E-03 5.0E-05 Sr-90 3.2E-03 8.2E-04 8.8E-05 3.2E-06 Y-90 5.9E-03 1.4E-03 2.2E-04 7.4E-07 Y-91 2.6E-01 6.4E-02 7.3E-03 2.3E-04 Y-91M 1.1E-02 3.7E-03 5.8E-04 6.2E-10 Sr-91 1.9E-02 5.9E-03 9.3E-04 9.5E-08 Mo-99 9.4E+00 2.3E+00 3.4E-01 1.2E-03 Ru-103 4.6E-02 1.1E-02 1.3E-03 3.9E-05 Ru-106 2.8E-03 7.2E-04 7.8E-05 2.8E-06 Ti-129 1.4E-01 4.8E-02 7.5E-03 1.5E-08 I-131 2.3E+01 5.2E+00 7.1E-01 9.8E-03 I-132 3.7E+00 1.3E+00 2.0E-01 1.4E-06 Te-132 2.8E+00 6.7E-01 9.8E-02 4.6E-04 I-133 2.0E+01 5.7E+00 8.9E-01 3.9E-04 I-134 1.6E+00 5.5E-01 8.7E-02 9.9E-08 Cs-134 8.5E-01 2.2E-01 1.2E-01 2.6E-04 I-135 7.7E+00 2.5E+00 3.9E-01 2.0E-05 Cs-136 1.3E-01 3.1E-02 2.0E-02 2.5E-05 Cs-137 3.5E+00 8.8E-01 4.7E-01 1.1E-03 Xe-131M 3.5E+01 3.0E+00 3.6E+00 3.5E-01 Xe-133 2.7E+03 3.7E+02 5.1E+02 2.5E+01 Xe-135M 3.0E+00 1.0E+00 1.7E+00 1.2E-06 Xe-135 2.9E+01 8.9E+00 1.4E+01 8.1E-03 Xe-138 1.5E+00 5.3E-01 8.4E-01 5.0E-07 Ba-140 6.5E-02 1.5E-02 1.9E-03 3.7E-05 La-140 4.7E-02 1.2E-02 1.9E-03 2.9E-06 Pr-143 5.8E-02 1.3E-02 1.7E-03 3.4E-05 Ce-144 4.3E-02 1.1E-02 1.2E-03 4.2E-05 Zr-95 6.3E-02 1.5E-02 1.7E-03 5.6E-05 Cr-51 1.8E-02 4.2E-03 5.1E-04 1.4E-05 Mn-54 3.0E-03 7.7E-04 8.3E-05 3.0E-06 Co-58 1.5E-01 3.8E-02 4.3E-03 1.4E-04 Fe-59 9.5E-03 2.3E-03 2.7E-04 8.1E-06 Co-60 2.0E-02 5.0E-03 5.4E-04 2.0E-05
(a) E-01 denotes 10-1 (1) This is historical data. The Flash Tank path is inactive per ER-W3-00-0225-00-00. Effluent now flows directly to the holdup tanks.
WSES-FSAR-UNIT-3 TABLE 12.2-8 (Sheet 2 of 2)
Boric Acid Nuclide Pre-Concentrator Pre-Concentrator Condensate Holdup Filter Ion-Exchanger Ion-Exchanger Tanks Volume 2.2 cu. ft.
36 cu. ft.
36 cu. ft.
47960 gallons Br-84 0.
4.8E-06(a) 2.4E-09 2.9E-03 Kr-84M 0.
0.
0.
5.1E+02 Kr-85 0.
0.
0.
1.9E+03 Kr-87 0.
0.
0.
1.3E+02 Kr-88 0.
0.
0.
7.1E+02 Rb-88 0.
6.5E-04 0.
6.4E-01 Sr-89 0.
9.9E-01 0.
2.7E-01 Sr-90 0.
2.4E-01 0.
1.6E-02 Y-90 0.
1.6E-03 0.
8.1E-03 Y-91 0.
5.2E+00 0.
1.2E+00 Y-91M 0.
1.1E-06 0.
4.3E-04 Sr-91 0.
1.7E-04 0.
5.7E-03 Mo-99 0.
2.8E+00 0.
1.3E+01 Ru-103 0.
6.1E-01 0.
2.1E-01 Ru-106 0.
1.7E-01 0.
1.4E-02 Ti-129 0.
2.7E-05 1.3E-08 7.4E-03 I-131 0.
4.0E+01 1.9E-02 6.5E+01 I-132 0.
2.6E-03 1.3E-06 3.6E-01 Te-132 0.
1.1E+00 5.3E-04 4.5E+00 I-133 0.
7.2E-01 3.5E-04 1.1E+01 I-134 0.
1.8E-04 8.9E-08 6.6E-02 Cs-134 0.
1.9E+02 0.
1.3E+01 I-135 0.
3.7E-02 1.8E-05 1.7E+00 Cs-136 0.
1.6E+00 0.
1.5E+00 Cs-137 0.
8.9E+02 0.
5.4E+01 Xe-131M 0.
0.
0.
5.0E+02 Xe-133 0.
0.
0.
6.4E+04 Xe-135M 0.
0.
0.
1.4E+01 Xe-135 0.
0.
0.
1.1E+03 Xe-138 0.
0.
0.
6.3E+00 Ba-140 0.
2.2E-01 0.
2.2E-01 La-140 0.
5.6E-03 0.
4.4E-02 Pr-143 0.
2.1E-01 0.
2.0E-01 Ce-144 0.
2.3E+00 0
2.1E-01 Zr-95 1.2E-02 1.4E+00 0.
2.9E-01 Cr-51 1.6E-01 0.
0.
7.5E-02 Mn-54 1.7E-01 0.
0.
1.5E-02 Co-58 3.7E+00 0.
0.
7.2E-01 Fe-59 1.4E-01 0.
0.
4.3E-02 Co-60 1.4E+00 0.
0.
9.9E-02 (a) E-06 denotes 10-6
WSES-FSAR-UNIT-3 TABLE 12.2-9 MAXIMUM ACTIVITY INVENTORY IN FPS COMPONENTS Nuclide Fuel Pool Pool Purification Fuel Pool Ion-Exchanger Filter Heat Exchanger Volume 36 cu. ft.
2.2 cu. ft.
383 gallons Br-84 0.
0.
0.
Kr-85M 0.
0.
5.8E-06 Kr-85 0.
0.
1.7E-02 Kr-87 0.
0.
4.5E-14 Kr-88 0.
0.
1.5E-07 Rb-88 0.
0.
0.
Sr-89 1.2E-01(a) 0.
1.7E-04 Sr-90 7.4E-03 0.
9.7E-06 Y-90 4.8E-03 0.
1.4E-05 Y-91 5.5E-01 0.
7.9E-04 Y-91M 0.
0.
0.
Sr-91 3.4E-04 0.
3.7E-06 Mo-99 7.9E+00 0.
2.3E-02 Ru-103 9.4E-02 0.
1.4E-04 Ru-106 6.4E-03 0.
8.6E-06 Te-129 7.3E-15 0.
0.
I-131 3.4E+01 0.
6.6E-02 I-132 4.1E-07 0.
1.7E-08 Te-132 2.6E+00 0.
7.2E-03 I-133 3.5E+00 0.
2.1E-02 I-134 0.
0.
0.
Cs-134 3.1E+00 0.
4.1E-03 I-135 2.4E-02 0.
3.5E-04 Cs-136 3.6E-01 0.
6.4E-04 Cs-137 1.3E+01 0.
1.7E-02 Xe-131M 0.
0.
7.1E-03 Xe-133 0.
0.
8.7E-01 Xe-135M 0.
0.
0.
Xe-135 0.
0.
9.7E-04 Xe-138 0.
0.
0.
Ba-140 1.1E-01 0.
1.9E-04 La-140 2.4E-02 0.
9.3E-05 Pr-143 9.9E-02 0.
1.7E-04 Ce-144 9.7E-02 0.
1.3E-04 Zr-95 1.3E-01 1.2E-03 1.9E-04 Cr-51 0.
3.5E-02 5.4E-05 Mn-54 0.
6.8E-03 9.2E-06 Co-58 0.
3.3E-01 4.6E-04 Fe-59 0.
2.0E-02 2.9E-05 Co-60 0.
4.5E-02 5.9E-05 (a) E-01 denotes 10-1
WSES-FSAR-UNIT-3 TABLE 12.2-10 MAXIMUM AND EXPECTED ACTIVITY INVENTORY IN SIS COMPONENTS (CURIES)
Shutdown Heat Exchanger MAXIMUM Shutdown LOCA Volume (gallons) 875 875 Br-84 1.3E-01(a) 2.9E+04 Kr-85M 7.8E+00 4.1E+02 Kr-85 1.7E+01 8.9E+00 Kr-87 4.2E+00 7.6E+02 Kr-88 1.4E+01 1.1E+03 Rb-88 1.3E+01 2.2E+03 Sr-89 3.5E-02 3.0E+03 Sr-90 1.9E-03 1.9E+02 Y-90 4.7E-03 1.9E+02 Y-91 1.6E-01 3.7E+03 Y-91M 1.3E-02 2.1E+03 Sr-91 2.0E-02 3.6E+03 Mo-99 7.4E+00 3.9E+03 Ru-103 2.8E-02 1.9E+03 Ru-106 1.7E-03 1.7E+02 Te-129 1.6E-01 6.7E+02 I-131 1.5E+01 9.1E+04 I-132 4.4E+00 1.4E+05 Te-132 2.1E+00 2.7E+03 I-133 1.9E+01 2.1E+05 I-134 1.9E+00 2.4E+05 Cs-134 5.1E-01 2.1E+01 I-135 8.5E+00 1.9E+05 Cs-136 8.7E-02 3.6E+00 Cs-137 2.1E+00 8.5E+01 Xe-131M 7.9E+00 6.3E+00 Xe-133 1.1E+03 2.0E+03 Xe-135M 3.6E+04 5.6E+02 Xe-135 3.1E+01 4.7E+02 Xe-138 1.8E+00 1.8E+03 Ba-140 4.2E-02 3.9E+03 La-140 4.1E-02 4.0E+03 Pr-143 3.7E-02 3.7E+03 Ce-144 2.6E-02 2.6E+03 Zr-95 3.8E-02 3.8E+03 Cr-51 1.1E-02 1.4E-05 Mn-54 1.8E-03 2.3E-06 Cc-58 9.3E-02 1.2E-04 Fe-59 5.8E-03 7.5E-06 Co-60 1.2E-02 1.5E-05 (a) E-01 denotes 10-1
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 1 of 6)
Revision 307 (07/13)
(LBDCR 13-009, R307)
MAXIMUM ACTIVITY INVENTORY IN WMS COMPONENTS (Curies)
(DRN 01-1249, R11-B)
Waste Nuclide Waste Laundry Waste Condensate Demineralizer Waste Condensate Tank Tank Tank System Ion Exchanger__
(DRN 01-1249, R11-B)
Volume(gallons) 4000.
4000.
17200.
1000.
36 cu. ft.
Br-84 4.8E-02(a) 6.0E-05 4.1E-05 2.4E-01 9.9E-07 Kr-85M
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
Kr-87
- 0.
- 0.
- 0.
- 0.
- 0.
Kr-88
- 0.
- 0.
- 0.
- 0.
- 0.
Rb-88 4.9E+00 6.1E-03 4.2E-03 2.5E+01 5.6E-05 Sr-89 1.3E-02 1.6E-05 1.1E-05 6.5E-02 6.0E-04 Sr-90 7.0E-04 8.8E-07 6.0E-07 3.5E-03 1.3E-04 Y-90 1.7E-03 2.2E-06 1.5E-06 8.6E-03 4.3E-06 Y-91 5.9E-02 7.3E-05 5.0E-05 2.9E-01 3.1E-03 Y-91M 4.7E-03 5.8E-06 4.0E-06 2.3E-02 1.5E-07 Sr-91 7.4E-03 9.3E-06 6.4E-06 3.7E-02 2.8E-06 Mo-99 2.7E+00 3.4E-03 2.3E-03 1.4E+01 7.0E-03 Ru-103 1.0E-02 1.3E-05 8.9E-06 5.2E-02 3.8E-04 Ru-106 6.2E-04 7.8E-07 5.3E-07 3.1E-03 9.0E-05 Ti-129 6.0E-02 7.5E-05 5.2E-05 3.0E-01 2.7E-06 I-131 5.7E+00 7.1E-03 4.9E-03 2.8E+01 4.3E-02 I-132 1.6E+00 2.0E-03 1.4E-03 8.0E+00 1.4E-04 Te-132 7.8E-01 9.8E-04 6.8E-04 3.9E+00 2.4E-03 I-133 7.1E+00 8.9E-03 6.1E-03 3.6E+01 5.8E-03 I-134 6.9E-01 8.7E-04 6.0E-04 3.5E+00 2.4E-05 Cs-134 1.9E-01 2.3E-04 1.6E-04 9.3E-01 3.1E-02 I-135 3.1E+00 3.9E-03 2.7E-03 1.6E+01 8.1E-04 Cs-136 3.2E-02 4.0E-05 2.8E-05 1.6E-01 3.9E-04 Cs-137 7.5E-01 9.4E-04 6.5E-04 3.8E+00 1.4E-01 Xe-131M
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
Xe-135M
- 0.
- 0.
- 0.
- 0.
- 0.
Xe-135
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
- 0.
Ba-140 1.5E-02 1.9E-05 1.3E-05 7.7E-02 1.8E-04 La-140 1.5E-02 1.9E-05 1.3E-05 7.4E-02 2.3E-05 Pr-143 1.4E-02 1.7E-05 1.2E-05 6.8E-02 1.7E-04 Ce-144 9.5E-03 1.2E-05 8.1E-06 4.7E-02 1.3E-03 Zr-95 1.4E-02 1.7E-05 1.2E-05 7.0E-02 8.1E-04 Cr-51 4.1E-03 5.1E-06 3.5E-05 2.0E-02 Mn-54 6.6E-04 8.3E-07 5.7E-06 3.3E-03
- 0.
Co-58 3.4E-02 4.3E-05 2.9E-04 1.7E-01
- 0.
Fe-59 2.1E-03 2.7E-06 1.8E-05 1.1E-02
- 0.
Co-60 4.3E-03 5.4E-06 3.7E-05 2.1E-02
- 0.
(a) E-02 denotes 10-2
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 2 of 6)
Revision 307 (07/13)
(LBDCR 13-009, R307)
Nuclide Waste Waste Oil Laundry Spent Resin Gas Gas Filter Filter Filter Tank Decay Tank Surge Tank Volume (gallons) 1.1 cu. ft..
6 cu. ft.
1.1 cu. ft.
3200.
44.88 150 Br-84
- 0.
- 0.
- 0.
2.8E-01 1.0E-03 2.8E-06 Kr-85M
- 0.
- 0.
- 0.
- 0.
5.9E+02 1.6E+00 Kr-85
- 0.
- 0.
- 0.
- 0.
4.0E+03 1.1E+01 Kr-87
- 0.
- 0.
- 0.
- 0.
3.2E+02 8.7E-01 Kr-88
- 0.
- 0.
- 0.
- 0.
1.0E+03 2.8E+00 Rb-88
- 0.
- 0.
- 0.
9.3E+00 5.1E-01 1.4E-03 Sr-89
- 0.
- 0.
- 0.
1.5E+02 1.1E-03 2.9E-06 Sr-90
- 0.
- 0.
- 0.
3.4E+01 7.7E-05 2.1E-07 Y-90
- 0.
- 0.
- 0.
1.1E+00 4.6E-05 1.3E-07 Y-91
- 0.
- 0.
- 0.
8.0E+02 5.0E-03 1.4E-05 Y-91M
- 0.
- 0.
- 0.
3.9E-02 9.7E-05 2.7E-07 Sr-91
- 0.
- 0.
- 0.
7.2E-01 1.5E-04 4.3E-07 Mo-99
- 0.
- 0.
- 0.
1.8E+03 7.2E-02 2.0E-04 Ru-103
- 0.
- 0.
- 0.
9.8E+01 7.9E-04 2.2E-06 Ru-106
- 0.
- 0.
- 0.
2.3E+01 6.5E-05 1.8E-07 Ti-129
- 0.
- 0.
- 0.
7.6E-01 1.3E-03 3.5E-06 I-131
- 0.
- 0.
- 0.
1.1E+04 2.1E-01 5.9E-04 I-132
- 0.
- 0.
- 0.
4.0E+01 3.3E-02 9.1E-05 Te-132
- 0.
- 0.
- 0.
6.6E+02 2.2E-02 6.0E-05 I-133
- 0.
- 0.
- 0.
1.6E+03 1.6E-01 4.4E-04 I-134
- 0.
- 0.
- 0.
6.6E+00 1.4E-02 4.0E-05 Cs-134
- 0.
- 0.
- 0.
5.0E+03 9.9E-02 2.7E-04 I-135
- 0.
- 0.
- 0.
2.3E+02 6.5E-02 1.8E-04 Cs-136
- 0.
- 0.
- 0.
6.4E+01 7.6E-03 2.1E-05 Cs-137
- 0.
- 0.
- 0.
2.3E+04 4.1E-01 1.1E-03 Xe-131M
- 0.
- 0.
- 0.
- 0.
9.6E+02 2.6E+00 Xe-133
- 0.
- 0.
- 0.
- 0.
1.1E+05 2.9E+02 Xe-135M
- 0.
- 0.
- 0.
- 0.
2.8E+02 7.6E-01 Xe-135
- 0.
- 0.
- 0.
- 0.
2.3E+03 6.4E+00 Xe-138
- 0.
- 0.
- 0.
- 0.
1.4E+02 3.9E-01 Ba-140
- 0.
- 0.
- 0.
4.7E+01 7.2E-04 2.0E-06 La-140
- 0.
- 0.
- 0.
6.0E+00 3.6E-04 1.0E-06 Pr-143
- 0.
- 0.
- 0.
4.4E+01 6.6E-04 1.8E-06 Ce-144
- 0.
- 0.
- 0.
3.3E+02 9.8E-04 2.7E-06 Zr-95 3.8E-03(a)
- 0.
- 0.
2.1E+02 1.2E-03 3.3E-06 Cr-51 5.6E-02 5.6E-02 2.3E-05
- 0.
2.7E-04 7.5E-07 Mn-54 4.9E-02 4.9E-02 2.0E-05
- 0.
6.9E-05 1.9E-07 Co-58 1.1E+00 1.1E-00 4.6E-04
- 0.
3.0E-03 8.3E-06 Fe-59 4.7E-02 4.7E-02 1.9E-05
- 0.
1.7E-04 4.6E-07 Co-60 4.0E-01 4.0E-01 1.7E-04
- 0.
4.7E-04 1.3E-06 (a) E-03 denotes 10-3
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 3 of 6)
Revision 307 (07/13)
Maximum Anticipated Activity per Container Stored in the Low Level Radioactive Waste Storage Facility (Curies)
Dry Active Waste Container Maximum Anticipated Activity per Container Nuclide 1280 Cubic Feet Container Curies per Container Cs-137 1.28E+03 2.15E+00 Co-58 1.28E+03 3.97E+00 Cs-134 1.28E+03 1.17E+00 Fe-55 1.28E+03 6.01E-01 Co-60 1.28E+03 1.09E+00 Mn-54 1.28E+03 3.86E-01 Ni-63 1.28E+03 2.77E+00 Ce-144 1.28E+03 7.21E-03 H-3 1.28E+03 2.17E-01 C-14 1.28E+03 1.13E-02 Co-57 1.28E+03 3.27E-04 Sb-125 1.28E+03 5.58E-05 Cr-51 1.28E+03 4.68E-03 Nb-95 1.28E+03 4.71E-04 Zr-95 1.28E+03 2.42E-04 Ce-141 1.28E+03 6.27E-05 Sn-113 1.28E+03 1.09E-04 1.24E+01 (LBDCR 13-009, R307)
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 4 of 6)
Revision 307 (07/13)
Maximum Anticipated Activity per Container Stored in the Low Level Radioactive Waste Storage Facility (Curies)
Plant Bead Resin Container Maximum Anticipated Activity per Container Nuclide Container 145 Cubic Feet Curies per Container Cs-137 1.45E+02 1.20E+02 Co-58 1.45E+02 1.08E+02 Cs-134 1.45E+02 6.06E+01 Fe-55 1.45E+02 2.77E+01 Co-60 1.45E+02 9.50E+00 Mn-54 1.45E+02 8.41E+00 Ni-63 1.45E+02 2.81E+01 Ce-144 1.45E+02 8.00E-02 H-3 1.45E+02 2.33E-01 C-14 1.45E+02 1.26E-01 Ru-106 1.45E+02 0.00E+00 Sb-125 1.45E+02 2.25E+00 Nb-95 1.45E+02 0.00E+00 Ce-144 1.45E+02 8.00E-02 Pu-241 1.45E+02 1.30E-02 3.66E+02 Liquid Waste Management Bead Resin Container Maximum Anticipated Activity per Container Nuclide Container 205 Cubic Feet Curies per Container Cs-137 2.05E+02 1.72E+00 Co-58 2.05E+02 3.64E-01 Cs-134 2.05E+02 7.26E-01 Fe-55 2.05E+02 6.91E-01 Co-60 2.05E+02 5.83E-01 Mn-54 2.05E+02 3.47E-02 Ni-63 2.05E+02 1.87E+00 Ce-144 2.05E+02 2.17E-02 H-3 2.05E+02 6.20E-02 C-14 2.05E+02 1.48E-02 Ru-106 2.05E+02 4.46E-04 Sb-125 2.05E+02 8.76E-02 Nb-95 2.05E+02 4.06E-02 Ce-144 2.05E+02 2.17E-02 Pu-241 2.05E+02 1.16E-02 6.25E+00 (LBDCR 13-009, R307)
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 5 of 6)
Revision 307 (07/13)
Maximum Anticipated Activity per Container Stored in the Low Level Radioactive Waste Storage Facility (Curies)
Plant Filters, Class A & B Container Maximum Anticipated Activity per Container Nuclide Container 120 Cubic Feet Curies per Container Cs-137 1.20E+02 2.51E+01 Co-58 1.20E+02 6.65E+01 Cs-134 1.20E+02 1.63E+01 Fe-55 1.20E+02 6.99E+01 Co-60 1.20E+02 1.10E+01 Mn-54 1.20E+02 7.63E+00 Ni-63 1.20E+02 1.15E+01 Ce-144 1.20E+02 3.60E-01 H-3 1.20E+02 4.04E-07 C-14 1.20E+02 7.74E-01 Ru-106 1.20E+02 7.74E-01 Sb-125 1.20E+02 5.83E+00 Nb-95 1.20E+02 1.79E+01 Ce-144 1.20E+02 3.60E-01 Cr-51 1.20E+02 2.40E+01 Zr-95 1.20E+02 2.95E+00 Sn-113 1.20E+02 1.42E+00 2.62E+02 Secondary Blowdown Bead Resin Container Maximum Anticipated Activity per Container Nuclide Container 205 Cubic Feet Curies per Container Cs-137 2.05E+02 1.89E-05 Co-58 2.05E+02 3.93E-06 Cs-134 2.05E+02 9.48E-06 Fe-55 2.05E+02 0.00E+00 Co-60 2.05E+02 4.80E-06 Mn-54 2.05E+02 1.78E-06 Ni-63 2.05E+02 6.44E-07 Ce-144 2.05E+02 0.00E+00 H-3 2.05E+02 0.00E+00 3.95E-05 (LBDCR 13-009, R307)
WSES-FSAR-UNIT-3 (LBDCR 13-009, R307)
TABLE 12.2-11 (Sheet 6 of 6)
Revision 307 (07/13)
Maximum Anticipated Activity per Container Stored in the Low Level Radioactive Waste Storage Facility (Curies)
Condensate Polisher Powder Resin Container Maximum Anticipated Activity per Container Nuclide Container 205 Cubic Feet Curies per Container Cs-137 2.05E+02 7.12E-06 Co-58 2.05E+02 0.00E+00 Cs-134 2.05E+02 3.11E-06 Fe-55 2.05E+02 0.00E+00 Co-60 2.05E+02 0.00E+00 Mn-54 2.05E+02 0.00E+00 Ni-63 2.05E+02 0.00E+00 Ce-144 2.05E+02 5.00E-07 H-3 2.05E+02 3.05E-07 1.10E-05 (LBDCR 13-009, R307)
WSES-FSAR-UNIT-3 TABLE 12.2-12 Revision 14 (12/05)
LOCA CORE INVENTORY (Curies/MWt) (a)
Nuclide (Ci)/MWt Nuclide (Ci)/MWt Nuclide (Ci)/MWt
(DRN 03-2066, R14)
Co-58 2.553E+02 Ru-103 4.509E+04 Cs-136 2.110E+03 Co-60 1.953E+02 Ru-105 2.340E+04 Cs-137 4.273E+03 Kr-85 3.562E+02 Ru-106 1.719E+04 Ba-139 4.976E+04 Kr-85m 1.052E+04 Rh-105 1.621E+04 Ba-140 5.088E+04 Kr-87 2.125E+04 Sb-127 2.208E+03 La-140 5.285E+04 Kr-88 3.003E+04 Sb-129 9.305E+03 La-141 4.615E+04 Rb-86 1.496E+01 Te-127 2.132E+03 La-142 4.449E+04 Sr-89 2.941E+04 Te-127m 2.823E+02 Ce-141 4.476E+04 Sr-90 2.867E+03 Te-129 9.162E+03 Ce-143 4.585E+04 Sr-91 3.927E+04 Te-129m 1.360E+03 Ce-144 3.421E+04 Sr-92 3.805E+04 Te-131m 4.161E+03 Pr-143 4.450E+04 Y-90 3.022E+03 Te-132 4.059E+04 Nd-147 1.911E+04 Y-91 3.683E+04 I-131 2.853E+04 Np-239 5.120E+05 Y-92 3.819E+04 I-132 4.127E+04 Pu-238 2.902E+01 Y-93 4.320E+04 I-133 5.769E+04 Pu-239 6.545E+00 Zr-95 4.785E+04 I-134 6.418E+04 Pu-240 8.254E+00 Zr-97 4.562E+04 I-135 5.412E+04 Pu-241 1.390E+03 Nb-95 4.782E+04 Xe-133 5.642E+04 Am-241 9.181E-01 Mo-99 5.173E+04 Xe-135 1.659E+04 Cm-242 3.514E+02 Tc-99M 4.532E+04 Cs-134 8.023E+03 Cm-244 2.056E+01 (a)
Core inventories are based on 100.5% of full power core conditions.
(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 TABLE 12.2-12A Revision 14 (12/05)
CORE INVENTORY FOR STEAMING EVENTS (Curies)
(DRN 03-2066, R14)
Isotope Curies Kr-83m 1.68E+07 Kr-85 1.17E+06 Kr-85m 3.93E+07 Kr-87 7.94E+07 Kr-88 1.12E+08 I-131 1.04E+08 I-132 1.50E+08 I-133 2.13E+08 I-134 2.40E+08 I-135 1.98E+08 Xe-133 2.05E+08 Xe-135 6.20E+07 Xe-131m 1.17E+06 Xe-133m 6.56E+06 Xe-135m 4.17E+07 Xe-138 1.95E+08
(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 TABLE 12.2-13 FISSION PRODUCT GAMMA SOURCE IN CONTAINMENT BUILDING (Mev/sec)
(assuming 100% noble gases, 50% halogens, 1% solids)
Energy Interval (Mev)
Time
.1 -.4
.4 -.9
.9 - 1.35 1.35 - 1.8 1.8 - 2.2 2.2 - 2.6 2.6 0
3.08(18)*
1.84(19) 7.30(18) 1.37(19) 8.90(18) 6.41(18) 2.90(18)
.5 hr.
2.93(18) 1.62(19) 6.65(18) 5.02(18) 4.63(18) 5.19(18) 3.71(17) 1 hr.
2.82(18) 1.43(19) 5.81(18) 4.45(18) 3.12(18) 4.30(18) 1.36(17) 2 hr.
2.68(18) 1.14(19) 4.70(18) 3.56(18) 2.14(18) 3.05(18) 4.38(16) 8 hr.
2.09(18) 6.41(18) 2.16(18) 1.58(18) 6.91(17) 5.76(17) 1.56(16) 24 hr.
1.16(18) 4.45(18) 8.68(17) 6.23(17) 2.32(17) 1.03(17) 2.57(14) 1 wk.
3.06(17) 1.07(18) 1.63(17) 1.75(17) 5.88(16) 3.11(16) 1.74(14) 1 mo.
4.10(16) 1.28(17) 3.17(15) 2.42(16) 1.46(15) 1.85(15) 5.52(13) 2 mo.
5.98(15) 7.86(16) 5.37(14) 4.30(15) 7.72(14) 2.85(14) 1.78(13) 4 mo.
7.51(14) 4.35(16) 1.98(14) 4.87(14) 6.45(14) 3.41(13) 7.72(12)
- Denotes power of ten (10)
WSES-FSAR-UNIT-3 TABLE 12.2-14 (Sheet 1 of 2)
Revision 12 (10/02)
ASSUMPTIONS AND PARAMETERS USED TO CALCULATE AIRBORNE CONCENTRATIONS
¨(DRN 99-2362)
Leakage Rates and Source Stream (SS):
Leakage into containment 1% of the noble gas inventory/d
.001% of the iodine inventory/d SS = 1
Steam leakage into turbine building 1700 lb/h*
Leakage into Reactor Auxiliary Building 160 lb./d, SS = 1 Leakage from the gas surge tank 0.005 scfm, **
Leakage from the waste gas com-0.02 scfm, **
pressor room Leakage from letdown heat 6.6 gal./d, SS = 1 exchanger room (DRN 00-1046)
Corridor elevation - 4 ft. MSL in RAB 10% of the exhaust from nearest rooms flows into the corridor Partition Factors:
Turbine Building 1 for iodine, 1 for noble gases Reactor Auxiliary Building 0.0075 for iodines, 1 for noble gases Letdown heat exchanger 0.1 for iodines, 1 for noble gases.
Ventilation Rates (cfm):
Containment Isolated case Turbine Building 636,000 Fuel Handling Building 26,000 Reactor Auxiliary Building 77,000 Gas surge tank room 365 Waste gas compressor room 600 Letdown heat exchanger room 450 Volumes (Cu. Ft.):
Containment 2.7 x 106 Turbine Building 2.5 x 106 Fuel Handling Building 4.08 x 105 Reactor Auxiliary Building 2.0 x 106 Other Factors:
Failed fuel fraction 0.12%
Plant load 80%
Outside air condition (winter) 56.1 F, 76.7% Relative Humidity
WSES-FSAR-UNIT-3 TABLE 12.2-14 (Sheet 2 of 2)
Other Factors: (Contd)
Fuel pool parameters:
Surface temperature 150°F Surface Area 1152 ft2 Air velocity over surface 10 ft./min.
Concentration of the liquid leaking into the turbine was assumed to be the same as that for secondary coolant.
Gaseous source terms were calculated assuming degassification of 0.540 gpm of reactor coolant and an over all decontamination factor of 2000 for iodine and waste gas (flow rate 30,000 scf/yr.)
WSES-FSAR-UNIT-3 TABLE 12.2-15 Revision 12 (10/02)
AVERAGE AIRBORNE C/MPC IN REACTOR AUXILIARY BUILDING, TURBINE BUILDING, CONTAINMENT AND FUEL HANDLING BUILDING ISOTOPE CONTAINMENT (C/MPC)
FUEL HANDLING BUILDING (C/MPC)
TURBINE BUILDING (C/MPC)
REACTOR AUXILIARY BUILDING (C/MPC)
Kr-83M 1.02(-1)*
7.43(-9) 7.16(-4)
Kr-85m 1.87(-1) 6.61(-8) 5.97(-4)
Kr-85 2.70(0) 8.21(-9) 1.39(-5)
Kr-87 2.06(-1) 2.05(-8) 1.98(-3)
Kr-88 1.39(0) 7.04(-8) 6.81(-3)
Kr-89 7.77(-4) 1.79(-9) 3.15(-5)
Xe-131m 3.23(-1) 2.19(-9) 1.78(-5)
Xe-133m 8.56(-1) 8.12(-9) 2.35(-4)
Xe-133 8.71(+1) 6.73(-7) 1.03(-2)
Xe-135m 9.81(-3) 4.60(-9) 2.56(-4)
Xe-135 1.5(0) 3.10(-8) 2.40(-3)
Xe-137 1.71(-3) 3.19(-9) 6.72(-5)
Xe-138 2.96(-2) 1.54(-8) 8.11(-2)
I-131 2.51(1) 8.77(-3) 1.49(-2)
I-133 1.00(0) 2.26(-3) 3.93(-4)
¨(DRN 02-110)
H-3 1.58(+1) 3.60(0) 2.04(-4) 2.92(-4)
(DRN 02-110) 1.02(-1) = 1.02 x 10-1 Indicates negligible C/MPC.
WSES-FSAR-UNIT-3
(DRN 02-110, R12)
Table 12.2-15a (Sheet 1 of 2)
Revision 14 (12/05)
Average Airborne C/DAC in Reactor Auxiliary Building, Turbine Building, Containment and Fuel Handling Building
(DRN 05-455, R14)
Isotope Containment Building Fuel Handling Building Turbine Building Reactor Auxiliary Building C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC Kr-83m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Kr-85m 1.45E-07 7.24E-03 6.13E-10 3.06E-05 Kr-85 9.26E-03 9.26E+01 3.21E-08 3.21E-04 Kr-87 4.30E-08 8.59E-03 5.60E-10 1.12E-04 Kr-88 1.01E-07 5.05E-02 6.62E-10 3.31E-04 Kr-89 0.00E+00 0.00E+00 Xe-131m 4.82E-04 1.21E+00 3.40E-08 8.49E-05 Xe-133m 7.75E-06 7.75E-02 2.92E-09 2.92E-05 Xe-133 7.97E-06 7.97E-02 1.26E-09 1.26E-05 Xe135m 6.75E-08 7.50E-03 2.45E-09 2.72E-04 Xe-135 1.24E-06 1.24E-01 2.66E-09 2.66E-04 Xe-137 4.32E-09 4.32E-03 2.41E-10 2.41E-04 Xe-138 3.46E-08 8.66E-03 1.21E-09 3.01E-04 I-130 0.00E+00 0.00E+00 0.00E+00 0.00E+00 I-131 1.31E-09 6.53E-02 2.07E-13 1.03E-05 1.03E-12 5.14E-05 I-132 3.00E-10 1.00E-04 1.61E-12 5.36E-07 1.76E-11 5.86E-06 I-133 1.52E-09 1.52E-02 1.96E-12 1.96E-05 1.10E-11 1.10E-04 I-134 1.82E-10 9.11E-06 1.25E-12 6.25E-08 2.36E-11 1.18E-06 I-135 8.76E-10 1.25E-03 2.70E-12 3.86E-06 1.93E-11 2.75E-05 H-3 7.90E-05 3.95E+00 2.08E-05 1.038E+00 1.02E-09 5.10E-05 1.46E-09 7.29E-05
(DRN 02-110, R12;05-455, R14)
WSES-FSAR-UNIT-3
(DRN 02-110, R12)
Table 12.2-15a (Sheet 2 of 2)
Revision 14 (12/05)
Average Airborne C/DAC in Reactor Auxiliary Building, Turbine Building, Containment and Fuel Handling Building
(DRN 05-455, R14)
Isotope Containment Building Fuel Handling Building Turbine Building Reactor Auxiliary Building C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC C (uCi/cc)
C / DAC Cr-51 7.09E-10 8.87E-05 3.38E-13 4.23E-08 1.63E-13 2.03E-08 Mn-54 2.32E-09 7.73E-03 1.70E-13 5.67E-07 8.49E-14 2.83E-07 Fe-55 2.24E-09 1.12E-03 1.28E-13 6.42E-08 6.36E-14 3.18E-08 Fe-59 1.11E-10 5.53E-04 3.13E-14 1.56E-07 1.58E-14 7.90E-08 Co-58 2.64E-09 8.81E-03 4.97E-13 1.66E-06 2.43E-13 8.10E-07 Co-60 1.05E-09 1.05E-01 5.78E-14 5.78E-06 2.82E-14 2.82E-06 Np-239 3.87E-11 4.30E-05 2.00E-13 2.22E-07 1.04E-13 1.16E-07 Br-83 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Rb-86 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Sr-89 5.83E-11 9.71E-04 1.49E-14 2.48E-07 7.38E-15 1.23E-07 Y-91 2.50E-12 5.01E-05 5.49E-16 1.10E-08 2.75E-16 5.50E-09 Mo-99 1.34E-10 2.23E-04 6.02E-13 1.00E-06 3.08E-13 5.14E-07 Tc-99m 6.71E-12 6.71E-08 2.10E-13 2.10E-09 1.62E-13 1.62E-09 Te-127m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te-127 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Te-129m 5.28E-11 5.28E-04 2.03E-14 2.03E-07 9.98E-15 9.98E-08 Te-129 5.83E-12 1.94E-07 3.87E-13 1.29E-08 6.08E-13 2.03E-08 Te-131m 1.32E-11 6.58E-05 1.21E-13 6.06E-07 6.62E-14 3.31E-07 Te-132 4.26E-11 4.74E-04 1.61E-13 1.79E-06 8.26E-14 9.18E-07 Cs-134 6.13E-11 1.53E-03 3.64E-15 9.10E-08 1.81E-15 4.53E-08 Cs-136 8.57E-11 2.86E-04 8.32E-14 2.77E-07 4.13E-14 1.38E-07 Cs-137 1.02E-10 1.71E-03 5.35E-15 8.92E-08 2.60E-15 4.33E-08 Ba-137m 7.40E-16 1.23E-08 2.61E-15 4.36E-08 3.28E-16 5.47E-09
(DRN 02-110, R12;05-455, R14)
WSES-FSAR-UNIT-3 TABLE 12.2-16 (Sheet 1 of 4)
Revision 11-A (02/02)
REACTOR AUXILIARY BUILDING ROOM BY ROOM C/MPC AND WHOLE BODY DOSE COMMITMENT VALUES Dose Commitment (mrem/hr occupancy)
Elevation Leakage Rate Inhalation External Item Location and/or Component (ft. MSL)
(gpd)
C/MPC Whole Body Whole Body 1
Shutdown Cooling Heat Exchanger A and B
-35 1.0(1)(a) 2.29 3.15(-2) 7.98(-1) 2 Valve Operating Closure A and B
-15.5 1.0 4.58(-1) 6.30(-3) 1.60(-1) 3 Below Valve Operating Closure B Sump #6 and Pumps
-35 0
4.64 6.37(-2) 1.62 4
Containment Spray Pump A, LPSI Pump A, HPSI Pumps A and A/B, Sumps #7 and #8 and Pumps
-35 6.6 1.56 2.15(-2) 5.45(-1) 5 Containment Spray Pump B, LPSI Pump B, HPSI Pump B, Equip.
Drain Tk Pump, Reactor Drain Tk Pump, Sump #5 and Pump
-35 1.96(1) 4.64 6.37(-2) 1.62 6
Sump #1 and Pumps
-35 6.6 8.09(-1) 6.30(-2) 2.63(-1) 7 Equip. Drain Tk
-35 3.3 1.48(-1) 1.16(-2) 4.82(-2) 8 Emergency FW Pump B
-35 0
-(b) 9 Emergency FW Pump A
-35 0
10 Emergency FW Pump (Turbine Driven)
-35 0
11 Component Cooling Water Makeup Pumps A and B, Oil Sump
- 3 and Pump
-35 0
¨(DRN 00-1046) 12 Gas Surge Tank
-35 5.0(-3)scfm 5.46 1.53(-1) 1.30 (DRN 00-1046) 13 Gas Decay Tank C
-35 2.0(-2)scfm 1.66(2) 4.66 39.5 14 Waste Gas Compressor B
-35 2.0(-2)scfm 3.32 9.32(-2) 7.90(-1) 15 Gas Decay Tank B
-35 2.0(-2)scfm 1.66(2) 4.66 39.5 16 Waste Gas Compressor A
-35 2.0(-2)scfm 3.32 9.32(-2) 7.90(-1) 17 Gas Decay Tank A
-35 2.0(-2)scfm 1.66(2) 4.66 39.5 18 Charging Pump A
-35 6.6 4.79 4.72(-2) 1.96 19 Charging Pump A/B
-35 6.6 4.79 4.72(-2) 1.96 20 Charging Pump B
-35 6.6 4.79 4.72(-2) 1.96 21 Waste Tank, Waste Tk Pump B
-35 9.9 7.76(-2) 2.58(-2) 8.0(-4) 22 Waste Tank, Waste Tk Pump A, Sump #11 and Pump
-35 9.9 7.76(-2) 2.58(-2) 8.0(-4)
WSES-FSAR-UNIT-3 TABLE 12.2-16 (Sheet 2 of 4)
Dose Commitment (mrem/hr occupancy)
Elevation Leakage Rate Inhalation External Item Location and/or Component (ft. MSL)
(gpd)
C/MPC Whole Body Whole Body 23 Laundry Filter
-35 6.6 1.78(-1) 1.76(-1) 5.87(-6) 24 Oil Separator
-35 6.6 5.69(-1) 1.89(-1) 5.87(-3) 25 Laundry Tank A and B, Laundry Pump A and B, Detergent Sump
- 1 and Pumps
-35 19.8 5.20(-2) 5.11(-2) 1.71(-6) 26 Waste Filter
-35 6.6 5.69(-1) 1.89(-2) 5.87(-3) 27 Waste Condensate Pumps A and B, Chem. Waste Tank and Pump Sample Recovery Tank and Pump
-35 6.6 2.76(-2) 2.73(-2) 8.84(-8) 28 Waste Condensate Tanks A and B
-35 6.6 8.10(-3) 7.99(-3) 2.59(-8) 29 Sump #10 and Pumps, Plumbing Valve Pit, Refueling Storage Pool Leak Detection Station, Condensate Storage Pool Leak Detection Station
-35 0
30 Elevator Machine Room
-35 0
31 Holdup Tank 1-D
-35 3.3 2.81(-1) 2.05(-2) 9.02(-3) 32 Holdup Tank 1-B
-35 3.3 2.81(-1) 2.05(-2) 9.02(-3) 33 Holdup Tank 1-C
-35 3.3 2.81(-1) 2.05(-2) 9.02(-3) 34 Holdup Tank 1-A
-35 3.3 2.81(-1) 2.05(-2) 9.02(-3) 35 Acid Neutralizing Tank
-35 0
36 Boric Acid Makeup Tanks A and B, Boric Acid Pumps A and B
-35 19.8 4.59(-2) 3.17(-2) 2.58(-4) 37 Holdup Drain Pump, Holdup Recirc Drain Pump, Holdup Recirc Pump
-35 19.8 1.58 1.16(-1) 5.08(-2) 38 Boric Acid Preconcentrator Filter B
-35 6.6 2.67 1.95(-1) 8.57(-2) 39 Boric Acid Preconcentrator Filter A
-35 6.6 2.67 1.95(-1) 8.57(-2) 40 Shield Door Area
-35 0
41 Boric Acid Cond. Tanks A, B, C and D, Boric Acid Cond.
Pumps A and B, Sump #9 and Pumps
-35 2.64(1) 2.86(-2) 2.81(-2) 9.12(-7) 42 Waste Condensate Ion Exchanger
-35 3.3 2.08(-2) 2.04(-2) 6.63(-7) 43 Spent Resin Tank
-35 3.3 2.88(1) 1.01 4.31(-1)
WSES-FSAR-UNIT-3 TABLE 12.2-16 (Sheet 3 of 4)
Revision 305 (11/11)
Dose Commitment (mrem/hr occupancy)
Elevation Leakage Rate Inhalation External Item Location and/or Component (ft. MSL)
(gpd)
C/MPC Whole Body Whole Body 44 Corridor
-35 1.98 3.59(-2) 3.55(-4) 1.47(-2) 45 Purification Filter
-4 6.6 2.27(1) 3.12(-1) 7.90 (DRN 00-805, R11-B) 46 Flash Tank (c)
-4 3.3 2.06 2.83(-2) 7.19(-1) 47 Flash Tk Pumps A and B (c)
-4 1.32(1) 1.43 1.04(-1) 4.57(-2)
(DRN 00-805, R11-B)
(DRN 99-1032, R12) 48 Boronometer (Note: This device has been functionally abandoned-in-place by DC 3432.)
-4 (DRN 99-1032, R12) 49 Volume Control Tank
-4 3.3 8.71(-1) 8.60(-3) 3.57(-1) 50 Fuel Pool Filter
-4 3.3 9.05(-2) 7.38(-2) 2.44(-4)
(EC-4019, R305) 51 Chemical Addition Tank and Strainer / Zinc Inj. Skid
-4 9.9 3.64 2.84(-1) 1.19 (EC-4019, R305) 52 Deborating Ion Exchanger, Purification Ion Exchanger B
-4 6.6 4.45 6.12(-2) 1.55 53 Purification Ion Exchanger A, Fuel Pool Ion Exchanger
-4 3.3 1.86 2.56(-2) 6.48(-1) 54 Preconcentrator Ion Exchanger B, Boric Acid Cond. Ion Exchanger B
-4 6.6 1.07 7.81(-2) 3.43(-2) 55 Preconcentrator Ion Exchanger A. Boric Acid Cond. Ion Exchanger A
-4 6.6 1.07 7.81(-2) 3.43(-2) 56 Letdown Heat Exchanger and Strainer
-4 1.32(1) 3.21(1) 1.49 3.83 57 Blowdown Pumps A and B
-4 6.6 9.48(-2) 9.47(-2) 3.99(-5) 58 Filter Flush Tank and Pump
-4 9.9 1.21 9.47(-2) 3.95(-1) 59 Blowdown Heat Exchanger A and B
-4 3.0 4.26(-2) 3.34(-2) 1.37(-4) 60 Blowdown Filters A and B
-4 1.32(1) 7.24(-2) 5.90(-2) 1.96(-4) 61 Blowdown Demineralizers A and B
-4 6.6 3.95(-2) 3.22(-2) 1.07(-4) 62 Acid Storage Tank, Caustic Storage Tank and Heaters, Chemical Feed Tank and Pump
-4 0
63 Boric Acid Concentrator A
-4 1.65(1) 8.21(-2) 2.06(-2) 9.17(-3) 64 Boric Acid Concentrator B
-4 1.65(1) 8.21(-2) 2.06(-2) 9.17(-3) 65 Waste Concentrator
-4 1.65(1) 5.37(-2) 1.79(-2) 5.53(-4) 66 Pipe Penetration Area
-4 1.0(1) 1.03 4.18(-2) 1.86(-1)
WSES-FSAR-UNIT-3 TABLE 12.2-16 (Sheet 4 of 4)
Revision 11-B (06/02)
Dose Commitment (mrem/hr occupancy)
Elevation Leakage Rate Inhalation External Item Location and/or Component (ft. MSL)
(gpd)
C/MPC Whole Body Whole Body 67 Corridor
-4 1.32 1.08(-1) 5.0(-3) 1.29(-2) 68 Component Cooling Water Chemical Feed Tank
+21 0
69 Boric Acid Batching Tank and Strainer
+21 9.9 2.29(-1) 4.82(-2) 2.79(-3) 70 Waste Concentrate Storage Tank and Metering Pump
+21 6.6 1.02 7.81(-2) 1.47(-2) 71 Vault Area
-35 3.3 1.52(-3) 1.24(-3) 4.11(-6) 72 Aux. Component Cooling Water Pumps A and B
-35 0
73 Refueling Water Pool Purification Pump, Sump #3 and Pumps
-35 3.3 1.52(-3) 1.24(-3) 4.11(-6) 74 Blowdown Tank
-4 3.3 1.41(-1) 1.35(-1) 5.70(-5)
Notes (a) represents powers of 10 (b) represents clean
¨(DRN 00-805)
(c) The Flash Tank and pumps have been made inactive per ER-W3-00-0225-00-00.
WSES-FSAR-UNIT-3
(DRN 02-110, R12)
Table 12.2-16a (Sheet 1 of 4)
Revision 14 (12/05)
Reactor Auxiliary Building Room by Room C/DAC and Dose Commitment Values
(DRN 05-455, R14)
Item Location and / or Component Elevation Ventilation Leakage Dose Commitment (mRem/hr Occupancy)
(Ft. MSL)
Rate (CFM)
Rate (gpd)
Conc. C (uCi/cc)
C/DAC DDE Submersion CEDE Inhalation CDE-Thyroid Inhalation 1
Shutdown Cooling Heat Exchanger A&B
-35 1,500 10.0 1.61E-06 4.28E-02 6.94E-02 5.68E-03 9.14E-02 2
Valve Operating Closure A and B
-35 750 1.0 3.01E-07 8.56E-03 1.39E-02 1.14E-03 1.83E-02 3
Below Valve Operating Closure B Sump #6 and Pumps
-35 0
3.05E-06 8.68E-02 1.41E-01 1.15E-02 1.85E-01 4
Containment Spray Pump A, LPSI Pump A, HPSI Pumps A and A/B, Sumps #7 and #8 and Pumps
-35 1,450 6.6 1.03E-06 2.92E-02 4.74E-02 3.88E-03 6.24E-02 5
Containment Spray Pump B, LPSI Pump B, HPSI Pump B Equip. Drain Tk Pump, Rx Drain Tk Pump, Sump #5 & Pump
-35 1,450 19.6 3.05E-06 8.68E-02 1.41E-01 1.15E-02 1.85E-01 6
Sump #1 and Pumps
-35 300 6.6 4.96E-07 1.41E-02 2.29E-02 1.87E-03 3.03E-02 7
Equipment Drain Tank
-35 820 3.3 9.08E-08 2.58E-03 4.19E-03 3.43E-04 5.52E-03 8
Emergency FW Pump B
-35 0
CLEAN 9
Emergency FW Pump A
-35 0
CLEAN 10 Emergency FW Pump (Turbine Driven)
-35 0
CLEAN 11 Component Cooling Water Makeup Pumps A and B, Oil Sump #3 and Pump
-35 0
CLEAN 12 Gas Surge Tank
-35 365 0.05 scfm 3.33E-05 8.70E-01 1.50E+00 6.02E-02 1.35E-01 13 Gas Decay Tank C
-35 120 0.02 scfm 4.05E-04 1.06E+01 1.82E+01 7.31E-01 1.63+00 14 Waste Gas Compressor B
-35 600 0.02 scfm 8.09E-06 2.12E-01 3.64E-01 1.46E-02 3.28E-02 15 Gas Decay Tank B
-35 120 0.02 scfm 4.05E-04 1.06E+01 1.82E+01 7.31E-01 1.63E+00 16 Waste Gas Compressor A
-35 600 0.02 scfm 8.09E-06 2.12E-01 3.64E-01 1.46E-02 3.28E-02 17 Gas Decay Tank A
-35 120 0.02 scfm 4.05E-04 1.06E+01 1.82E+01 7.31E-01 1.63E+00 18 Charging Pump A
-35 400 6.6 3.66E-06 9.43E-02 1.68E-01 1.41E-03 2.26E-02 19 Charging Pump A/B
-35 400 6.6 3.66E-06 9.43E-02 1.68E-01 1.41E-03 2.26E-02 20 Charging Pump B
-35 400 6.6 3.66E-06 9.43E-02 1.68E-01 1.41E-03 2.26E-02 21 Waste Tank, Waste Tank Pump B
-35 1,100 9.9 2.38E-08 1.21E-03 1.17E-03 7.67E-04 1.23E-02 22 Waste Tank, Waste Tank Pump A, Sump #11 and Pump
-35 1,100 9.9 2.38E-08 1.21E-03 1.17E-03 7.67E-04 1.23E-02 23 Laundry Filter
-35 100 6.6 1.46E-07 3.72E-03 6.69E-03 5.62E-06 9.05E-05
(DRN 02-110, R12;05-455, R14)
WSES-FSAR-UNIT-3
(DRN 02-110, R12)
Table 12.2-16a (Sheet 2 of 4)
Revision 14 (12/05)
Reactor Auxiliary Building Room by Room C/DAC and Dose Commitment Values
(DRN 05-455, R14)
Item Location and / or Component Elevation Ventilation Leakage Dose Commitment (mRem/hr Occupancy)
(Ft. MSL)
Rate (CFM)
Rate (gpd)
Conc. C (uCi/cc)
C/DAC DDE Submersion CEDE Inhalation CDE-Thyroid Inhalation 24 Oil Seperator
-35 100 6.6 1.75E-07 8.90E-03 8.55E-03 5.62E-03 9.05E-02 25 Laundry Tank A and B, Laundry Pump A and B, Detergent Sump #1 and pumps
-35 1,025 19.8 4.28E-08 1.09E-03 1.96E-03 1.65E-06 2.65E-05 26 Waste Filter
-35 100 6.6 1.75E-07 8.90E-03 8.55E-03 5.62E-03 9.05E-02 27 Waste Condensate Pumps A and B, Chem Waste Tank and Pump, Sample Recovery Tank and Pump
-35 645 6.6 2.26E-08 5.77E-04 1.04E-03 8.72E-08 1.40E-06 28 Waste Condensate Tanks A and B
-35 2,015 6.6 7.25E-09 1.85E-04 3.32E-04 2.79E-08 4.49E-07 29 Sump #10 and Pumps, Plumbing Valve Pit, Refueling Storage Pool Leak Detection Station, Condensate Storage Pool Leak Detection Station
-35 0
CLEAN 30 Elevator Machine Room
-35 0
CLEAN 31 Holdup Tank 1-D
-35 540 3.3 3.61E-08 4.63E-03 2.11E-03 4.68E-03 7.54E-02 32 Holdup Tank 1-B
-35 600 3.3 3.61E-08 4.63E-03 2.11E-03 4.68E-03 7.54E-02 33 Holdup Tank 1-C
-35 540 3.3 3.61E-08 4.63E-03 2.11E-03 4.68E-03 7.54E-02 34 Holdup Tank 1-A
-35 600 3.3 3.61E-08 4.63E-03 2.11E-03 4.68E-03 7.54E-02 35 Acid Neutralizing Tank
-35 0
CLEAN 36 Boric Acid Makeup Tanks A and B, Boric acid Pumps A & B
-35 1,750 19.8 2.60E-08 8.15E-04 1.21E-03 1.93E-04 3.10E-03 37 Holdup Drain Pump, Holdup Recirc Drain Pump, and Holdup Recirc Pump
-35 810 19.8 1.60E-07 2.06E-02 9.38E-03 2.08E-02 3.35E-01 38 Boric Acid Preconcentrator Filter B
-35 160 6.6 2.71E-07 3.47E-02 1.58E-02 3.51E-02 5.66E-01 39 Boric Acid Preconcentrator Filter A
-35 160 6.6 1.20E-06 9.56E-01 1.58E-02 3.51E-02 5.66E-01 40 Shield Door Area
-35 0
CLEAN 41 Boric Acid Cond. Tanks A, B, C and D, Boric Acid Cond. Pumps A and B, Sump #9 and Pumps
-35 2,500 26.4 2.34E-08 5.96E-04 1.07E-03 8.99E-07 1.45E-05 42 Waste Condensate Ion Exchanger
-35 430 3.3 1.70E-08 4.33E-04 7.77E-04 6.54E-07 1.05E-05 43 Spent Resin Tank
-35 660 3.3 3.28E-06 4.21E-01 1.92E-01 4.26E-01 6.86E+00 44 Corridor
-35 19,840 1.98 3.71E-08 1.10E-03 1.71E-03 1.40E-04 2.26E-03
(DRN 02-110, R12;05-455, R14)
WSES-FSAR-UNIT-3 (DRN 02-110, R12)
Table 12.2-16a (Sheet 3 of 4)
Revision 305 (11/11)
Reactor Auxiliary Building Room by Room C/DAC and Dose Commitment Values (DRN 05-455, R14)
Item Location and / or Component Elevation Ventilation Leakage Dose Commitment (mRem/hr Occupancy)
(Ft. MSL)
Rate (CFM)
Rate (gpd)
Conc. C (uCi/cc)
C/DAC DDE Submersion CEDE Inhalation CDE-Thyroid Inhalation 45 Purification Filter
-4 100 6.6 1.49E-05 4.24E-01 6.87E-01 5.62E-02 9.05E-01 46 Flash Tank
-4 550 3.3 1.35E-06 3.85E-02 6.25E-02 5.11E-03 8.23E-02 47 Flash Tank Pumps A and B
-4 600 13.2 1.44E-07 1.85E-02 8.45E-03 1.87E-02 3.02E-01 48 Boronometer
-4 140 3.3 5.32E-06 1.51E-01 2.45E-01 2.01E-02 3.23E-01 49 Volume Control Tank
-4 1,100 3.3 6.65E-07 1.71E-02 3.05E-02 2.56E-04 4.11E-03 50 Fuel Pool Filter
-4 120 3.3 6.21E-08 1.77E-03 2.86E-03 2.34E-04 3.77E-03 (EC-4019, R305) 51 Chemical Addition Tank and Strainer / Zinc Inj.
-4 100 9.9 2.23E-06 6.36E-02 1.03E-01 8.43E-03 1.36E-01 (EC-4019, R305) 52 Deborating Ion Exchanger, Purification Ion Exchanger B
-4 510 6.6 2.92E-06 8.31E-02 1.35E-01 1.10E-02 1.77E-01 53 Purification Ion Exchanger A, Fuel Pool Ion Exchanger
-4 610 3.3 1.22E-06 3.47E-02 5.63E-02 4.61E-03 7.42E-02 54 Preconcentrator Ion Exchanger B, Boric Acid Cond. Ion Exchanger B
-4 400 6.6 1.08E-07 1.39E-02 6.33E-03 1.41E-02 2.26E-01 55 Preconcentrator Ion Exchanger A, Boric Acid Cond. Ion Exchanger A
-4 400 6.6 1.08E-07 1.39E-02 6.33E-03 1.41E-02 2.26E-01 56 Letdown Heat Exchanger and Strainer
-4 450 13.2 6.70E-06 3.91E-01 4.07E-01 1.97E-01 5.36E+00 57 Blowdown Pumps A and B
-4 1,350 6.6 2.89E-11 1.94E-05 9.23E-06 1.73E-05 4.47E-04 58 Filter Flush Tank and Pump
-4 300 9.9 7.45E-08 2.12E-03 3.44E-03 2.81E-04 4.53E-03 59 Blowdown Heat Exchanger A and B
-4 1,300 3.0 5.24E-09 2.22E-04 2.77E-04 9.82E-04 2.11E-03 60 Blowdown Filters A and B
-4 600 13.2 4.96E-08 1.41E-03 2.29E-03 1.87E-04 3.02E-03 61 Blowdown Demineralizers A and B
-4 550 6.6 2.71E-08 7.70E-04 1.25E-03 1.02E-04 1.65E-03 62 Acid Storage Tank, Caustic Storage Tank and Heaters, Chemical Feed Tank and Pump
-4 0
CLEAN 63 Boric Acid Concentrator A
-4 900 16.5 4.85E-08 2.47E-03 2.37E-03 1.56E-03 2.51E-02 64 Boric Acid Concentrator B
-4 900 16.5 4.85E-08 2.47E-03 2.37E-03 1.56E-03 2.51E-02 65 Waste Concentrator
-4 1,100 16.5 3.97E-08 2.02E-03 1.94E-03 1.28E-03 2.06E-02 66 Pipe Penetration Area
-4 6,710 10 3.38E-07 1.43E-02 1.79E-02 5.29E-03 1.36E-01 67 Corridor
-4 16,250 1.9 4.61E-10 9.20E-05 2.13E-05 1.74E-06 2.80E-05 (DRN 02-110, R12;05-455, R14
WSES-FSAR-UNIT-3
(DRN 02-110, R12)
Table 12.2-16a (Sheet 4 of 4)
Revision 14 (12/05)
Reactor Auxiliary Building Room by Room C/DAC and Dose Commitment Values
(DRN 05-455, R14)
Item Location and / or Component Elevation Ventilation Leakage Dose Commitment (mRem/hr Occupancy)
(Ft. MSL)
Rate (CFM)
Rate (gpd)
Conc. C (uCi/cc)
C/DAC DDE Submersion CEDE Inhalation CDE-Thyroid Inhalation 68 Component Cooling Water Chemical Feed Tank
+21 0
CLEAN 69 Boric Acid Batching Tank and Strainer
+21 630 9.9 4.84E-08 3.35E-03 2.48E-03 2.68E-03 4.31E-02 70 Waste Concentrate Storage Tank and Metering Pump
+21 400 6.6 4.37E-07 2.22E-02 2.14E-02 1.41E-02 2.26E-01 71 Vault Area
-35 7,130 3.3 1.04E-09 2.97E-05 4.82E-05 3.94E-06 6.35E-05 72 Aux Component Cooling Water Pumps A and B
-35 0
CLEAN 73 Refueling Water Pool Purification Pump, Sump
- 3 and Pumps
-35 7,130 3.3 1.04E-09 2.97E-05 4.82E-05 3.94E-06 6.35E-05 74 Blowdown Tank
-4 350 3.3 3.88E-08 2.14E-03 2.92E-05 3.57E-03 8.62E-04 Reactor Auxiliary Building 77,000 160 (Lb/day) 8.01E-08 2.23E-03 3.68E-03 2.98E-04 4.78E-03
(DRN 02-110, R12;05-455, R14
WSES-FSAR-UNIT-3 Table 12.2-17 Revision 14 (12/05)
(DRN 99-1098, R11; 03-2066, R14) 18-GROUP GAMMA-RAY SOURCE STRENGTHS PER FUEL ASSEMBLY 3 DAYS AFTER SHUTDOWN
(DRN 03-2066, R14)
E Mean (Mev)
Photons/sec 1.00E-02 2.86E+17 3.00E-02 8.47E+16 5.50E-02 3.94E+16 8.50E-02 4.77E+16 1.20E-01 1.22E+17 1.70E-01 2.94E+16 3.00E-01 1.02E+17 6.50E-01 2.12E+17 1.13E+00 1.83E+16 1.58E+00 3.36E+16 2.00E+00 2.81E+15 2.40E+00 1.68E+15 2.80E+00 2.44E+13 3.25E+00 9.77E+12 3.75E+00 2.63E+08 4.25E+00 3.33E+07 4.75E+00 1.96E+07 5.50E+00 1.74E+07
(DRN 99-1098, R11)
WSES-FSAR-UNIT-3 12.3-1 Revision 12 (10/02) 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES
¨(DRN 02-110)
Compliance with the design feature guidance specified in 10 CFR 20 and Regulatory Guide 8.8 is discussed in Subsection 12.1.2 in detail. The layout of plant radiation zones is described in Subsection 12.3.2.2 and Section 12.4. The counting room is described in Section 12.5. The locations of sampling ports are discussed in Subsection 9.3.2.
(DRN 02-110) 12.3.1.1 Description of Plant Shielding Plant layouts and cross sections of buildings containing process equipment for treatment of radioactive fluids, and also a plot plan are shown on Figures 1.2 - 1 through 27 and Drawing G136.
12.3.1.2 Primary Shield The primary shield consists of reinforced concrete which surrounds the reactor vessel. The primary shield is designed to meet the following objectives:
a) to attenuate the core neutron flux in order to limit the activation of component and structural materials, b) to limit the radiation level after shutdown in order to permit access to the Reactor Coolant System equipment, c) to reduce, in conjunction with the secondary shield and the neutron streaming shield, the radiation level from sources within the reactor vessel in order to allow limited access to the containment during normal operation, and d) to permit access during shutdown for inspections required by ASME,Section XI.
For purposes of primary shielding design, the normal full power operation of the core and resultant neutron and gamma fluxes are the controlling factors.
12.3.1.3 Secondary Shield
¨ (DRN 99-2362)
The secondary shield surrounds the primary shield and reactor coolant loops and attenuates, to a safe level, the radiation originating in the reactor coolant and steam generators. During full power operation, the major radiation source in the Reactor Coolant System is N-16, which is created by neutron activation of oxygen during passage of coolant through the core.
The secondary shield is designed to permit limited access to certain areas within the Reactor Building during full power operation. The secondary shield also serves to reduce the full power radiation levels outside the Reactor Building so that normal continuous occupancy outside the Reactor Building is afforded.
In addition, the secondary shield helps in reducing the radiation intensity outside the Reactor Building in the unlikely event of an accidental release of fission products into the containment. This function, however, is primarily accomplished by the Reactor Building concrete structure.
WSES-FSAR-UNIT-3 12.3-2 Revision 306 (05/12)
After reactor shutdown, the fission and corrosion product activities in the Reactor Coolant System listed in Tables 11.1-2 and 11.1-10 become the dominant sources.
12.3.1.4 Fuel Transfer Shield (EC-14275, R306)
The fuel transfer shield protects plant personnel from fission product gamma radiation emitted from the spent fuel elements during core refueling operations. The fuel is removed from the reactor through a canal to a water filled spent fuel storage area located in the Fuel Handling Building. After sufficient decay, the spent fuel is transferred under water to canisters within the transfer cask.
The fuel transfer shielding consists of five parts:
(EC-14275, R306) a)
The water and concrete of the refueling cavity, spent fuel storage areas and the transfer canal.
b)
The fuel transfer tube shield structure which shields the space between the transfer canal and the steel containment structure, and the space between the steel containment and the concrete Shield Building (annulus).
c)
The fuel transfer tube in the gap between the Reactor Building and the Fuel Handling Building is shielded by a crescent shaped lead shield covering the top and sides of the tube.
(EC-14275, R306) d)
The spent fuel canisters within the transfer cask are loaded underwater.
e)
The spent fuel transfer cask contains and shields the loaded canister. The transfer cask is a steel, lead, steel layered cylinder with a water jacket attached to the exterior. The loaded canister has a steel shielded lid to allow access for sealing operations.
(EC-14275, R306)
The fuel transfer tube shield structure (part b above) is a combination of concrete, steel, lead and silicone foam. The design objective of the shield structure is to enclose the fuel transfer tube by shielding materials to prevent any inadvertent exposure to this high radiation source. Access openings are provided for periodic leakage inspections and any necessary testing of the fuel transfer tube. An access way below the shield structure in the annulus is incorporated into the design to provide an alternate exit route in the event of a fire or other accident in the annulus. The shield design has provisions for the expansion of the steel containment vessel during normal operations, and also for the maximum expansion during and following a LOCA.
The general dose rate limits used in the shield design are:
a) 5.0 mrem/hr for the areas most likely to be occupied and away from narrow gap areas.
These areas include from floor level up to head height on the outside of the shield structure and away from narrow shield gaps.
b) 25 mrem/hr for the areas of infrequent occupancy and narrow exposure areas. These areas include elevation above head height outside the shield structure and the narrow shield gaps.
Areas that exceed the 25 mrem/hr (i.e., Shielding Defective Areas) may be posted with radiological signs/barricades to restrict personnel occupancy.
WSES-FSAR-UNIT-3 12.3-3 Revision 14-B (06/06)
(DRN 99-2362, R11;06-319, R14-B)
The 5.0 mrem/hr allows 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per week occupancy without exceeding 100 mrem/week and this level is not exceeded in the occupied locations around the fuel transfer tube shielding. The areas that would see infrequent occupancy, and those that have narrow exposure areas are allowed to have higher dose rates. This is a design compromise that reduces the weight of shielding-material (welded to containment vessel), and the volume of shielding material so as not to overly restrict the access opening into the shield structure. It is assumed that a spent fuel assembly will pass through the transfer tube in one minute and that a maximum of 358 assemblies (217 offload + (217 - 76) maximum reload) will be transferred during a refueling outage. A dose rate of 5.0 mrem/hr will result in a total exposure of thirty mrem maximum if a person is present for all transfers. If it is assumed that a maximum of 32 spent fuel assemblies are transferred in an eight hour shift, then the average exposure per shift will be 2.7 mrem for the same conditions. A dose rate of 25 mrem/hr will result in exposures five times greater than those cited above; however, the larger dose rates apply to areas of infrequent occupancy and the narrow gaps. Thus, the actual doses in these areas are expected to be much lower. The narrow gaps are the two in. gaps between the steel containment vessel and the concrete, steel or lead shields. These gaps are shielded by compressable silicone foam and steel. Personnel will have a low exposure at these gaps if they approach the shield structure and press against the steel containment vessel. In these cases the exposures will be to the extremities, as defined in 10CFR20.1003.
(DRN 99-2362, R11;06-319, R14-B)
(DRN 99-1098, R11)
The refueling canal, the fuel transfer tube, the steel containment vessel, the concrete containment building, and the shield structure were originally modelled for dose rate calculations on the SPAN-4 program (Reference 12). This is a point kernel shielding program that uses three dimensional rectangular, cylindrical and spherical geometries in any combination to describe the physical features and radiation sources. For this application the exact geometry was modelled and no compromise of shape was necessary. The program uses Gaussian quadrature, and the Gauss point distributions are determined by Legendre and/or Laguerre polynomials. The data for calculations, such as, cross-sections, buildup factors, material compositions, energy structures, flux to dose conversion factors and Gauss quadratures are contained in the program library, and this data can be changed or added to by the user. For this application, considerable attention was given to the Gauss point distributions for the calculations through the narrow gaps. Trial and error and an excess of Gauss points were used in the narrow gap calculations to assure that the dose rates were adequately calculated.
As part of the spent fuel pool reracking, licensed in 1998, the fuel transfer tube was reanalyzed using the QAD-CGGP program. The results are documented in Reference 13.
(DRN 99-1098, R11)
The fuel transfer tube shield structure has two parts, which are separated by the steel containment vessel. The first part is between the steel containment vessel and the refueling canal, and it completely blocks off this passageway up to elevation 21.00 feet.
The second part is between the steel containment vessel and the concrete reactor building wall. There is a passageway beneath this shield structure to provide an alternate exit from this annular area. These two parts of the shield structure are not connected at any point, and each is physically separated from the steel containment vessel by a two in. gap. This gap provides space for thermal expansion of the steel vessel.
The first part of the shield structure consists of:
a) a four ft. thick column on the west side between the floor and elevation 17.00 feet.
WSES-FSAR-UNIT-3 12.3-4 Revision 11 (05/01) b)
a four ft. thick column on the east side between the floor and elevation 5.50 ft. and 13.50 ft. and 17.00 ft. The space between elevations 5.50 ft. and 13.50 ft. is shielded by stacked lead brick to a thickness of 1.0 ft., which are retained in place by a removable steel frame. The lead brick can be removed to provide access to the interior of this shield structure. The access width is about 33 in.
c) a four ft. thick slab on the top of the structure with a length of 18.5 ft. The top of the slab is at an elevation of 21.00 ft.
d) a bottom shield of concrete filled between the side shield columns (a and b above) to an elevation of 6.00 ft. A one ft. by two ft. lead brick slab is embedded in this lower shield adjacent to the steel containment vessel. This lead brick is needed to shield the diagonal radiation paths through the two in. gaps and the two in. steel containment vessel to the space below the shield in the adjacent annulus.
e) a filler material in the two in. gap to provide shielding and flexibility for thermal expansion of the steel vessel. The material chosen for this application is a cellular polymeric silicone material (silicone foam) which is densified with powdered lead to a total density of 100 pounds per cubic foot. The steel containment vessel is expected to have its minimum radius (at its minimum temperature) at the time of construction, so the gap inside the containment will range from two in.
to approximately three in. for a short time following a LOCA. Under normal operating conditions the maximum gap is expected to be approximately 2.25 in. The growth in the gap width can be compensated for by installing the silicone foam with an initial compression so that it will automatically expand as the gap grows. This is accomplished by precasting the silicone foam at an appropriately lower density and installing it against the inside surface of the steel containment vessel. The weight of the wet concrete poured against it will compress it to the desired density and develop the necessary initial compression. For a 1/4 in. deflection in two in. the pressure developed is estimated to be approximately 13 pounds per square inch. It is not considered necessary to compensate for a gap growth of greater than 1/4 in., since the fuel transfer tube will not be used during accident conditions.
f)
Two silicone foam blocks adjacent to the stacked lead bricks and the steel containment vessel.
The smaller block fills the outer portion of the W8x24 post and the larger block is adjacent to the west side of the W8x24 post and the steel containment vessel. The latter block is fastened to the containment vessel so as to move with its expansion and contraction. The two blocks extend from elevations 6.00 ft. to 13.00 ft. It is not necessary to fill the inner portion of the W8x24 post with silicone foam since its contribution in reducing dose rates is negligible. The silicone foam material for this application is lead loaded to a total density of 150 pounds per cubic foot, and they are not subject to compressive loads. The size of the blocks are limited so as not to overly restrict the access opening size or the maneuvering space inside the shield structure.
The second part of the shield structure consists of:
a) a four ft. thick column on the west side between elevations 4.625 ft. and 17.00 ft.
WSES-FSAR-UNIT-3 12.3-5 Revision 11 (05/01) b)
a four ft. thick column on the east side between elevation 13.50 ft. and 17.00 ft. The space between elevations 5.50 ft. and 13.50 ft. is shielded by stacked lead bricks to a thickness of 1.0 ft.
Its function and dimensions are similar to item b) under the first part of the shield structure.
c) a four ft. thick slab on the top of the structure with a length of 18.5 ft. The top of the slab is at an elevation of 21.00 ft.
d) a 1.5 ft. thick steel slab on the bottom with its top elevation at 6.00 ft. It is supported by steel columns, which provide a passageway beneath the shield with a width of approximately 25 in. and a height of 6.0 ft.
e) five steel slabs with cross sectional dimensions of 3 x 9 in. These slabs are retained against the containment vessel by steel brackets which are welded to the steel containment vessel, and as close to the main shield structure as construction will permit. The nine in. dimension is adjacent to the steel vessel. The slab on the west side of the shield structure extends from elevation 3.00 ft. to 21.75 ft. Two of the slabs are on the east side of the structure between elevation 3.00 ft. and 4.63 and elevation 13.50 ft. and 21.75 feet. A third slab in the east side is adjacent to the stacked lead bricks between elevation 4.63 ft. and 13.00 ft. The fifth slab is on the top and it fills the space between the side slabs. The two in. gaps between the concrete shields and the steel containment vessel are air gaps, the silicone foam filler material cannot be used on this side of the containment vessel since the approximate one in. maximum expansion of the containment vessel would result in excessive pressures (buckling forces) that were not included in the vessel design.
f)
The two silicone foam blocks on the outside of the containment vessel and adjacent to the stacked lead brick serve the same function as the two blocks in item f) under the first part of the shield structure. These blocks are also lead loaded to a density of 150 pounds per cubic foot.
g)
A silicone foam slab with a thickness of three in. is placed below the 18 in. thick steel slab. This shield is used to reduce the dose rates in the passageway below the steel slab in the annulus.
The silicone foam for this application is 150 pounds per cubic ft., and it is not subject to compressive forces. It is shaped around the support posts to avoid compression when the containment vessel is expanded to its maximum dimension. This shield extends from the floor to elevation 4.50 ft., and azimuthally along the containment vessel for the same length as the steel slab.
All of the shield components are shaped to conform to the curvature of the steel containment vessel.
Following an accident inside the Containment Building it would be necessary to inspect these shields to determine if any have shifted position.
Gamma-ray dose rates from the spent fuel in transit in the fuel transfer tube were calculated with the QAD-CGGP program (Reference 14). QAD is a combinatorial geometry point kernel code system that uses the geometric progression buildup factor. The source-term input for QAD was determined with SAS2H-ORIGEN-S (Reference 15), which gave the 18-group gamma-ray source strengths shown in Table 12.2-17. The gamma intensities shown in the table are those for a single, average assembly with an initial fuel enrichment of 5.5% (Note: enrichment used for conservative calculation purposes only; licensee can not posess fuel of this enrichment level) and a fuel exposure of 70,000 Mwd/mtU. In calculating the dose rates, a peaking factor of 1.8 was used in the QAD program.
The fuel carrier, which holds the spent fuel assemblies as they move through the fuel transfer tube, is equipped to carry two assemblies simultaneously; the assemblies are essentially piggy-backed one
WSES-FSAR-UNIT-3 12.3-6 Revision 14 (12/05)
(DRN 99-1098, R11) above the other, forming a gamma source that is long horizontally, tall vertically, and narrow laterally.
This two-assembly source assumed for the dose rate calculations is shielded by the water in the inner steel tube of the transfer-tube assembly, the inner tube itself, and the outer steel tube of the transfer-tube assembly. The source is centered vertically and horizontally within the transfer tube.
The gamma dose rate from the two, in-transit, spent fuel assemblies was calculated at six locations. The locations and the calculated dose rates at those locations are described in the following paragraphs.
The dose rate at the outer surface of the transfer tube assembly, at its vertical mid-height and in the annulus between the steel and concrete of the reactor building, is 86.4 x 106 mR/hr.
(DRN 03-2066, R14)
At a point whose elevation is that of the vertical mid-height of the transfer tube, on the east face of the special shielding to the east of the transfer tube, the dose rate is calculated to be 1.60 x 105 mR/hr in the annulus between the steel and concrete of the reactor building. This high dose rate is very conservative, for the calculation did not consider the shielding benefit of the lead-loaded silicon foam (silfoam) material that fills the interstice between the reactor building steel and the north-south lead shielding.
(DRN 03-2066, R14)
For a dose rate point 9-9 west of the transfer tube centerline, and at the elevation of the tubes vertical mid-height, the dose rate (through the 4-foot concrete shield wall) is 36.4 mR/hr in the annulus between the steel and concrete of the reactor building.
At a point between the refueling canal wall and the reactor building steel at elevation 21.00 (on a concrete floor which has a thickness of 4 feet), above the transfer tube, the dose rate is 0.18 mR/hr.
For a location between the refueling canal wall and the reactor building steel, at elevation 21.00 and at the top of a crack that is actually filled with silfoam, but is assumed to be air, the dose rate is 1.79 x 105 mR/hr. Again, this high dose rate results from the assumption the crack is empty and the dose rate point looks directly at the unshielded transfer tube. The conservatism introduced by this assumption can be seen by comparing the high dose rate with that for a point shifted slightly so that there is no unshielded view of the transfer tube; in the latter case, the dose rate drops to less than 0.20 mR/hr.
In a related calculation, the dose rate at the bottom of the silfoam-filled crack of the preceding calculation, which is at elevation 17.00, the dose rate is 2.73 x 105 mR/hr.
Most of the dose rates given above represent conservative maximums, and this was the intent in performing the calculations. The conservatism was introduced into the calculations a number of ways, including the following specified input assumptions.
Two fuel assemblies for the source, each assumed to have the maximum source term.
A fuel exposure of 70,000 Mwd/mtU, with 5.5% initial enrichment (Note: enrichment used for conservative calculation purposes only; licensee can not posess fuel of this enrichment level).
A conservative peaking factor of 1.8 for both fuel assemblies. A cooling time of 3 days.
The use of AP (anterior-posterior) gamma-ray fluence-to-dose conversion factors (Reference 16).
The dose rates actually experienced by personnel should be significantly lower than the hypothetical dose rates shown above. In regions between the fuel transfer tube and the special shielding (regions not accessible to personnel), the dose rates may approach the values given above during periods when fuel assemblies are in transit.
(DRN 99-1098, R11)
WSES-FSAR-UNIT-3 12.3-7 Revision 11 (05/01)
The fuel transfer tube is shielded in the space between the Reactor Building and the Fuel Handling Building (Part C) by a crescent shaped lead shield.
12.3.1.5 Shield Building The Shield Building is a reinforced concrete structure with cylindrical wall three ft. thick and a 2.5 ft. thick dome. In conjunction with the primary and secondary shields, it limits the radiation level outside the Shield Building from all sources inside the containment to no more than 0.25 mrem/hr at full power operation.
The Shield Building provides protection to plant personnel from radiation sources inside the containment following a design basis accident. These radiation sources are discussed in Subsection 12.2.1.9. The Shield Building walls will act to greatly attenuate the direct offsite gamma dose following a design basis accident.
12.3.1.6 Neutron Streaming Shield The potential for radiation streaming (neutron and gamma) through the annulus around the reactor vessel has been analyzed to determine the radiation fields that could occur in areas of containment which may require occupancy.
Because operating experience indicates that streaming gamma dose rates during operation are a relatively small fraction of the corresponding neutron dose rates, the analyses of the streaming dose rates in containment has been limited to the determination of neutron dose rates. The angular neutron flux as a function of energy which emerges from the surface of the reactor vessel at selected locations has been derived using the DOT 3(5) computer program utilizing an S4 angular quadrature and a P3 Legendre expansion coefficient for anisotropic scattering. Refer to Subsection 12.2.1.2 for a description of these sources.
These vessel emergent angular fluxes have been used as input to a Monte Carlo analysis of the streaming problem.
Morse-CG(6) a general purpose Monte Carlo multigroup neutron (and gamma ray) transport code with combinational geometry, has been used to compute the neutron streaming and the resultant dose rates at various locations inside containment.
The cross section library used in the calculations is based on DLC-23 or CASK library(7). This library is a coupled neutron and gamma ray library and the data in the library are obtained by collapsing cross sections over a PWR core spectrum.
The geometry model used for the containment dose rate calculation includes the containment vessel steel shell and Shield Building concrete, the major features of the containment internal structures such as the refueling cavity, the shield walls around the steam generators and pumps, and a detailed description of the reactor vessel cavity, the reactor vessel, the primary piping, the missile shield, the primary shield, and the ring girder support.
A check on the accuracy of the modelled geometrical representation of the reactor cavity and containment has been obtained by verification by means of computer generated pictures of the model taken at different elevations and sections.
The ring girder support of the reactor vessel has been designed in such manner as to also provide shadow shielding against neutrons and gamma rays streaming up the annular gap between the vessel
WSES-FSAR-UNIT-3 12.3-8 Revision 11 (05/01) and the cavity walls. The ring girder design also reduces the neutron and gamma ray streaming through the primary penetrations. Vessel cooling and insulation requirements impose a limitation on the minimum ring girder to vessel gap that can be achieved.
The neutron histories start on the surface of the reactor pressure vessel with an energy and a direction determined by processing the annular fluxes at the outermost mesh of the vessel determined by an R-Z DOT 3 calculation, with the DOMINO code(8), which is explicity set up to provide the proper source information for the MORSE program. Variation in source strength along the circumference has been neglected for conservatism.
The MORSE calculation has been performed in two stages by a MORSE to MORSE coupling technique.
The first stage stops the random walk at the vessel flange level. Particle escapes at the flange as computed in the first stage are written to collision tapes which are used as inputs to subsequent MORSE runs utilizing the entire containment model.
No biasing has been used for either stage of the calculation. The Monte Carlo calculation of the dose rates in containment, however, has been limited to the energy group of neutrons emerging from the vessel with energies above 0.11 MeV in the interest of saving computer time and cost, as it is known from experience at operating plants that the dose rates in containment are due primarily to fast neutrons. The contribution to the containment dose rates from the lower energy groups has been estimated to be approximately three percent. The response to the energy groups considered is followed down to thermal energies.
Table 12.3-4 lists the estimated neutron dose rates at selected points within containment. These values are corrected for a factor of two conservatism in the calculated vessel emergent fluxes. The uncertainty in the Monte Carlo calculation is in the neighborhood of 30 percent.
While the computed neutron dose rates on the containment operating floor are relatively high, the dose rates in the general areas of the lower floor where personnel may require access are expected to be much lower. Computed values of the floor dose rates are less reliable due to the difficulty in modelling the problem and the large uncertainties in any ensuing results; however a reasonable estimate of the expected levels can be made by comparison of the dose rates measured at several operating plants on the lower floors with the corresponding dose rates on the operating floors of the same plants, and scaling of these lower floor dose rates by the ratio of the predicted Waterford 3 operating floor dose rates to the measured operating dose rates.
Measurements conducted at Calvert Cliffs (9), St. Lucie 1 (10), and Millstone 2 (11), scaled for full power, indicate that neutron dose rates at locations where the Waterford 3 dose rates are approximately 20 rem/hr, range from 60-65 rem/hr.
Middle floor dose rates in these plants range from 75-900 mrem/hr, while bottom floors dose rates are in the range of 15-250 mrem/hr.
Expected lower floors neutron dose rates for Waterford 3 should therefore be one third of those of the referenced operating plants, and can thus range from 25-300 mrem in the middle floor, and from five up to 83 mrem in the bottom floors.
Streaming gamma dose rates at the same plants were measured to be one fourth or less of the neutron dose rates for general containment areas. A similar ratio is expected for this plant.
WSES-FSAR-UNIT-3 12.3-9 Revision 14 (12/05)
Information was obtained relative to the estimated frequency, length of visit and number of persons to enter the containment during full power operation. Experience has indicated that most of the visits are related to instrument failures. The visits anticipated for Waterford 3 are listed below.
For the series of cabinets 1A, 1B, 1C, 1D, 2A, 2B, 2C and 2D at elevation +21.0 ft. MSL, the I&C personnel expect a visit once every two months for a time of one hour and with three persons. Other visits by I&C personnel at middle and lower levels are anticipated to be once every six months for two hours each visit and with three persons.
The operations personnel expect to visit the containment during full power operation at the middle and lower levels once every two months for 30 minutes per visit and with two persons. It is also anticipated by the operating personnel that visits to the operating floor will be made once every two months for five minutes each visit, and by two persons. These visits would be limited to the outer periphery of the operating floor where the dose rates are estimated to range from 700 to 3000 mrem/hr.
The annual man-rem contribution from these containment visits are conservatively estimated to range from 1.6 to 13.8 man rem using the dose rate estimates for the middle floor levels (25 to 300 mrem/hr) and 700 to 3000 mrem/hr for the operating floor.
12.3.1.7 Reactor Auxiliary Building Shielding Reactor Auxiliary Building shielding includes concrete walls, covers, and removable blocks which will shield the sources of radiation originating from the Chemical and Volume Control System, Boron Management System, Safety Injection System, the Waste Management System, and portions of the Fuel Pool System. Typical components which require shielding include holdup tanks, decay tanks, demineralizers, fitters, heat exchangers, and associated piping. Activities in these systems are based upon normal system operation with clad defects in fuel rods generating one percent of rated core thermal power and are specified in Tables 12.2-7 through 12.2-11.
12.3.1.8 Main Control Room For purposes of designing main control room shielding, the radioactivity releases from the maximum loss of coolant accident are controlling. The two sources considered in designing the shielding are:
a)
Direct gamma radiation from the containment atmosphere and emergency filters.
(DRN 03-2066, R14)
Alternative Source Terms based on Regulatory Guide 1.183, release fractions and plate-out fractions are considered. A uniform distribution of radioactivity within the containment is assumed. Credit for post-accident decay is considered.
(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 12.3-10 Revision 309 (06/16)
(DRN 03-2066, R14)
Doses to Control Room personnel following a LOCA are presented in Section 15.6. Credit for seven ft. of concrete (three ft. for the containment shield wall and four ft. for the main control room shielding) was taken. A minimum shield thickness of four ft. separates the emergency filters from the main control room.
b)
Direct Gamma Radiation from Radiation Leakage External to Containment (Cloud)
Thirty day post-LOCA Control Room doses are reported in FSAR Section 15.6.3.3.5 and FSAR Table 15.6-18.
(DRN 03-2066, R14)
In addition to shielding from external exposures, the main control room is designed to be pressurized under accident conditions with filtered makeup and recirculation in order to minimize the quantity of airborne radioactivity which enters the main control room and thereby ensure compliance with GDC 19 of 10CFR50. A detailed description of the system design is provided in Subsection 9.4.1. Chapter 15 includes an evaluation of the exposures to main control room personnel following the design basis accident.
12.3.1.9 Fuel Handling Building (EC-14275, R306)
Shielding is provided for protection during all phases of spent fuel removal and storage. Operations requiring shielding of personnel are spent fuel removal from the reactor, spent fuel transfer through the refueling canal and transfer tube, spent fuel storage, spent fuel transfer cask loading, spent fuel canister seal welding, vacuum drying, helium backfilling, movement into a storage cask, and unloading from a storage cask into the spent fuel pool; and maintenance and inspection of the spent fuel pool purification loop of the Fuel Pool System.
(EC-14275, R306)
(DRN 99-1098, R11)
Since all spent fuel removal and transfer operations are carried out under borated water, a minimum water depth above the top of the fuel assemblies is established to provide radiation shielding protection.
The dose rate at the water surface is less than 15 mrem/hr. The concrete walls of the fuel transfer canal and spent fuel pool supplement the water shielding and limit the maximum continuous radiation dose levels in working areas to less than 15 mrem/hr from spent fuel sources.
(DRN 99-1098, R11)
(LBDCR 16-016, R309)
The refueling water and concrete walls also provide shielding from activated control element assemblies (CEAS), reactor internals removed at refueling times, excores, filters, canisters, and other radiological waste less active than spent fuel. Although dose rates will generally be less than 2.5 mrem/hr in working areas, certain manipulation of fuel assemblies, CEAs, or reactor internals may produce areas where dose rates exceed 2.5 mrem/hr for short periods. However, the radiation levels will be closely monitored during refueling operations to establish the allowable exposure times for plant personnel in order not to exceed the dose limits specified in 10CFR20.
(LBDCR 16-016, R309)
The spent fuel pool shielding is based upon the following considerations:
WSES-FSAR-UNIT-3 12.3-11 Revision 306 (05/12)
The controlling factor in the design of the spent fuel pool and fuel transfer canal walls are the irradiated fuel assemblies.
(DRN 99-1098, R11)
The shielding design of the fuel transfer canal is based upon consideration of a spent fuel assembly with a source strength 1.8 times greater than that derived from Table 12.2-17.
For the spent fuel pool (from Reference 13):
a)
In the room below the spent fuel pool, the dose rate from the full pool of spent fuel. The fuel in the pool will consist of 48 hot assemblies (3 day cooled, 1.8 peaking factors) and the remainder of the pool will be filled with assemblies which are 1 year cooled with peaking factors of 1.00. The dose rate is 4.17 mR/hr at an elevation of -30.00 (five feet above the floor).
b)
In the room below the spent fuel pool, the dose rate from the full pool of spent fuel. The fuel in the pool will consist of 116 hot assemblies (5 day cooled, 1.8 peaking factors), 101 assemblies (5 day cooled, 1.4 peaking factors) and the remainder of the pool will be filled with assemblies which are 1 year cooled with peaking factors of 1.00. The dose rate is 11.7 mR/hr at an elevation of -
30.00 (five feet above the floor).
c)
At the pipe chase at the wall north of the fuel stored in the spent fuel pool racks. The northern-most row (outer row of assemblies next to the wall) will be 1 year cooled, 1.00 peaking factor assemblies. The remaining assemblies which form the source (interior rows) will be 3 day cooled, 1.80 peaking factor assemblies. The dose rate in the pipe chase is.17 mR/hr.
d)
At the pipe chase at the wall north of the fuel stored in the spent fuel pool racks. The northern-most row (outer row of assemblies next to the wall) will be assumed to be water. The remaining assemblies which form the source (interior rows) will be 3 day cooled, 1.80 peaking factor assemblies. The dose rate in the pipe chase is 2.15 mR/hr.
e)
In the fuel pool cooling pump room with the refueling canal racks fully filled with fuel (all assemblies will be 1 year cooled with 1.00 peaking factors). The dose rate is 0.05 mR/hr.
f)
At the pipe chase at the wall north of fuel stored in the refueling canal racks (all assemblies will be 1 year cooled with 1.00 peaking factors). The dose rate is 0.05 mR/hr.
g)
At the north face of Gate #3B, in the Cask Decontamination Pit, from a single assembly moved (suspended approximately 7 underwater from the Spent Fuel Handling Machine) in the Cask Storage Pit. The dose rate from the fuel assembly (5.5% initial enrichment, 70,000 Mwd/mtU bumup, 3-day cooling, 1.8 peaking factor) is 52.2 mRem/hr. The fuel assembly is no closer than above the eighth row of cells from the gate (Gate #3A). If, for any reason, the need arises to store irradiated fuel closer (than seven spaces) to Gate #3A during normal SFP operation the appropriate shielding calculations will be performed, prior to placing fuel into these cells, to ensure that dose rates, in this area, will remain acceptable.
(EC-14275, R306) h)
In the Rail Bay Area from a single assembly moved (suspended approximately 7 underwater from the Spent Fuel Handling Machine) in the Cask Storage Pit. The dose rate point in this calculation is farther away from the source than the point in g above and receives additional shielding from Gate #4 (the gate north of the Cask Decontamination Area). The separation distance of the fuel assembly from Gate #3A (the gate leading to the Cask Decontamination Area) will be the same as in g above. The dose rate from the fuel assembly (5.5% initial enrichment, 70,000 Mwd/mtU burnup, 3-day cooling, 1.8 peaking factor) is 0.75 mRem/hr. If, for any reason, (DRN 99-1098, R11; EC-14275, R306)
WSES-FSAR-UNIT-3 12.3-12 Revision 14 (12/05)
(DRN 99-1098, R11) the need arises to store irradiated fuel closer (than seven spaces) to Gate #3A during normal SFP operation the appropriate shielding calculations will be performed, prior to placing fuel into these cells, to ensure that dose rates, in this area, will remain acceptable.
i)
At the surface of the pool from an assembly suspended 74 below the surface of the water (3 day cooled, 1.80 peaking factor assembly). The dose rate directly overhead is 9.07 mr/hr. The maximum dose rate, slightly away from the vertical position, is 10.8 mR/hr.
j)
At the surface of the pool from typical, measured, radionuclides in the pool water. The dose rate is 0.21 mR/hr.
k)
At the surface of the pool from the expected (FSAR Table 11.1-17) radionuclides in the pool water. The dose rate is 7.29 mR/hr.
I)
At the surface of the pool from the maximum (FSAR Table 11.1-17) radionuclides in the pool water. The dose rate is 72.7 mR/hr.
(DRN 99-1098, R11)
Shielding for the Fuel Pool System is based upon source terms derived from normal system operation as specified in Table 12.2-9.
12.3.2 SHIELDING 12.3.2.1 Design Objectives The primary design objective of the plant radiation shielding is to protect plant operating personnel and the general public against radiation exposure from the reactor, power conversion, auxiliary, and waste processing systems during normal operation, including anticipated operational occurrences, postulated accident conditions, and maintenance.
This objective is accomplished by designing the shielding to perform the following functions:
a)
Limit inplant exposure to radiation of plant personnel, contractors, and authorized site visitors to as far below the limits set forth in 10CFR20 as reasonably achievable for normal operation, including anticipated operational occurrences and maintenance, in conformance with Regulatory Guide 8.8 (March 1977).
b)
Limit radiation exposure of main control room personnel, in the unlikely event of an accident, to allow habitability of the main control room as specified in 10CFR50, Appendix A, Criterion 19, by limiting the total integrated dose over 90 days following the accident to three rem.
(DRN 05-144, R14) c)
Limit exposures to the general public offsite from direct and air scattered radiation to a small fraction of the limits set forth in 10CFR20 during normal operation and anticipated operational occurrences, and to within the limits specified in 10CFR50.67 for postulated accident conditions.
(DRN 05-144, R14) d)
Provide barriers for restricting personnel access to high radiation areas and for controlling the spread of contaminants. The plant radiation shielding is also designed to protect certain plant components from excessive radiation damage or activation.
WSES-FSAR-UNIT-3 12.3-13 Revision 11 (05/01)
To accomplish this objective, the plant shielding functions to:
1)
Reduce neutron activation of equipment, piping, supports and other materials by the use of suitable shielding around the reactor vessel, designed to minimize neutron streaming into the reactor cavity upper reaches, steam generator subcompartments, and general containment spaces.
2)
Limit radiation damage to equipment and materials to below the specific integrated life dose limits.
To comply with the above objectives, the plant shielding is designed to attenuate radiation levels throughout the plant, from direct and scattered neutron and gamma radiation to the dose limits specified in Table 12.3-1.
In part, the criteria which is used in determining shielding requirements for pumps and valve galleries are the following:
a)
The dose rate in the near vicinity (few ft.) from the equipment in question.
b)
The annual exposure time anticipated for personnel with respect to the specific equipment.
Exposure times are classified according to the modes of activity of the exposed individual relative to the equipment during the exposure period. These are:
1) function or operation of equipment, 2) control or surveillance of equipment, and/or 3) maintenance of equipment.
c)
Background radiation dose due to adjacent potentially radioactive equipment.
Average exposure distances are determined for each of the exposure time modes and the resultant dose rate calculated for each distance using the average dose rates prevalent at that distance for that particular exposure mode. Typical exposure times for the various modes of exposures, i.e., operation, maintenance and repair, have been obtained from data on similar operations at other nuclear plants.
The following guidelines are applicable with respect to shielding requirements of pumps and valve galleries:
WSES-FSAR-UNIT-3 12.3-14 Revision 307 (07/13) a)
All pumps and valve galleries involved in the transmission of fluids (liquid and gases) of potential reactor coolant nuclide concentrations have been shielded.
b)
All pumps and valves involved in the transmission of secondary system fluids, excluding resins and concentrates, have generally not been shielded.
12.3.2.2 Design Description For shielding design purposes, the plant has been divided into radiation access zones, based on the maximum zone dose rate levels listed in Table 12.3-1.
These zone designations were used for initial shielding purposes only. Design criteria was based on regulatory dose limits applicable at the time. Specific regulations cited in zone descriptions are not applicable after January 1, 1994 due to changes to 10CFR20. Compliance with new dose limits effective January 1, 1994 will be achieved through appropriate administrative controls established in plant operating procedures.
(LBDCR 13-010, R307)
A description of each radiation zone chosen for design purposes is given below and shown on Figures 12.3-1 through 8a.
(LBDCR 13-010, R307) a)
Zone I (DRN 99-2362, R11; LBDCR 13-010, R307)
This zone has no restriction on occupancy. Such a zone would represent areas in the plant where radiation due to occupancy on a 40 hr/wk, 50 wk/yr basis, will not exceed the whole body dose of 500 mrem/yr, as specified in paragraph 20.105 of 10CFR20. Most non-employees and visitors to the site will receive considerably less than 500 mrem/yr because of the relatively short time interval during which they are onsite.
(DRN 99-2362, R11; LBDCR 13-010, R307) b)
Zone II This zone is a restricted radiation area which can be occupied by plant personnel and authorized visitors on a 40 hr/wk, 50 wk/yr basis without exceeding the allowable total effective dose equivalent of 5000 mrem/calendar year.
(DRN 99-2362, R11) c)
Zone III This is a restricted radiation area that plant personnel can occupy on a periodic basis. The average radiation level in this zone may vary from 2.5 to 15.0 mrem/hr.
d)
Zone IV This zone represents a restricted radiation area. The average radiation level may vary from 15.0 mrem/hr to 100 mrem/hr. Occupancy will be limited. However, qualified personnel who have been issued a Radiation Work Permit can enter these areas for brief periods of time to operate and inspect components. The length of stay in these areas will be determined by the Radiation Protection Staff based on the (DRN 99-2362, R11)
WSES-FSAR UNIT-3 12.3-15 Revision 11 (05/01) actual radiation level in the area, the past radiation history of the person entering, and the nature of the radiation.
e)
Zone V This zone represents areas with high potential for becoming high radiation areas. Areas exceeding 100 mrem/hr @ 30 cm2 from the source will be posted as CAUTION HIGH RADIATION AREA.
12.3.2.3 Methods of Shielding Design Shield wall thicknesses are determined by using basic shielding data and equations. Data is taken from the Table of Isotopes, "Reactor Physics Constants, ANL-5800," XDC-59-8-179, and other pertinent texts.
Radiation sources are determined as indicated in Section 12.2. The method of calculation normally employed is that of the point kernel integration, outlined in Reference 1 hereafter referred to as Rockwell's method. A computer code, ISOSHLD(2), has been used for some of the actual calculations. This program calculates the decay gamma ray and Bremstrahlung dose rate at the exterior of a shielded radiation source for a number of common geometric arrangements of sources and shields such as encountered in nuclear power plants. Source geometries that can be used include: point, linear, spherical, truncated conical, disc, cylindrical, and parallel piped sources. Slab shields are used for all cases.
Spherical shields can only be used for spherical sources. In addition, special computer codes such as SPAN-4(3) and MORSE-CG(4) have been employed.
The correct combination of source and shield is used to approximate the actual configuration in the plant.
Tanks and large pipes containing liquid are approximated by cylindrical sources. Gas filled tanks and pipes are simulated by line sources. Small liquid carrying pipes are also approximated by line sources.
Where the source shield dose point of geometry was sufficiently complex to preclude use of the Rockwell method or the ISOSHLD program, a point kernel integration, SPAN-4, was utilized. This program calculates the dose rate at a point from any number of sources having complex geometry and complex shield configurations. The geometry of the sources and shields are described by suitable intersection of quadratic surfaces. ISOSHLD, SPAN-4, and Rockwell's method account for scattering effects in the shields by using appropriate build-up factors.
Other computer codes have been utilized for updated calculations of dose rates and shielding at various points of interest. These additional computer codes utilize similar methods and geometries as the above mentioned codes. They have been verified and validated throughout industry and at W3SES to calculate both accurate and conservative radiation dose rate and shielding results.
None of the methods considers the energy degradation (softening) of the radiation spectrum as it emerges from the shield, and thus each predicts conservative values of the dose rate at the point of interest.
Whenever scattering effects were expected to be important or dominant such as in the calculation of the neutron and gamma ray streaming inside containment, the MORSE-CG, Monte Carlo code has been used. This code solves the neutron or gamma ray transport problem in arbitrary geometry by following a sufficient number of random "flight paths" of individual particles or rays through the physical system.
Importance sampling was used to reduce the number of "histories" which has to be followed, by arbitrarily terminating the "history" of certain rays in regions which are not considered important. Volumetric sources were simulated by a number of point sources.
Comparison of the measured dose rates at operating plants, both BWR and PWR type with corresponding theoretically calculated dose rates, indicate that the models and method of
WSES-FSAR UNIT-3 12.3-16 Revision 11 (05/01) calculation used predict higher dose rates than actually observed. Therefore, shielding calculations based on such models and methods are conservative.
To ensure that occupational exposures would be kept as low as reasonably achievable, the shielding design has been constantly reviewed, updated, and modified as necessary during all the phases of the plant design and construction.
12.3.2.4 Compliance with Regulatory Guide 1.69 Regulatory Guide 1.69, Concrete Radiation Shields for Nuclear Power Plants, December, 1973 generally invokes ANSI Standard N101.6-1972, Concrete Radiation Shields, as an acceptable method to the NRC for the design of concrete radiation shields for nuclear power plants. Waterford 3 complies with the intent of this guide with the following exceptions and clarifications:
ANSI N101.6-72 Section Exceptions and Clarifications 4.3.1 Concrete shielding at Waterford 3 pertains to gamma and/or neutron shielding only. There are no significant sources of alpha or beta radiation within the plant which could affect concrete shield design. The maximum temperature of the primary shield wall will be 150°F. This wall will be designed to afford required shielding at this temperature.
4.3.4 "The possibility of an explosion in the cell" is not applicable to Waterford 3, since there are no explosive materials contained within shielded compartments.
4.3.5 Assumptions and methods used for accident analyses are those given in Chapter 15. These assumptions and methods result in an acceptably conservative design.
4.3.6 Regulatory Guide 8.8 is used as guidance in limiting personnel exposure and determining shielding practices.
4.7 No design drawings will be prepared specifically for formwork. Concrete design drawings are provided in sufficient detail to allow proper design of formwork according to good construction practice. Formwork specifications are provided which require conformance to ACI-347-1968.
4.8 Not applicable to Waterford 3. No heavy aggregates are used.
5.1.2 Not applicable to Waterford 3. No high density concrete is used.
5.1.3 Not applicable to Waterford 3. No hydrous aggregate is used.
5.1.4 Not applicable to Waterford 3. No boron containing aggregates are used.
WSES-FSAR UNIT-3 12.3-17 Revision 11 (05/01) 5.1.6 Coatings of clay, silt, gypsum, calcite or caliche on coarse aggregate will total no more than three Percent of the total weight of the aggregate. Radiation attenuation calculations take this into account.
5.3.4 Not applicable to Waterford 3. No pozzolans are used.
5.3.5 Not applicable to Waterford 3. No grout fluidifiers are used.
5.4 For Waterford 3, a maximum slump of four in. is permitted for certain application: where less slump is impracticable.
5.4.2 Not applicable to Waterford 3. The preplaned-aggregate (PA) method is not used.
5.4.4 Not applicable to Waterford 3. Heavy aggregates are not used.
6.1 The use of noncombustible or fire retardant formwork for shielding; is impractical. Formwork for shielding is consistent with good construction practice and as required by ACI-347-1968.
6.2.1 ACI-347-1968 is used for the design of formwork.
6.2.2 Approval of concrete forms prior to construction is per ACI-347-1968.
ANSI N101.6-72 Section Contd.
Exceptions and Clarifications 6.4 No substitute for detailed thermal stress analysis is made.
6.5 See position for Section 4.7 with regard to shop drawings.
7.2 See position on Section 4.7 regarding shop drawings. Any changes in specifications must be reviewed and approved prior to construction activity. The effects of supplemental tracings on shield adequacy is evaluated at that time.
8.1.3 Not applicable to Waterford 3. No high-density concrete is used.
8.1.8 Aggregate is from one source and is continually sampled throughout the construction phase for conformance to project specifications. Considering these controls, bagging and retention of samples is not necessary.
8.2.6 Vibrators having a speed of 6000 cpm are used. This speed is adequate for producing satisfactory consolidation. Sufficient spare vibrators are maintained but not necessarily one for every two being used.
8.4 Not applicable to Waterford 3. The puddling method is not used.
8.6.1 The composition and fluidity of the mortar or grout when used in pressure grouting, is specified in project specifications.
8.6.2 Filling of forms is done in accordance with good construction practice.
Specifications require that no voids will be left in the concrete.
WSES-FSAR UNIT-3 12.3-18 Revision 11 (05/01) 8.7.1 The only construction joints shown on drawings are those essential to the design of the structure. Therefore, construction joints at other locations do not require approval of the engineer responsible for the design.
Construction joints are not stepped (i.e., not provided with offsets to prevent radiation streaming). Streaming between joints is not considered to be a problem since sufficient amplitude between joints is provided.
8.7.2 Concrete is cured for the specified times. The requirements of ACI-347-1968 are not met regarding time limits for removing forms.
8.7.5 Patching and finishing is performed, as soon as practicable to ensure a quality product; however, not necessarily within the specified times.
8.7.6 Traffic or other operations is restricted after curing and finishing to prevent damage to the concrete but not necessarily for the time specified.
9.1 Only certain areas subject to contamination by radioactive substances have a protective coating. ANSI N101.4-72 and ANSI N512-72 are utilized for such applications within Waterford 3.
10.1.2 Dimensional tolerances for hatches and openings as specified in ACI-347-1968 are used rather than those given in Table 1. Minimum practicable joint clearances are specified.
10-1.3 Not applicable to Waterford 3. Service trenches are not used.
10.2.2 The weight of each block is clearly marked on the block; however, not necessarily by stenciling.
10.2.3 Blocks are cured according to good construction practice but not necessarily in the absence of direct sunlight or heat. This sunlight or heat, however, does not result in the loss of shielding efficiency.
10.3.1 There are no present plans for penetrations through shielding plugs. However, if they are required, streaming is prevented by proper design of the penetration.
10.6 All precast shielding components are fabricated at the site. If, however, offsite fabrication is used, precast shielding components are not necessarily protected from direct sunlight or high temperatures during transit or storage. This exposure is not expected to result in loss or shielding efficiency.
11.5.1 Preoperational tests of shielding are not performed. Normal post-operational tests identify any areas necessary.
WSES-FSAR UNIT-3 12.3-19 Revision 14 (12/05)
(DRN 03-2066, R14) 11.5.2 Not applicable to Waterford 3. Containment leak testing is performed in accordance with 10CFR50 Appendix J.
(DRN 03-2066, R14) 12.3.3 VENTILATION The plant ventilation systems are designed to provide a suitable environment for personnel and equipment during normal plant operation and to provide a safe environment for operating personnel and the public during design basis accident conditions when controlling the plant to a safe shutdown conditio n.
12.3.3.1 Design Objectives
(DRN 03-2066, R14)
The plant ventilation system for normal plant operation and design basis accident conditions is designed to meet the requirements of 10CFR20, 10CFR50, and 10CFR50.67.
(DRN 03-2066, R14)
Design criteria for the plant ventilation systems include the following:
a)
During normal operation and design basis accident conditions, the maximum airborne radioactive material concentrations in air breathed by personnel in restricted areas of the plant must be as low as is reasonably achievable and within the limits specified in Appendix B of 10CFR20.
The maximum airborne radioactive material concentrations in unrestricted areas of the plant must be within the limits specified in Appendix B of 10CFR20.
b)
During normal operation and design basis accident conditions, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the site boundary are as low as is reasonably achievable and within the limits specified in 10CFR20 and 10CFR50.
(DRN 03-2066, R14) c)
The dose guidelines of 10CFR50.67 must be satisfied following postulated design basis accidents.
d)
The dose to main control room personnel shall not exceed the limits specified in General Design Criterion 19 of Appendix A to 10CFR50 and 10CFR50.67.
(DRN 03-2066, R14) e)
Airborne radioactivity monitoring is provided in compliance with General Design Criteria 63 and 64 of Appendix A to 10CFR50.
In the design of all ventilation systems, the following guidelines are used:
a)
The airflow is directed from areas of low potential airborne contamination to areas of higher potential airborne contamination.
b)
Airborne radiation monitoring is provided (see Subsection 12.3.4).
WSES-FSAR UNIT-3 12.3-20 Revision 11 (05/01) c)
Consideration is given to the disruption of normal airflow patterns by maintenance operations.
Ventilation systems are provided with back draft dampers and/or adjustable isolating dampers to allow servicing of redundant equipment without discontinuing system operation.
d)
Ventilation fans and filters are provided with adequate space around the units to allow servicing and replacement of sections.
e)
Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. In addition, the following concepts are used to minimize the spread of contamination:
1)
Equipment vents and drains are piped directly to the collection device connected to the collection system thus preventing spread of contamination.
2)
All welded piping systems and ductwork are employed on contaminated systems to the extent possible to reduce system leakage to a minimum acceptable level.
f)
Filters containing radioactivity can be easily maintained and will not create additional radiation hazard to personnel in normally occupied areas.
12.3.3.2 Design Description The air conditioning, heating, and ventilation systems are described for all plant buildings in Section 9.4 and Subsections 6.2.5 and 6.5.1. The aspects of the design that relate to removal of airborne radioactivity from equipment rooms, corridors and normally occupied areas are discussed under Subsection 11.3.2.
12.3.3.3 Air Cleaning System Design Air cleaning systems are either safety-related fission product removal systems which operate following a design basis accident or non-safety related systems which control airborne radioactivity in normally occupied areas during normal operation. The central exhaust system of the Reactor Auxiliary Building Normal Ventilation System is an illustrative example of a non-safety related air cleaning system which functions during normal operation (refer to Subsection 9.4.3).
An example layout of the RAB central exhaust system housing showing filter mountings, access doors, aisle space, service galleries and provision for testing, isolation and decontamination is provided on Figures 12.3-9 and 12.3-10.
Periodic testing for filters and adsorbers will be performed after initial operation. The frequency of changeout of adsorbers will be determined from periodic testing. The frequency of changeout of filters will be determined by monitoring the filters loading levels.
WSES-FSAR UNIT-3 12.3-21 Revision 11 (05/01) 12.3.3.4 Ventilation Systems Compliance to Regulatory Guides The ventilation systems with atmosphere cleanup features meet the intent of Regulatory Guide 1.52, Design, Testing and Maintenance Criteria for Engineered-Safety Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, June 1973. Waterford 3 compliance to Regulatory Guide 1.52 is described in Subsection 6.5.1.
12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION The radiation monitoring system provided for Waterford 3 consists of the following:
a)
Area Radiation Monitoring System, b)
Airborne Radiation Monitoring System, c)
Process and Effluent Radiological Monitoring and Sampling System, and d)
High Range Area Radiation Monitoring System.
The radiation monitoring systems for in-plant personnel radiation exposure determinations consist of (1)
Area Radiation Monitoring System, and (2) In-Plant Airborne Radiation Monitoring System which is part of the Airborne Radiation Monitoring System.
The Process and Effluent Radiological Monitoring and Sampling System is discussed in Section 11.5.
The radiation Monitoring system block diagram is shown on Figure 12.3-11.
The Area Radiation Monitoring System informs operations personnel, both locally and in the main control room, of radiation levels in areas where Area Radiation Monitoring System detectors are located, provides warning when abnormal radiation levels occur in specific plant areas, and warns of possible equipment malfunctions. Some channels of the Area Radiation Monitoring System are designed to Class 1E requirements and can withstand loss-of-coolant accident environmental conditions. Some of these Class 1E channels provide a containment purge isolation signal (refer to Subsection 7.6.1.5) in the event of abnormally high radiation inside the containment and enable main control room operators to monitor radioactivity levels inside the containment. The post-LOCA instrumentation provides information on the general direction of the accident. The High Range Area Monitors indicate radiation levels in areas requiring post-accident access or in areas with containment penetrations or hatches. In the event of a fuel handling accident, the Area Radiation Monitoring System provides a signal to isolate the Fuel Handling Building and start the emergency ventilation system.
The Airborne Radiation Monitoring System provides information, both locally and in the main control room, for the purpose of maintaining low in-plant personnel radiation exposure in accordance with 10CFR20 and Regulatory Guide 8.8 (March 1977). It provides information on the airborne activity levels inside the control room outside air intakes and in the event of detection of high airborne activity generates a signal to isolate the normal outside air intakes and start the emergency ventilation system.
The detectors assist operators in picking the one of two emergency intakes with the lowest airborne activity levels, thereby
WSES-FSAR UNIT-3 12.3-22 Revision 11 (05/01)
(DRN 99-2362) minimizing the amount of noble gases entering the control room environment and also minimizing the amount of emergency ventilation system filter loading. The In-Plant Airborne Radiation Monitoring System consists of three channel, movable, radiation monitors. The monitors consist of a particulate channel, iodine channel, and noble gas channel. The primary purpose of these monitors is: (a) To assure that the in-plant personnel are not over-exposed to airborne radionuclides during maintenance of routine operations and that no localized high activity concentrations exist in any of the plant areas, and (b) to assist personnel in deciding whether or not breathing apparatus is necessary prior to entering high or airborne radiation areas. Monitors comprising the in-plant Airborne Radiation Monitoring and the ranges are given in Table 12.3-3..
Regulatory Guide 8.10 (September 1975) is discussed in Section 12-1.
12.3.4.1 Area Radiation Monitoring System 12.3.4.1.1 Design Objectives The objectives of the Area Radiation Monitoring System during normal operating plant conditions and anticipated operational occurrences are:
a) to measure ambient gamma radiation and to indicate to operations personnel the ambient gamma radiation in specific areas of the plant, b) to annunciate and warn of abnormal radiation levels in specific areas of the plant, c) to furnish records of radiation levels in specific areas of the plant, d) to provide base data in controlling access of personnel to radiation areas, e) to warn of uncontrolled or inadvertent movement of radioactive material in the plant, f) to provide local indication and alarms at key points where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area,
(DRN 99-2362) g)
to assist operations and plant personnel in decisions on deployment of personnel in the event of an accident or equipment malfunction resulting in a release of radioactive material in the plant,
(DRN 99-2362) h)
to annunciate and warn of possible equipment malfunctions in specific areas of the plant, i) to furnish information for making radiation surveys, and j) to assist the supervisors in planning work schedules.
WSES-FSAR UNIT-3 12.3-23 Revision 11 (05/01)
The objectives of the Area Radiation Monitoring System during postulated accidents are:
(DRN 99-2362) a)
provide the capability to alarm and initiate a containment purge isolation signal in the unlikely event of a loss-of-coolant accident, fuel handling accident, or abnormally high radiation inside the containment, (Note, CIAS picks up LOCA indications before CPIS does)
(DRN 99-2362) b)
provide long term post-accident monitoring of conditions inside the containment, and c) provide a signal to isolate the Fuel Handling Building and start the emergency ventilation system in the event of a fuel handling accident.
12.3.4.1.2 Criteria for Location of Monitors Considerations for area monitor locations are based on the following:
a) frequency and length of personnel occupancy of a specific area, b) potential for personnel unknowingly to receive high radiation doses, c) potential for equipment malfunction, d) areas where during normal plant operation including refueling, radiation exposures could exceed the radiation limits due to system failure or personnel error, e) areas where new and spent fuel is received and stored, f) containment area for indicating the level of radioactivity and detecting the presence of fission products due to a reactor coolant pressure boundary leak, and
(DRN 99-2362) g)
normally or potentially radioactive release areas.
(DRN 99-2362) 12.3.4.1.3
System Description
The Area Radiation Monitoring channels are located at selected places inside the plant to detect and store information on the radiation levels and, if necessary, annunciate abnormal radiation conditions.
The areas where the gamma monitors are located are shown in Table 12.3-2. Indication, annunciation and storage is provided for all 39 channels in the main control room and other cathode ray tubes (CRTS) and computer magnetic storage tape. The instrument locations, ranges, sensitivities, accuracies and alarm set points are shown in Table 12.3-2. Environmental design conditions are discussed in Subsection 11.5.2.3.
A typical channel consists of a gamma sensitive Geiger-Muller (GM) or ion chamber detector, a microprocessor, power supply, a local indicator and audio-visual alarm and a check source or test current. The system utilizes local microprocessors with inputs to the radiation monitoring computers for purposes of data logging, processing, editing and displaying of information obtained from the radiation sensors. This microprocessor approach provides considerable flexibility in the means of collecting data and manners of displaying and utilizing such data.
WSES-FSAR UNIT-3 12.3-24 Revision 306 (05/12)
All channel information is processed through a dedicated microprocessor which is then interrogated by the radiation monitoring computer for processing, indications on CRTS, storage, alarming and hard copy production if so desired (EC-12329, R306)
Those channels identified on Table 12.3-2 as safety related are first indicated and recorded on digital ratemeters and recorders housed on the radiation monitoring panels in the main control room as shown on Figure 12.3-11. Through a properly qualified isolation buffer the signal is then transmitted to the Radiation Monitoring System computers for the kind of processing mentioned above. Upon a seismic event where the Radiation Monitoring System computers and peripherals are presumed to fail, the safety related channels maintain their function ability.
(EC-12329, R306)
The detectors are wall-mounted gamma sensitive Geiger-Muller tubes or ion chambers. Their energy dependence is typically flat (within +/- 20 percent) from 80 KeV to 1.5 MeV and each GM detector is provided with an integral check source, operated from the local microprocessor or the CRT in the main control room. The ion chamber area monitors have a test current to test circuit integrity.
Each monitor channel provides three alarms. One alarm level is set high enough above the normal measured radiation levels (background) in the area to prevent spurious alarms, yet low enough to indicate transient radiation level increases. A second alarm is set at a higher level. A third alarm acts as a tube failure, circuit failure or cable disconnect alarm. On occasion, other alarm points may be selected depending either upon work in progress in the area or operations that will vary the normal (background) measured radiation levels in the area. Some monitors have different alarm points when the reactor is critical than when the reactor is shut down. Plant chemistry personnel approve the alarm points consistent with radiological safety controls for the area.
(DRN 99-2362, R11,00-801 The instruments are calibrated and maintained on a routine schedule, as discussed in the Off Site Dose Calculation Manual, Technical Requirements Manual, the Technical Specifications, or plant procedures.
All alarms initiate continuous audible and visual alarms at the detector. The tone and volume of the local audible alarm varies in intensity to be easily heard in operating areas. The radiation monitoring computer through the PMC provides alarms for any channel detecting high radiation levels.
Identification of the channel alarmed is done at the CRT in the main control room and -4 Access Control point office.
(DRN 99-2362, R11,00-801)
Area radiation monitoring channels 24 through 31 have a 1,500 volt isolation buffer between the microprocessor and the radiation monitoring computer. Upon a postulated seismic event, all non-seismic channels, including the micro processor, and Radiation Monitoring Computers are presumed to fail. The buffer, designed in accordance with Reactor Research and Development Standards C163T-1971, Plant Protection System Buffers, and Sections 4.3.2 and 4.3.3 of C16-1T-1969, Supplementary Criteria and Requirements for RDT Reactor Plant Protection System, isolates the essential channels and maintains the signal flow to the ratemeters in the main control room. The main control room panel housing all safety-related ratemeters is qualified to seismic Category I requirements.
Area radiation monitoring channels 24 through 27 (ion chambers) (Containment Purge Isolation Detectors) are designed to Class 1E requirements and can withstand a LOCA environment for a period of at least 10 min. after the accident. Channels 24 through 27 are powered from two 120 V ac nuclear instrumentation buses SA and SB and are arranged in two groups with two monitors in each group. The output signals from these monitors make up
WSES-FSAR UNIT-3 12.3-25 Revision 11 (05/01) part of the containment purge isolation signal. Channels 28 through 31 are designed to Class 1E requirements. These are powered by two 120 V AC vital buses (SA and SB). All other monitors power supplies feed from an interruptible source and become inoperative during a loss of offsite power.
In addition to qualification to Class 1E requirements, channels 24 through 31 are physically and electrically separated from each other in accordance with the criteria set forth in IEEE 279-1971 and IEEE 308-1971.
A retractible solenoid operated radioactive check source provides a means of checking the integrity of GM tube detector area radiation monitors. A test current provides a means of checking the integrity of ion chamber detector area radiation monitors.
The microprocessor and computer receives, processes and displays information on request. Three alarms are provided: one for alert radiation, the second for high radiation and a third for when a channel becomes inoperative. The microprocessor has the ability to activate the check source into position or activate the test current for operational check purposes.
Four redundant fuel pool monitors are provided to detect radioactivity in the event of a fuel handling accident in the Fuel Handling Building. In the event of a refueling accident, four GM tubes positioned above the fuel pool area will sense the radioactivity released and will supply a signal for the startup of the Fuel Handling Building Ventilation System (only emergency portion, refer to Subsection 9.4.2) as well as the closure of isolation dampers in the normal ventilation system. These monitors are designed to Class IE requirements and in accordance with IEEE-279-1971, IEEE-308-1971, and IEEE-344-1971. The calibration performed is in accordance with vendor supplied transfer calibration procedures.
Recalibration will be performed at periodic intervals as set by plant procedures.
(DRN 99-2362) 12.3.4.2 Airborne Radiation Monitoring System 12.3.4.2.1 Design Objectives The objectives of the Airborne Radiation Monitoring System during normal operating plant conditions and anticipated operational occurrences are:
a) to inform operations personnel of airborne particulate, gaseous and iodine (-, gross, and,
respectively) activity trends in the various buildings and structures of the plant,
(DRN 99-2362) b)
to alarm in case of abnormal increase in the airborne activity levels,
(DRN 99-2362) c)
to furnish records of gross airborne trends in the various plant areas and of the amount of radioactive releases to the environment through the plant buildings or structures during normal, or abnormal operational occurrences, d) to help detect identified or unidentified leaks inside the reactor coolant pressure boundary (as recommended in Regulatory Guide 1.45, May 1973) and other areas of the plant,
WSES-FSAR UNIT-3 12.3-26 Revision 11 (05/01)
(DRN 99-2362) e)
to assist personnel in deciding whether or not breathing apparatus is necessary when entering a high airborne radiation area, and
(DRN 99-2362) f)
to provide information for evaluation of the performance of all plant systems that function to minimize the release of radioactivity to accessible areas of the plant and to the environment.
The objective of the main control room airborne radiation monitoring system during postulated accidents is to provide the capability to alarm and initiate isolation of the main control room normal ventilation system and actuate the emergency ventilation system in the unlikely event that radioactivity is introduced into the main control room intake ductwork.
12.3.4.2.2 Criteria for Location of Monitors Considerations for locating the Airborne Radiation Monitoring System monitors are based on the following:
a) paths that normally, or potentially, may release airborne radioactivity to the environment, b) areas where the airborne radioactivity can abruptly increase and where personnel normally have access to the areas, c) in ventilation ducts where the monitors can survey, at large, the airborne radioactivity level, and d) inside the containment for the purpose of monitoring unidentified leaks.
12.3.4.2.3
System Description
Airborne radioactivity detection devices are provided in the plant to monitor normal radiation levels and to detect and annunciate any abnormal radiation conditions. The Airborne Radiation Monitoring System consists of monitors for the containment, the main control room, as well as certain areas in the Reactor Auxiliary Building (RAB). Table 12.3-3 lists the Airborne Radiation Monitoring System monitors including the locations, number of channels, type, range, sensitivity, accuracy, and alarm set points.
Environmental design conditions are discussed in Subsection 11.5.2.3.
The RAB is an area which has the possibility of containing airborne contamination when personnel are present. Calculations were performed to determine the capability of the airborne radioactivity monitors in the RAB to detect 1 DAC in air of particulate or iodine radioactivity in one hour in any compartment sampled by these monitors. Based on the calculations, monitor PRM-IR-6710D cannot satisfy the subject detection requirement for those compartments with flowrates less than 198 cfm for iodine and 88 cfm for particulates. In those compartments, during extended maintenance where the 1 DAC in one hour requirement cannot be met, administrative procedures will provide guidance.
WSES-FSAR UNIT-3 12.3-27 Revision 15 (03/07)
The guidance of regulatory position paragraph C.2.g of Regulatory Guide 8.8 (March 1977) has been factored into the design of the Airborne Radiation Monitoring System.
This system also utilizes the microprocessor approach as described in Subsection 12.3.4.1.3.
Additional alarms are provided for malfunction of the particulate and iodine filters and high pressure and low pressure across the filters.
12.3.4.2.3.1 Containment Atmosphere Radiation Monitor The containment atmosphere radiation monitor is designed to provide a indication in the main control room of the particulate, iodine and gaseous radioactivity levels inside the containment. Radioactivity in the containment atmosphere indicates the presence of fission products due to a reactor coolant pressure boundary leak.
The containment atmosphere sample is drawn into the monitoring assembly by a one in. stainless steel sampling line as shown in Figure 12.3-15. The flow sample is distributed to the iodine and the particulate samplers by a V pipe connection. The iodine sample is obtained when the sample passes through a fixed charcoal filter bed. The fixed charcoal filter bed is then monitored by a gamma sensitive scintillation detector. The particulate sampler collects particles greater than or equal to 0.3 microns on a moving paper filter. A beta sensitive scintillation detector aimed at the moving paper filter monitors for particulate radiation. After passing through particulate and iodine filters, the sample will be dried in a moisture control unit and then monitored for radioactive gas content in the gas sampler. The sample lines are kept as short as possible and horizontal runs are minimized to reduce plate-out and losses due to gravity deposition prior to iodine and particulate monitoring. The moisture control unit can be isolated from the radiation monitor via isolation valves installed between the two units.
(DRN 06-1029, R15)
The pumping system consists of two sample pumps; one pump draws air through the iodine and gas monitors and the other draws air through the particulate monitor. Flowmeters are placed downstream of each pump, to indicate flow rates and to alarm abnormal flow rates. Mass flow probes control the flow of the monitor pathways via the monitor microprocessor. The mass flow probes also signal alarms in the microprocessor when a sudden change in flow takes place indicating abnormal filter function in the particulate and iodine monitors. The isolation of the noble gas monitor allows remote purging with clean air for background checkup and maintenance. The air stream bypasses the noble gas monitor whenever it is isolated by its isolation valves. The purpose of having the air stream bypass the noble gas monitor is to allow operation of the iodine and particulate monitors when the noble gas monitor is undergoing testing or maintenance, thus achieving additional system flexibility.
(DRN 06-1029, R15)
A containment isolation actuation signal will isolate this monitor from the containment.
The iodine filter is periodically replaced and can be analyzed in the laboratory. Adequate shielding is arranged in a 4 geometry around the detectors to prevent interference from background radiation and electromagnetic fields.
A solenoid operated check source is provided for verifying detector operation.
WSES-FSAR UNIT-3 12.3-28 Revision 11 (05/01)
Each channel provides three alarm modes; FAIL, ALERT and HIGH. ALERT and HIGH alarms are adjustable over the full span of the scale. One alarm is set to alarm for detector signal failure, power failure, or from a failure due to a disconnected cable. The second is set to alert that a specified radiation level has been exceeded. The third is set at a higher level to alarm at higher radiation levels.
The containment atmosphere radiation monitor is designed to Class 1E requirements and in accordance with IEEE-279-1971, IEEE-308-1971 and IEEE-344-1971.
Indication devices are located on the main control room radiation monitoring panel. All radiation channels are indicated, stored, as required, and annunciated on the main control room radiation monitoring panel. Abnormal radiation levels are indicated both visually and audibly, locally and in the main control room. Each detector is calibrated at the factory using two or more National Institute Standards Technology calibrated isotopes prepared in the proper form to simulate the effluent for which the system will be used.
A recalibration of the detectors will be performed at periodic intervals as set by the Offsite Dose Calculation Manual, Technical Requirements Manual, the Technical Specifications, or plant procedures.
Calibration is performed in accordance with vendor supplied transfer calibration procedures.
As described in Subsection 11.5.2.1, the following items are applicable to the containment atmosphere radiation monitor: monitor cabinet, sampler, check source, detector assembly, recorders and power supplies.
12.3.4.2.3.2 Plant Stack Radiation Monitor The plant stack radiation monitor is designed to representatively sample, monitor, indicate and store the radioactivity levels in the plant effluent gases being discharged from the plant stack. It provides a continuous indication in the main control room of the activity levels of radioactive materials released to the environs so that determination of the total amount of activity release is possible. These monitors have been designed to the applicable requirements of Regulatory Guide 1.21 (June 1974), regulatory position paragraphs C.2 through C.11.
A schematic diagram of the plant stack radiation monitor is shown on Figure 12.3-13.
The plant stack radiation monitor monitors the plant stack for particulates, iodine and noble gases at the point of release to the atmosphere. Its function is to confirm that releases of radioactivity do not exceed the predetermined limits set by the Offsite Dose Calculation Manual (ODCM) or Technical Requirement Manual (TRM).
The sample flow is withdrawn from the stack through an isokinetic nozzle located at a minimum of eight stack diameters from the last point of radioactivity entry. The nozzles are designed such that the sampling velocity is the same as that in the stack pipe so that preferential selection does not occur, i.e.,
so that the weights of the radioactive particles do not become a factor in obtaining a representative sample. The isokinetic sampling system is designed in accordance with ANSI N13.1-1969.
The particulate iodine and gaseous detectors used for each plant stack monitor have the same technical description as those for the containment atmosphere radiation monitor described above (with the exception of a moisture control unit). The calibration performed
WSES-FSAR UNIT-3 12.3-29 Revision 12 (10/02) is similar to that for the containment atmosphere radiation monitor, also discussed in that subsection. As described in Subsection 11.5.2.1, the following items are applicable to the containment atmosphere radiation monitor; monitor cabinet, sampler, check source, detectors assembly, recorders and power supplies.
12.3.4.2.4 Main Control Room Radiation Monitors Two redundant pairs of radiation monitors, as shown on Figure 12.3-14, are provided for monitoring radioactive airborne concentration levels inside the two outside air intake plenums. Each of the four monitors consists of two separate plastic scintillator detectors; one detector with a beta shield and one without. The detector with a beta shield is not used. It is an installed spare for the beta-gamma detector.
The detector with the beta shield does not have any alarm functions. All alarms are applicable for the beta-gamma detector only. Using the beta-gamma detector operators are able to pick the air intake with the lowest airborne concentration levels.
Under normal operation the control room ventilation system draws air from the north side intake plenums.
Upon detection of high radiation in any one of the radiation detectors, a signal is developed to isolate the control room. The operator shall examine the radiation monitor outputs to determine which intake point has the lowest radiation level and choose the intake point with the lowest airborne contamination level to pressurize the control room.
These monitors are designed to Class 1E requirements and in accordance with IEEE 279-1971, IEEE 308-1971, and IEEE 344-1971.
The calibration performed is in accordance with vendor supplied transfer calibration procedures.
12.3.4.2.5 Other Airborne Radiation Monitors
¨(DRN 02-407)
Other airborne radiation monitors exist in the plant as shown in Table 12.3-3. Of those listed, the hot machine shop and the decontamination facility monitors are movable carts.
(DRN 02-407) 12.3.4.3 High Range Area Radiation Monitoring System 12.3.4.3.1 Design Objectives The objectives of the High Range Area Radiation Monitoring System during and following postulated accidents are:
a)
To indicate radiation levels in area with containment penetrations or hatches to provide indication of breach, and b)
To indicate radiation exposure rate in areas where access is required to service equipment post-accident.
WSES-FSAR UNIT-3 12.3-30 Revision 11 (05/01) 12.3.4.3.2
System Description
The High Range Area Radiation Monitoring channels are located at selected places inside the plant to detect and store information on the radiation levels and, if necessary, annunciate abnormal radiation conditions. The area where the gamma monitors are located are shown in Table 12.3-2.
Indication, annunciation and storage is provided for all 11 channels in the main control room and other cathode ray tubes (CRTS) and computer magnetic storage tape. The instrument locations, ranges, sensitivities, accuracies and alarm setpoints are shown in Table 12.3-2.
For the High Range Area Radiation Monitoring System, a typical channel consists of a gamma sensitive ion chamber detector, a microprocessor, power supply, a local indicator and audio-visual alarm and a test current. The system utilizes local microprocessors with inputs to the radiation monitoring computers for purposes of data logging, processing, editing and displaying of information obtained from the radiation sensors. This microprocessor approach provides considerable flexibility in the means of collecting data and manners of displaying and utilizing such data.
All channel information is processed through a dedicated microprocessor which is then interrogated by the radiation monitoring computer for processing, indications on CRTS, storage, alarming and hard copy production if so desired.
The detectors are wall-mounted gamma sensitive ion chambers. Their energy dependence is typically flat (within +/- 20 percent) from 60 keV to 3 MeV and each monitor is provided with a test current, operated from the local microprocessor or the CRT in the main control room.
Each monitor channel is provided with three alarms. One alarm level is set high enough above the normal measured radiation levels in the area to prevent spurious alarms, yet low enough to indicate transient radiation level increases. A second alarm is set at a higher level. A third alarm acts as a tube failure, circuit failure or cable disconnect alarm. On occasion, other alarm points may be selected depending upon work in progress in the area or operations that will vary the normal measured radiation levels in the area. Some monitors are expected to have different alarm points when the reactor is critical then they will have when the reactor is shut down. Plant chemistry personnel will specify the alarm points consistent with radiological safety controls for the area.
The instruments are calibrated and maintained on a routine schedule. All alarms initiate continuous audible and visual alarms at the detector. The tone and volume of the local audible alarm is of variable intensity to be easily heard in operating areas. The radiation monitoring computer through the PMC provides alarms for any channel detecting high radiation levels. Identification of the channel alarmed is done at the CRT in the main control room and the -4 Access Control Point Office.
A means of checking the integrity of the High Range Area Radiation Monitoring System is accomplished through the use of a test current.
The microprocessor and computer receives, processes and displays information on request. Three alarms are provided: one for high radiation, the second for high-high radiation and a
WSES-FSAR UNIT-3 12.3-31 Revision 11 (05/01) third for when a channel becomes inoperative. The microprocessor has the ability to activate the test current for circuit verification purposes.
SECTION 12.3: REFERENCES 1.
Rockwell T., "Reactor Shielding Design Manual," USAEC Report 7004, 1956.
2.
ISOSHLD "Kernel Integration Code, General Purpose Isotope Shielding Analysis." CCC-79.
Oak Ridge RSIC, 1973.
3.
SPAN-4, "A Point Kernel Computer Program for Shielding," WAPD-TM-809(L), O.J. Wallace, October 1973.
4.
MORSE-G, "General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code with Combinational Geometry," CCC-203, E.A. Strakes, Oak Ridge, 1970.
5.
"DOT II W Users Manual" WANL-TME-1982, December 1969.
6.
E.A. Straker et al, "MORSE-CG General Purpose Monte Carlo Multigroup Neutron and Gamma Ray Transport Code with Combinational Geometry," CCC-203, ORNL-RSIC, 1973.
7.
DLC-23, "CASK 40 Group Coupled Neutron and Gamma Ray Cross Section Data," ORNL-RSIC, 1974.
8.
M.B. Emmett et al, "DOMINO: A General Purpose Code for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations," ORNL-4853, 1973.
9.
Baltimore Gas & Electric Co. "Calvert Cliffs Unit No. 1 FSAR Amendment No. 51, Docket No.
50-317, November 1975."
10.
Florida Power & Light. "St. Lucie Unit 1 Docket No. 50-355 License Condition D, Neutron Streaming Shield," April 1977.
11.
Northeast Nuclear Energy Company "Radiation Survey Results In and Around Millstone Unit 2 Containment Building," Docket No. 50-336. April 1976.
12.
O.J. Wallace. SPAN-4. "A Point Kernel Computer Program for Shielding." WAPD-TM-809(L),
Bettis Atomic Power Laboratory, October 1972.
(DRN 99-1098) 13.
Waterford 3 Rerack Radiological and Shielding Calculations, Holtec Report HI-971735, Calculation EC-M98-021.
14.
QAD-CGGP, A Combinatorial Geometry Point Kernel Code, Report CCC-493, Radiation Shielding Information Center, July 16, 1980.
15.
SAS2H-ORIGEN-S, in Scale 4.3-Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG-CR-0200, Rev. 5, September 1995 (Issued by Radiation Shielding Information Center).
16.
American National Standard ANSI/ANS-6.1.1-1991, neutron and gamma-ray fluence-to-dose factors, American Nuclear Society, 1992.
WSES-FSAR-UNIT-3 TABLE 12.3-1 Revision 308 (11/14)
ALLOWABLE DOSE RATES Location (Area)
Dose Rate mrem/hr (max)
Site Boundary 0.05 (DRN 99-2362, R11)
(DRN 99-2362, R11)
Outside Tool Room 0.05 Maintenance Support Building 0.05 Service Building 0.05 (DRN 99-2362, R11)
Main Control Room, Turbine Building, 0.25 Westside Access Enclosure and Reactor Auxiliary Building areas outside controlled access areas (DRN 99-2362, R11)
Reactor Auxiliary Building and Fuel Handling 2.5 Building areas inside controlled access areas:
areas of 40 hr/wk occupancy Reactor Auxiliary Building, Fuel Handling 15.0 Building and containment: areas of six hr/wk occupancy Reactor Auxiliary Building, Fuel Handling Building and containment:
areas where occupancy is determined by 100 health physics staff High radiation areas
>100 (DRN 03-2066, R14)
Main control room following maximum 5 rem TEDE hypothetical accident over 30 days following the accident (DRN 03-2066, R14)
(LBDCR 13-009, R307)
Low Level Radwaste Storage Facility 0.6 mrem/hr (LLRWSF) Restricted Area Boundary (LBDCR 13-009, R307)
(LBDCR 13-010, R307)
Original Steam Generators Storage Facility 0.05 At the fence boundary (LBDCR 13-010, R307)
(LBDCR 14-007, R308)
Entrapment Area 0.05 (LBDCR 14-007, R308)
WSES-FSAR-UNIT-3 TABLE 12.3-2 (Sheet 1 of 4)
Revision 11-A (02/02)
AREA RADIATION MONITORS Channel Location(3)
Safety Classification Range (mR/hr)
Sensitivity (cpm/mR/hr)
Accuracy
(%)
Maximum Alarm Setpoint Local Alarm
Background
Radiation (mR/hr)
Reactor Auxiliary Bldg. (RAB)
ARM-IR-5001(4)
RE-1 Main Control Room Non-safety 0.01-1000 104
+/- 20(1) 1.0 Yes 0.25 ARM-IR-5002 RE-2 EL. +46 ft. MSL, Column K-5A Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5003 RE-3 EL. +21 ft. MSL, Column H-4A Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5004 RE-4 EL. -4 ft. MSL, Column K-2A Non-safety 0.01-1000 104
+/- 20 7.5 Yes 2.5 ARM-IR-5005 RE-5 EL. -4 ft. MSL, Wall H bet. Column 4A and 5A Non-safety 0.01-1000 104
+/- 20 30.0 Yes 15 ARM-IR-5006 RE-6 Counting Room, EL. -4 ft. MSL Non-safety 0.01-1000 104
+/- 20 1.0 Yes 0.25 ARM-IR-5007 RE-7 Sampling Room, EL. -4 ft. MSL Non-safety 0.01-1000 104
+/- 20 1.0 Yes 0.25 ARM-IR-5008 RE-8 Boric Acid Precon, filters -
EL. -35 ft. MSL (Column H-3A)
Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5009 RE-9 Waste Filters Wall, EL. -35 ft. MSL Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5
¨ (DRN 99-2362; 00-1046)
ARM-IR-5016 RE-16 Hot Tool Room, Column H-2A Non-safety 0.01-1000 36.7
+/- 20 5.0 Yes 2.5 (DRN 99-2362; 00-1046)
ARM-IR-5017A RE-17 Charging Pumps Wall, EL. -35 ft. MSL Non-safety 0.1-104 36.7
+/- 20 30.0 Yes 15 ARM-IR-5019 RE-19 Elevator Shaft, EL. +21 ft. MSL Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5020 RE-20 Radio Chemistry Lab bet. Columns 10A and 11A, EL. -4 ft. MSL Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5021 RE-21 Corridor near Valve Gallery (Column 4A) EL. -4 ft. MSL Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5 ARM-IR-5022 RE-22 Corridor near Filter flush Tank (Wall 6A-L) EL. -4 ft. MSL Non-safety 0.01-1000 104
+/- 20 5.0 Yes 2.5
WSES-FSAR-UNIT-3 TABLE 12.3 2 (Sheet 2 of 4) Revision 7 (10/94)
AREA RADIATION MONITORS Channel Location(3)
Safety Classification Range (mR/hr)
Sensitivity (cpm/mR/hr)
Accuracy
(%)
Maximum Alarm Setpoint Local Alarm
Background
Radiation (mR/hr)
ARM IR 5022B RE 22B EL. 35ft. MSL, South of Column H 4A Non safety 0.01 1000 104
+/- 20 5.0 Yes 2.5 ARM IR 5022C RE 22C EL. 4 ft. MSL, West of Column H 3A Non safety 0.01 1000 104
+/- 20 30.0 Yes 15 ARM IR 5023 RE 23 Corridor near Gas Decay Tanks (Column K 4A) EL. 35 ft. MSL Non safety 0.01 1000 104
+/- 20 30.0 Yes 15 ARM IR 5023A RE 23A Decontamination Room EL. +21 ft.
MSL, near Column 3A H Non safety 0.01 1000 104
+/- 20 5.0 Yes 2.5
ARM IR 5200 Recirc. Penetration Area EL. 4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5201 Personnel Air Lock Area #1; EL. 4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 2.5 ARM IR 5202 SIS Sump Penetration Area; EL.
4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5204 Post Accident Sampling System Area; EL. 4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5205 DG Room 3A S EL. +21 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5206 DG Room 3B S EL. +21 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5207 Electrical Equipment Area EL.
+21 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5208 Component Cooling Water Area EL. 4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 0.25 ARM IR 5209 Waste & Boron Control Panel Area EL 4 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 30.0 ARM IR 5210 Turbine Driven EFW Pump Area; EL. 35 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 1x103 Yes 30.0
WSES-FSAR-UNIT-3 TABLE 12.3 2 (Sheet 3 of 4) Revision 10 (10/99)
AREA RADIATION MONITORS Channel Location(3)
Safety Classification Range (mR/hr)
Sensitivity (cpm/mR/hr)
Accuracy
(%)
Maximum Alarm Setpoint Local Alarm
Background
Radiation (mR/hr)
Reactor Bldg. (RB)
- ARM IR 5013 RE 13 Refueling Machine (during shutdown only)
Non safety 0.1 104 36.7
+/- 20 2 x Bkg Yes 15,000n:1,000 (power operation),
2.5 (shutdown)
- ARM IR 5014 RE 14 Southwest Staircase Non safety 0.1 104 36.7
+/- 20 2 x Bkg Yes 1,000n:1,000
- ARM IR 5015 (4)
RE 15 Northeast Staircase Non safety 0.1 104 36.7
+/- 20 2 x Bkg Yes 1,000n:1,000
- ARM IR 5018 RE 18 EL. 4 ft. MSL, Personnel Lock Non safety 0.1 104 36.7
+/- 20 2 x Bkg Yes(2) 15n:10 ARM IR 5024S RE 24 North side of SG Shield Wall, EL. +46 ft. MSL (containment purge isolation) 1E 20 5x105 7x10 10 A/R/hr
+/- 20 2 x Bkg Yes 2,000n:10,000
ARM IR 5025S RE 25 South side of SG Shield Wall, EL. +46 ft. MSL (containment purge isolation) 1E 20 5x105 7x10 10 A/R/hr
+/- 20 2 x Bkg Yes 2,000n:10,000
ARM IR 5026S RE 26 Southwest side of SG Shield Wall, EL. +21 ft. MSL (containment purge isolation) 1E 20 5x105 7x10 10 A/R/hr
+/- 20 2 x Bkg Yes 300n:2,000 ARM IR 5027S RE 27 North side of SG Shield Wall, EL. +21 ft. MSL 1E 20 5x105 7x10 10 A/R/hr
+/- 20 2 x Bkg Yes 300n:2,000 ARM IR 5028S RE 28 Outside RB Shield Wall (105°F),
EL. +46 ft. MSL (Post LOCA) 1E 101 105 36.7
+/- 20 None Yes 0.25 ARM IR 5029S RE 29 Outside RB Shield Wall (230°F),
EL. +21 ft. MSL (Post LOCA) 1E 101 105 36.7
+/- 20 None Yes 2.5 ARM IR 5030S RE 30 Outside RB Shield Wall (115°F),
EL. +21 ft. MSL (Post LOCA) 1E 101 105 36.7
+/- 20 None Yes 0.25
- Used by HP during shutdown for personnel protection. Not required during normal operation.
WSES-FSAR-UNIT-3 TABLE 12.3 2 (Sheet 4 of 4) Revision 10 (10/99)
AREA RADIATION MONITORS Channel Location(3)
Safety Classification Range (mR/hr)
Sensitivity (cpm/mR/hr)
Accuracy
(%)
Maximum Alarm Setpoint Local Alarm
Background
Radiation (mR/hr)
ARM IR 5031S RE 31 Outside RB Shield Wall (240°F),
EL. +46 ft. MSL (Post LOCA) 1E 101 105 36.7
+/- 20 None Yes 0.25
Fuel Handling Bldg. (FHB)
ARM IR 5010 RE 10 EL. +50 ft. MSL, Column U 7FH Non safety 0.01 1000 104
+/- 20 20 Yes 2.5 ARM IR 5011 RE 11 EL. +46 ft. MSL, New Fuel Vault Non safety 0.01 1000 104
+/- 20 20 Yes 2.5 ARM IR 5012 RE 12 Fuel Pool Pumps Hall Wall near Equipment Hatch Non safety 0.01 1000 104
+/- 20 30 Yes 15 ARM IR 0300.1S RE 32 FHB Adjacent to Spent Fuel Pool 1E 0.1 104 36.7
+/- 20 103 Yes 2.5 ARM IR 0300.2S RE 33 FHB Adjacent to Spent Fuel Pool 1E 0.1 104 36.7
+/- 20 103 Yes 2.5 ARM IR 0300.3S RE 34 FHB Adjacent to Spent Fuel Pool 1E 0.1 104 36.7
+/- 20 103 Yes 2.5 ARM IR 0300.4S RE 35 FHB Adjacent to Spent Fuel Pool 1E 0.1 104 36.7
+/- 20 103 Yes 2.5 ARM IR 5203 Refueling Canal Area, EL. +46 ft.
Non safety 102 107 1.2x10 10 A/R/hr
+/- 20 103 Yes 2.5 (1) Of indicated field intensity due to a combined action of +/- 20% voltage variation and environmental design conditions as discussed in Subsection 11.5.2.3.
(2) Outside containment on platform.
(3) Elevations are floor levels.
(4) LP&L tag numbers.
WSES FSAR UNIT 3 TABLE 12.3 3 Revision 14 (12/05)
AIRBORNE RADIATION MONITORS Numbers Location Type Number of Channels Range
(ci/cc)
Sensitivity
(ci/cc)
Ratemeter Accuracy Maximum Alarm Setpoint
(ci/cc)
Background
Radiation (mR/hr)
(DRN 05-455, R14)
- 1.
Containment Atmosphere (PRM IR 0100Y)***
Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75x10 10 1.85x10 7 (Cs 137)
(Xe 133)
20% of scale 3.01E 10 3.39E 8 1.33E 2 2.5
(DRN 05-455, R14)
- 2.
Main Control Room, at outside air intakes (PRM IR 0200.1, 0200.2, 0200.5, 0200.6
Scintillator 4
10 8 10 2 3.56X10 4 (Kr 85)
20% of scale 5.45 E 6 0.25
- 3.
Hot Machine Shop (PRM IR 5132)***
Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75X10 10 1.85X10 7 (Cs 137)
20% of scale 6E 8 2E 8 1E 4 2.5
- 4.
Decontamination Facility (PRM IR 5144)***
Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75X10 10 1.85X10 7 (Cs 137)
(Xe 133)
20% of scale 6E 8 2E 8 1E 4 2.5
- 5.
RAB Monitor A (PRM IR 6710A)***
located in ductwork on EL. 4 ft. MSL at Column J 4A Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75X10 10 1.85X10 7 (Cs 137)
(Xe 133)
20% of scale 6E 8 2E 8 1E 4 2.5
- 6.
RAB Monitor B (PRM IR 6710B)***
located in ductwork on EL. +21 ft. MSL west of Column M 5A Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75X10 10 1.85X10 7 (Cs 137)
(Xe 133)
20% of scale 6E 8 2E 8 1E 4 2.5
- 7.
RAB Monitor C PRM IR 6710C)***
located in ductwork on EL. 4 ft.
MSL column J 6A Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.56X10 12 1.75X10 10 1.85X10 7 (Cs 137)
(Xe 133)
20% of scale 6E 8 2E 8 1E 4 2.5
- 8.
RAB Monitor D (PRM IR 6710D)***
located in ductwork on EL. +46 ft. MSL Column K 4A Particulate Iodine Gaseous 3
10 11 10 5 10 9 10 3 10 7 10 1 1.546X10 12 1.75X10 10 1.85X10 7 (Cs 137)
(Xe 133)
20% of scale 6E 8 2E 8 1E 4 2.5
(DRN 99-2362, R11;02-407, R12)
(DRN 99-2362, R11;02-407, R12)
Setpoints determined in accordance with Offsite Dose Calculation Manual.
W 3 tag number.
WSES-FSAR-UNIT-3 TABLE 12.3-4 NEUTRON STREAMING DOSE RATES IN CONTAINMENT Containment Location Neutron Dose Rate (Rem/hr) 1.
Edge of Refueling Cavity at 16.5 Refueling Machine Normal Parking Location-Operating Floor 2.
Extreme Edge of Refueling Cavity 12.0 Near Steel Containment-Operating Floor 3.
Edge of Refueling Cavity Near 22 Steam Generator Shield Wall-Operating Floor 4.
Corner of Refueling Cavity Below 25.6 Missile Shield-Operating Floor 5.
Near Access Lock-Mezzanine Floor 0.25 6.
+21 EL General Mezzanine Floor Area 0.025-0.3 7.
-4 EL General Lower Floor Area 0.005-0.06
WSES-FSAR-UNIT-3 Revision 11 (05/01)
Figure 12.3-13 has been Intentionally Deleted.
WSES-FSAR-UNIT-3 12.3A-1 Revision 301 (09/07) 12.3A TMI SHIELDING STUDY 12.3A.1 INTRODUCTION In July 1979, NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report) was published. One requirement of NUREG-0578 was a design review of plant shielding of areas necessary for plant operation. That review, as clarified by NUREG-0737, is presented here.
The design basis of this review includes the effects of large and small break LOCAs. The dose rates and integrated doses were calculated by standard shielding techniques, utilizing the actual geometries and expected source strengths as a function of time and space. All equipment, piping, and areas which could contain radioactivity under accident conditions were considered (systems considered are listed in Table 12.3A-8).
<<(DRN 05-1249, R14)
Dose rate calculations were performed for those areas containing radioactive sources and the surrounding areas affected by these sources to ensure that vital areas (those necessary for post-accident operations) are accessible and habitable for the time necessary to perform the required tasks in that area.
This was done by placing piping diagram transparencies over general arrangement drawings to determine the source term for each area. The original design calculations were done by 3 dimensional point kernel technique which is contained in the computer program SPAN-4(1). This program has rectangular, cylindrical and spherical geometries available-thus, all the source and shielding shapes were accurately described in this analysis. The SPAN-4 program contains libraries of material densities and cross section, energy structures, buildup factors, flux to dose conversion factors and quadrature weights. The library data can be changed or added to (on a temporary basis) by the user through input data on cards. This calculational model is used for the direct dose rates from the contained sources.
Other computer codes have been utilized for updated calculations of dose rates at various points of interest. These codes have been verified and validated throughout industry and by WSES to calculate both accurate and conservative radiation dose rate and shielding results.
ø(DRN 05-1249, R14)
<<(DRN 99-2362, R11; 03-2066, R14; 05-1249, R14; EC-5000082374, R301)
The original TMI design basis dose analyses were reviewed for impact based on EPU conditions. A comparison of the average or expected radioisotope activities based on ANSI N237 (FSAR design basis) and ANSI 18.1 (EPU basis) and an evaluation of the change in flux-to-dose conversion factors between ANSI 6.1.1 1977 and ANSI 6.1.1 1991 indicate that normal operation and anticipated occurrences dose rates due to EPU are bounded by the current FSAR design basis dose rates.
ø(DRN 99-2362, R11; 03-2066, R14; 05-1249, R14; EC-5000082374, R301)
The dose rates calculated for this analysis used the worst conditions as their basis; thus, for practical operating conditions the Health Physics personnel will need to take dose rate measurements to guide personnel in their trips about the plant. Some areas may be found accessible which are indicated as inaccessible in this Appendix.
12.3A.2 SOURCE TERMS
<<(DRN 99-2362, R11;05-144, R14)
The source terms used for areas other than the control room and the diesel generator rooms are consistent with the requirements of NUREG-0737 and Regulatory Guide 1.4. The core inventories were taken from Table 4.3-1 of the Combustion Engineering document SYS80-PE-RG, Revision 4, Radiation Design Guide issued July 12, 1979 and are listed in Table 12.3A-1. The source data is tabulated in Table 12.3A-2 through 12.3A-7.
ø(DRN 99-2362, R11)
The source term used for the main control room and the emergency diesel generator rooms are consistent with the requirements of Regulatory Guide 1.183 and are listed in Table 12.2-12.
ø(DRN 05-144, R14)
WSES-FSAR-UNIT-3 12.3A-2 Revision 14 (12/05)
The source strengths used in this analysis were divided into three categories; liquid, gaseous and plateout. The source strengths (/cc-sec) are calculated for nine energy groups, and 11 times following the accident.
(DRN 99-2362, R11; 03-2066, R14)
For the liquid source, two conditions were studied to ensure that calculated dose rates are conservative.
The first condition is one in which a large break LOCA occurs in the primary loop, and during the injection phase a large quantity of the injection water spills from the primary loop to the floor of the containment and the SIS sump. The injection water comes from four safety injection tanks, the boric acid make-up tanks and the refueling water storage pool. The reactor coolant water volume is 11,100 cubic feet, and the minimum combined water 6 volume of reactor coolant and injection water is 69,525 cubic feet (1.97x109 cc). The minimum time for the injection phase to be completed (and the beginning of the recirculation phase) is 20 minutes. The total activity in the SIS sump water and the primary loop water is specified in Regulatory Guide 1.4 and NUREG-0737 as:
(DRN 99-2362, R11; 03-2066, R14) a)
100% of the core inventories for the noble gases, b) 50% of the core inventories for the halogens, and c) 1% of the core inventories for the other isotopes None of this activity leaves the containment until the recirculation phase is started. In the first few minutes of recirculation, it is assumed that this activity will be uniformly mixed in the total water volume of the primary loop and the SIS sump.
The second condition for the water volume activity concentration is one in which a small break occurs in the primary loop so that only a small amount of water is injected into the primary loop. It is also assumed that all the core inventories listed above in the first assumption are released into this smaller volume of water in the primary loop. This ratio of the two volumes is 6.264; thus, the source strengths in the second condition will be 6.264 times greater than in the first. The greater water/activity concentrations resulting from the second condition were used in the analysis of the hot leg sampling lines and the shutdown cooling system. They were not used for the Safety Injection or Containment Spray Systems because these systems are not aligned to their recirculation mode for a small break LOCA.
(DRN 03-2066, R14)
Waterford 3 implemented Alternative Source Term (AST) dose methodology for use in Chapter 15 analyses. Continued use of the older dose methodology based on Regulatory Guide 1.4 and NRC document TID-14844 is allowed by the NRC. Current TMI Action Plan doses and Equipment Qualification doses remain bounding and are not revised using AST with the exception of the CVAS and SBVS charcoal filter trains. The revised CVAS and SBVS filter train analyses use AST methodologies and assumptions and are similar to the LBLOCA dose analyses documented in Section 15.6.3.3.
(DRN 03-2066, R14)
For the small break scenario, Waterford 3 assumed that Shutdown Cooling is initiated six hours into the LOCA. Cooldown is assumed to be initiated two hours after the LOCA; the cooldown is assumed to take four hours, corresponding to a 50°F/hour cooldown rate from an initial RCS temperature of 550°F to SDC entry conditions at 350°F.
The source strengths for the liquid source are summarized in Tables 12.3A-2, 3, 4 and 5 for the SIS sump water total noble gases, halogens, and all other inventories respectively. Table 12.3A-2 is the total source strength and Tables 12.A-3, 4 and 5 are the ratios relative to the total.
WSES-FSAR-UNIT-3 12.3A-3 Revision 301 (09/07)
<<(DRN 03-2066, R14)
The gaseous source model used 100 percent of the core noble gases and 25 percent of the core halogens as an instantaneous release into the containment vessel. The free volume of the containment, vessel has been conservatively assumed to be 2,500,000 cubic feet (7.08x109 cc), and it is assumed that the activity is uniformly distributed within this volume. The source strengths for the gaseous source are summarized in Table 12.3A-6.
ø(DRN 03-2066, R14)
The plateout source model assumed 25 percent of the core iodine and one percent of the core rubidium and cesium instantaneously deposited on the internal surface areas of the containment. In addition to these source terms, 100 percent of the core inventories of gaseous Krypton-88, Krypton-89 and Xenon-138 are assumed in the containment and their daughter product decay to Rubidium-88, Rubidium-79 and Cesium-138, respectively are included as plateout isotopes. The internal surface area has been conservatively estimated to be 434,000 square feet (4.03x1O8 cm2), and the activity was assumed to be uniformly distributed on that surface area. The source strengths are summarized in Table 12.3A-7.
The Reactor Building contains a very large source of radiation following an accident and its effects can be noted at great distances. The dose rates at selected points within the plant site are shown on Figure 12.3A-1 and are a result of the gaseous atmosphere and plateout from within the containment. As can be seen from the figure, the five minute dose rates are high, but by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the dose rates are reduced to 0.4 percent of the five minute dose rates. This rapid decline in the dose rates is a result of the short half lives of most of the noble gases. Sixty five percent of the one year integrated dose from the containment source is received in the first six hours and 90 percent in the first day.
12.3A.3 ANALYZED SYSTEMS AND AREA DOSE RATES The systems analyzed as sources of radiation are those listed in Table 12.3A-8. All other systems such as the Chemical and Volume Control System (CVCS), Waste Management, and Boron Management are not necessary for post-LOCA operation. Degassing of the Reactor Coolant System (RCS) will be done using the RCS vent system rather than the CVCS system. The RCS vents discharge directly to the containment atmosphere or the quench tank (see Subsection 5.4.15).
The following is a description of assumptions made in computing the maximum possible dose rates in each vital area, problems associated with unacceptable dose rates, and projected solutions to those problems. The occupancy requirements and dose levels for each vital area are summarized in Table 12.3A-9.
<<(EC-5000082374, R301)
Dose rates are acceptable if they meet the requirements of NUREG 0737, i.e., less than 15 mrem/hour (averaged over 30 days) for areas requiring continuous occupancy, and GDC-19 requirements (less than 5 rem for the duration of the accident) for areas requiring irregular, not continuous, occupancy.
ø(EC-5000082374, R301) 12.3A.3.1 Control Room
<<(DRN 03-2066, R14)
The Control Room and Technical Support Center (item I on Table 12.3A-9) are located at elevation
+46.00' in the RAB (Figure 12.3A-8). Continuous occupancy is required in the control room for the duration of the accident. The bounding event with respect to control room dose are DBA events resulting in potential ADV releases based on the proximity of the east atmospheric dump valve release location and the east control room air intake. The bounding event with respect to shine doses from filter trains, containment and airborne sources is the large break LOCA analysis documented in Section 15.6.3.3.
There are six sources of radiation considered for this area. The individual contributors to the dose in the control room from each source for a large break LOCA are:
ø(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 12.3A-4 Revision 307 (07/13)
(DRN 03-2066, R14)
Approximate %
Dose Contribution
- 1)
Shield Building Ventilation System (SBVS) negligible (EC-5000082374, R301)
- 2)
Control Room Air Conditioning System (CRACS) 1.1
- 3)
Inhalation / Submersion 49.0
- 4)
Outside Cloud 0.6
- 5)
Containment Gas and Plateout 0.2
- 6)
Controlled Ventilation Area System (CVAS) 49.1 (EC-5000082374, R301)
The SBVS filters are shielded by 2.5 ft. thick concrete walls on the west side and a portion of the north side of the control room. The CVAS filters are shielded by 2.5 ft. thick concrete walls on the west side of the control room. The CRACS emergency filtration unit filters are shielded by a 1.0 ft. thick concrete wall on the north side of the control room. Additional shielding was provided to shield against shine from the CRACS Emergency Filtration Units. The dose from inhalation cannot be reduced completely; however, air tight air locks limit leakage into the area and the emergency filters remove the iodine from the control room air. The control room is shielded from the cloud outside the containment by the 2.0 ft. thick roof slab above the control room and the internal walls separating the Reactor Building and the RAB.
(DRN 03-2066, R14)
Therefore, with the additional shielding provided on the north side of the control room, the dose rates in the control room are acceptable.
12.3A.3.2 Valve Operation In order to operate the Shutdown Cooling System, valves SI-125A, SI-125B, SI-412A, SI-412B (providing a flow path between the LPSI pump discharge lines and the shutdown heat exchangers) and SI-135A and SI-135B (provides a flow path from the LPSI pump discharge lines to the LPSI pump suction lines in order to reduce thermal shock) must be opened (items 2, 3 and 4 on Table 12.3A-9). As a result of the shielding review, these valves have been provided with motor operated actuators.
Operation of the Shutdown Cooling System also requires that valves CS-117A and CS-117B be closed manually from the RAB -15' level valve gallery to isolate the Containment Spray headers. Analysis indicates that the dose rate near these valve operators is less than or equal to 23 R/hr at six hours. The resulting dose for a 10 minute entry is less than 4 Rem. Manual operation of these valves results in radiation dose within the criteria of GDC 19 and is therefore acceptable. (
Reference:
Calculation OSA-RC-CALC-91-001).
(EC-30976, R307)
Operation of the Shutdown Cooling System, after a loss of instrument air, requires that valves SI-129A and/or SI-129B be closed manually and remotely. The location of the supplemental air supplies used to close the SI-129A and SI-129B valves remotely are located in the RAB -15 level valve gallery near the handwheel operators for valves CS-117A and CS-117B. Therefore, the dose is the same experienced to close CS-117A or CS-117B at six hours.
(EC-30976, R307) 12.3A.3.3 Sampling Area The dose rates in the present reactor coolant system sampling area (item 5A on Table 12.3A-9) in the southeast corner of elevation -4.00' of the RAB are unacceptable. Post-accident sampling will, therefore, be conducted in the wing area of the RAB at elevation +21.00'. Based on the maximum dose rate of 16.3 R/hr, the post-accident sampling station is accessible following an accident for sampling and maintenance for entry durations of up to 18 minutes. The expected dose rates at the post-accident sampling station are given in Table 12.3A-9, Item 5B. Refer to FSAR Subsection 9.3.8 for further detail on post-accident sampling.
WSES-FSAR-UNIT-3 12.3A-5 Revision 11 (05/01) 12.3A.3.4 Radiochemistry Laboratory The Radiochemistry Laboratory area (item 6 on Table 12.3A-9) at elevation -4.00 of the RAB (Figure 12.3A-5) will have post-LOCA dose rates of less than 10 mrem/hr for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident and less than 1 mrem/hr thereafter. At the elevation below the radiochemistry lab, there is highly radioactive equipment such as the shutdown heat exchangers and equipment for the Containment Spray and Safety Injection Systems. However, the radiochemistry lab is well shielded from the high activity sources by the long slant paths in the walls and floors separating the two areas.
It is important to note that the reactor coolant system post-accident sampling will not be conducted in the
-4 RAB sampling area. Therefore, the dose rates generated for the Radiochemistry Laboratory Area reflect this fact. Consequently the Radiochemistry Laboratory Area will be available for continuous post-accident use for analyzing samples. This area is accessible from the control room by way of stairway A to elevation +21.00 and then stairway B to elevation -4.00.
(DRN 99-2362) 12.3A.3.5 Electrical Equipment Area The electrical equipment area (Table 12.3A-9) on the east side of elevation +21.00 of the RAB (Figure 12.3A-7) contains switches which must be activated before equipment on the diesel generator manual load block may be sequenced onto the diesel generator. This is normally done between 30 minutes and one hour into the accident.
The maximum post-accident dose rate in the "B" Switchgear Room, which extends into the RAB Wing Area, is 16.3 R/hr at 20 minutes into the postulated large break LOCA. The maximum small break LOCA dose of 14.7 R/hr occurs at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> into the event, the assumed time of SDC initiation. Personnel can enter the "B" Switchgear area for as long as 18 minutes without exceeding the GDC19 dose limit of 5 Rem.
Dose rates for the other Electrical Equipment areas are considerably lower. The Electrical Equipment areas other than the "B" Switchgear Room are continuously accessible.
The Electrical Equipment area is accessible from the control room via stairway A.
12.3A.3.6 Security Room The doses in the security room (item 8 on Table 12.3A-9) are acceptable for continuous use. However, the location of the security room is considered confidential information and as such is not included here.
The dose rates and accessibility to the security room is discussed in FSAR Subsection 13.6A.
(DRN 99-2362) 12.3A.3.7
-4 Access Point Offices The dose rates in the present -4 Access Point offices (item 9 on Table 12.3A-9; Figure 12.3A-5) at elevation -4.00 of the RAB are unacceptable. This area is located directly above the heat exchangers for the shutdown cooling system and the containment spray system which can become highly radioactive. An additional health physics area has been established at the
WSES-FSAR-UNIT-3 12.3A-6 Revision 14 (12/05) southeast corner of the +7.00 foot elevation in the RAB. This health physics area will be used during post accident conditions. It is conveniently located near the sampling areas and the chemistry laboratories.
Access will be required in this room on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis, and the dose rates are not expected to exceed 1.0 mrem/hr.
12.3A.3.8 Diesel Generator Rooms
(DRN 05-144, R14)
The Diesel Generator rooms (item 10 on Table 12.3A-9) are located on the west side of the RAB at elevation +21.00' (Figure 12.3A-7). The diesel generators are immediately to the south of the Component Cooling Water (CCW) area, and they are shielded from the CCW area and the highly radioactive Shield Building Ventilation System (SBVS) and Controlled Ventilation Area System (CVAS) filters at elevation
+46.00' by the 1.0 ft. thick ceiling slab. The slant paths and additional distance significantly reduces the dose rates in the diesel generator area so that the highest dose rate is 240 mrem/hr. Access is to these rooms for 15 minutes every eight hours for an operating parameter review. This area is accessible from the control room by stairway A and then through the electrical equipment area.
(DRN 05-144, R14) 12.3A.3.9 Waste Management System (WMS) and Boron Management System (BMS)
Control Panels The WMS and BMS control panels (item 11 on Table 12.3A-9) are in the middle of elevation -4.00' of the RAB (Figure 12.3A-5). This area must be accessible for two to four hours to run the systems in a "cold" state prior to the introduction of radioactive fluid and an additional hour to align the systems for radioactive fluid processing. The dose rates in this area are acceptable for continuous post-accident use.
This area is accessible from the control room by using stairway A to elevation +21.00' and then stairway B to elevation -4.00'.
12.3A.3.10 Heating, Ventilation and Air Conditioning (HVAC)
(DRN 03-2066, R14)
The HVAC equipment room at elevation +46 of the RAB is not accessible following an accident, and the air intake valves are controlled by remote manual operation. The emergency operation of the equipment is discussed in Subsections 9.4.3.4 and 9.4.5.9.
(DRN 03-2066, R14)
REFERENCES:
SECTION 12.3A
- 1)
- 0. J. Wallace, "SPAN-4: A Point Kernal Computer Program for Shielding," WAPD-TM-809(L),
October 1972.
- 2)
G. Martin, Jr., D. Michlewicz and J. Thomas, "FISSION 2120: A Program for Assessing the Need for Engineered Safety Feature Grade Air Cleaning Systems in Post Accident Environments",
Proceedings of the 15th DOE Nuclear Air Cleaning Conference, pp. 266 to 278, August 7-10, 1978.
(DRN 03-2066, R14)
- 3) Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors
(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 TABLE 12.3A-1 (Sheet 1 of 2)
Revision 14 (12/05)
(DRN 03-2066, R14)
TMI SOURCE TERM FOR SHIELDING EVALUATION Inventory Inventory Nuclide (Curies)
Nuclide (Curies)
(DRN 03-2066, R14)
SE-84 2.17(+7)*
TE-131m 1.77(+7)
BR-84 2.27(+7)
TE-131 1.02(+8)
AS-85 3.76(+6)
I-131 1.17(+8)
SE-85 1.34(+7)
XE-131m 8.24(+5)
BR-85 2.91(+7)
SN-132 2.01(+7)
KR-85m 2.95(+7)
SB-132 5.64(+7)
KR-85 9.36(+5)
TE-132 1.68(+8)
SE-87 2.14(+7)
I-132 1.72(+8)
BR-87 4.69(+7)
SN-133 6.98(+6)
KR-87 5.41(+7)
SB-133 6.37(+7)
BR-88 4.94(+7)
TE-133m 8.48(+7)
KR-88 7.73(+7)
TE-133 1.35(+8)
RB-88 7.85(+7)
I-133 2.36(+8)
BR-89 3.40(+7)
XE-133 2.37(+8)
KR-89 9.48(+7)
CS-134 2.23(+7)
RB-89 1.02(+8)
SB-134 1.13(+7)
SR-89 1.09(+8)
TE-134 1.79(+8)
BR-90 2.15(+7)
I-134 2.55(+8)
KR-90 9.37(+7)
SB-135 7.08(+6)
RB-90 9.63(+7)
TE-135 9.32(+7)
SR-90 7.61(+6)
I-135 2.20(+8)
Y-90 7.98(+6)
XE-135m 4.78(+7)
KR-91 6.90(+7)
XE-135 4.24(+7)
RB-91 1.24(+8)
CS-135 2.96(+1)
SR-91 1.34(+8)
CS-136 6.22(+6)
Y-91m 7.72(+7)
I-137 9.84(+7)
Y-91 1.42(+8)
XE-137 2.09(+8)
SR-95 1.43(+8)
CS-137 1.02(+7)
Y-95 1.88(+8)
BA-137m 9.68(+6)
ZR-95 1.97(+8)
I-138 4.95(+7)
NB-95 1.99(+8)
XE-138 1.89(+8)
ZR-99 1.94(+8)
CS-138 2.01(+8)
NB-99 2.04(+8)
XE-140 9.68(+7)
MO-99 2.14(+8)
CS-140 1.82(+8)
TC-99m 1.85(+8)
BA-140 2.06(+8)
MO-103 1.88(+8)
LA-140 2.12(+8)
TC-103 1.92(+8)
XE-143 2.31(+6)
RU-103 1.93(+8)
CS-143 3.92(+7)
WSES-FSAR-UNIT-3 TABLE 12.3A-1 (Sheet 2 of 2)
Revision 14 (12/05)
(DRN 03-2066, R14)
TMI SOURCE TERM FOR SHIELDING EVALUATION Inventory Inventory Nuclide (Curies)
Nuclide (Curies)
(DRN 03-2066, R14)
TC-106 7.95(+7)
BA-143 1.56(+8)
RU-106 5.45(+7)
LA-143 1.76(+8)
SN-129 1.25(+7)
CE-143 1.77(+8)
SB-129 3.89(+7)
PR-143 1.74(+8)
TE-129m 1.01(+7)
XE-144 5.15(+5)
TE-129 3.68(+7)
CS-144 1.20(+7)
I-127 1.85(+25)**
BA-144 1.16(+8)
I-129 2.94(+0)
LA-144 1.53(+8)
SN-131 3.45(+7)
CE-144 1.40(+8)
SB-131 9.50(+7)
PR-144 1.41(+8)
- Numbers in parentheses denote powers of ten
- I-127 is a Stable Isotope therefore the inventory is expressed in atoms
(DRN 03-2066, R14)
Note:
This table was evaluated for EPU and the TMI source term was determined to be bounding for EPU.
(DRN 03-2066, R14)
WSES-FSAR-UNIT-3 TABLE 12.3A-2 SIS SUMP WATER TOTAL SOURCE STRENGTHS Includes 100% Core Noble Gases, 50% Core Halogens, 1% Core All Others Diluted in 69,525 Ft3 Water - (Source Strengths /cc-sec)
Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MINS 1.69+9 6.30+9 8.13+9 6.36+9 1.55+9 1.88+9 1.02+9 1.08+8 8.99+6 20 MINS 1.69+9 4.87+9 7.09+9 5.44+9 1.24+9 1.18+9 8.59+8 2.12+7 3.42+5 1 HR 1.68+9 3.57+9 5.81+9 3.96+9 1.04+9 5.88+8 6.30+8 7.51+6 5.60+1 6 HRS 1.63+9 1.93+9 2.56+9 1.14+9 4.66+8 1.89+8 1.30+8 1.09+4 0+0 12 HRS 1.58+9 1.42+9 1.65+9 5.57+8 2.33+8 8.00+7 2.76+7 4.25+0 0+0 1 DAY 1.48+9 1.06+9 1.07+9 1.56+8 7.88+7 1.96+7 2.13+6 0+0 0+0 3 DAYS 1.14+9 7.63+8 3.41+8 3.55+6 1.12+7 1.48+5 3.46+5 0+0 0+0 1 WK 6.76+8 5.31+8 1.51+8 1.26+6 2.13+6 2.06+3 6.61+4 0+0 0+0 2 WKS 2.72+8 2.88+8 1.12+8 7.81+5 2.50+5 4.25+1 3.66+3 0+0 0+0 1 MO 3.51+7 7.40+7 7.50+7 4.33+5 1.39+5 5.96-3 4.90+0 0+0 0+0 1 YR 2.16+5 1.19+6 8.92+6 8.77+4 1.02+5 0+0 0+0 0+0 0+0
WSES-FSAR-UNIT-3 TABLE 12.3A-3 SIS SUMP WATER NOBLE GAS SOURCE STRENGTH RATIOS Source Strength Ratios Relative To Total Source Strengths Includes 100% Core Noble Gases, 50% Core Halogens and 1% Core All Others Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MINS 0.964 0.813 0.208 0.052 0.205 0.610 0.796 0.795 1.000 20 MINS 0.964 0.772 0.139 0.012 0.138 0.549 0.843 0.171 1.000 1 HR 0.967 0.720 0.099 0.003 0.164 0.484 0.926 0.000 1.000 6 HRS 0.967 0.528 0.046 0.000 0.150 0.429 0.994 0.000 12 HRS 0.968 0.354 0.017 0.000 0.068 0.249 0.976 0.000 1 DAY 0.968 0.156 0.004 0.000 0.010 0.056 0.722 3 DAYS 0.966 0.006 0.000 0.000 0.000 0.000 0.000 1 WK 0.959 0.000 0.000 0.000 0.000 0.000 0.000 2 WKS 0.940 0.000 0.000 0.000 0.000 0.000 0.000 1 MO 0.844 0.001 0.000 0.000 0.000 0.000 0.000 1 YR 0.000 0.000 0.000 0.000 0.000
WSES-FSAR-UNIT-3 TABLE 12.3A-4 SIS SUMP WATER HALOGEN SOURCE STRENGTH RATIOS Source Strength Ratios Relative To Total Source Strengths Includes 100% Core Noble Gases, 50% Core Halogens and 1% Core All Others Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MIN 0.015 0.160 0.619 0.698 0.331 0.144 0.012 0.202 0.000 20 MINS 0.015 0.689 0.794 0.410 0.206 0.010 0.823 0.000 1 HR 0.016 0.245 0.738 0.894 0.527 0.331 0.006 0.999 0.000 6 HRS 0.016 0.411 0.695 0.929 0.754 0.570 0.000 1.000 12 HRS 0.016 0.562 0.630 0.867 0.759 0.750 0.000 1.000 1 DAY 0.016 0.734 0.535 0.664 0.592 0.942 0.000 3 DAYS 0.018 0.872 0.227 0.016 0.018 0.885 0.000 1 WK 0.021 0.877 0.062 0.000 0.000 0.003 0.000 2 WKS 0.028 0.858 0.032 0.000 0.000 0.000 0.000 1 MO 0.053 0.773 0.009 0.000 0.000 0.000 0.000 1 YR 0.000 0.000 0.000 0.000 0.000
WSES-FSAR-UNIT-3 TABLE 12.3A-5 SIS SUMP WATER ALL OTHERS SOURCE STRENGTH RATIOS Source Strength Ratios Relative To Total Source Strengths Includes 100% Core Noble Gases, 50% Core Halogens and 1% Core All Others Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MINS 0.021 0.027 0.173 0.250 0.464 0.246 0.192 0.003 0.000 20 MINS 0.021 0.031 0.172 0.194 0.452 0.245 0.148 0.006 0.000 1 HR 0.017 0.035 0.253 0.103 0.309 0.185 0.068 0.001 0.000 6 HRS 0.017 0.061 0.259 0.071 0.096 0.001 0.006 0.000 12 HRS 0.016 0.084 0.353 0.133 0.173 0.001 0.024 0.000 1 DAY 0.016 0.110 0.461 0.336 0.398 0.002 0.278 3 DAY 0.016 0.122 0.773 0.984 0.982 0.115 1.000 1 WK 0.020 0.123 0.938 1.000 1.000 0.997 1.000 2 WKS 0.032 0.142 0.968 1.000 1.000 1.000 1.000 1 MO 0.103 0.226 0.991 1.000 1.000 1.000 1.000 1 YR 1.000 1.000 1.000 1.000 1.000
WSES-FSAR-UNIT-3 TABLE 12.3A-6 CONTAINMENT GASEOUS TOTAL SOURCE STRENGTHS Includes 100% Core Noble Gases and 25% Core Halogens Diluted in 2,500,000 Ft3 Air - (Source Strengths /cc-sec)
Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MINS 4.61+7 1.52+8 1.32+8 9.86+7 2.61+7 4.79+7 2.72+7 2.64+6 2.50+5 20 MINS 4.61+7 1.13+8 1.09+8 8.00+7 1.97+7 2.81+7 2.29+7 3.36+5 9.50+3 1 HR 4.59+7 7.91+7 8.56+7 5.58+7 1.63+7 1.21+7 1.70+7 1.04+5 1.56+0 6 HRS 4.47+7 3.74+7 3.56+7 1.57+7 6.61+6 3.44+6 3.57+6 1.51+2 0+0 12 HRS 4.32+7 2.45+7 2.18+7 7.72+6 2.94+6 1.30+6 7.38+5 0+0 0+0 1 DAY 4.05+7 1.60+7 1.36+7 2.17+6 7.55+5 2.81+5 3.72+4 0+0 0+0 3 DAYS 3.12+7 1.04+7 3.37+6 1.49+4 5.07+3 1.80+3 0+0 0+0 0+0 1 WK 1.84+7 7.31+6 7.94+5 0+0 0+0 0+0 0+0 0+0 0+0 2 WKS 7.38+6 4.00+6 3.86+5 0+0 0+0 0+0 0+0 0+0 0+0 1 MO 9.16+5 1.01+6 9.89+4 0+0 0+0 0+0 0+0 0+0 0+0 1 YR 0+0 0+0 1.88+3 0+0 0+0 0+0 0+0 0+0 0+0
WSES-FSAR-UNIT-3 TABLE 12.3A-7 CONTAINMENT PLATEOUT TOTAL SOURCE STRENGTHS
- Includes 25% Core Iodine and 1% Core Cesium and Rubidium Plus Daughter Products Of Noble Gas Decay - (Source Strengths /cc-sec)
Upper Mev 0.106 0.440 0.865 1.332 1.720 2.210 2.754 3.930 4.702 ENG GP #
8 7
6 5
4 3
2 1
0 Time 5 MINS 7.07+8 3.34+10 1.83+11 2.29+11 5.33+10 2.35+10 1.01+10 0+0 0+0 20 MINS 7.07+8 3.22+10 1.86+11 2.12+11 8.73+10 3.10+10 1.35+10 0+0 0+0 1 HR 7.05+8 2.98+10 1.52+11 1.24+11 7.13+10 2.18+10 7.33+9 0+0 0+0 6 HRS 6.92+8 2.50+10 5.84+10 2.80+10 9.48+9 3.59+9 5.60+7 0+0 0+0 12 HRS 6.78+8 2.36+10 3.76+10 1.37+10 4.66+9 1.72+9 9.97+6 0+0 0+0 1 DAY 6.50+8 2.19+10 2.42+10 3.88+9 1.31+9 4.66+8 5.15+5 0+0 0+0 3 DAYS 5.47+8 1.82+10 6.49+9 8.17+7 1.58+7 3.16+6 0+0 0+0 0+0 1 WK 3.89+8 1.29+10 1.96+9 4.60+7 6.91+6 1.50+2 0+0 0+0 0+0 2 WKS 2.14+8 7.06+9 1.23+9 3.34+7 6.87+6 0+0 0+0 0+0 0+0 1 MO 5.47+7 1.78+9 6.99+8 1.75+7 6.77+6 0+0 0+0 0+0 0+0 1 YR 0+0 0+0 3.99+8 4.24+6 4.97+6 0+0 0+0 0+0 0+0
- Uniformly distributed over the estimated 434,000 Ft2 of surface area inside the containment.
WSES-FSAR-UNIT-3 TABLE 12.3A-8 Revision 11 (05/01)
SYSTEMS CONTAINING RADIOACTIVE MATERIAL
System and Major Components Location Figure Safety Injection System Safety Injection Tanks EL + 46.00 of RCB 1.2-17 LPSI pumps NE corner RAB 12.3A-4 HPSI pumps
@ elevation - 35.00 Piping & Valves SIS sump
Shutdown Cooling System Shutdown Heat Exchangers NE corner of RAB @
12.3A-4 Piping & Valves elevation - 35.00 Containment Spray System Containment spray pumps NE corner of RAB @
12.3A-4 Containment spray nozzles elevation - 35.00 Piping & Valves Sampling System Sample heat exchangers Sample collecting tank Piping & Valves Controlled Ventilation Area System, Shield Building Ventilation System & Control Room Emergency Ventilation System Exhaust Fans NW corner of RAB 12.3A-8 Filters
@ elevation + 46.00 Ductwork & Dampers
WSES-FSAR-UNIT-3 TABLE 12.3A-9 (Sheet 1 of 2)
Revision 307 (07/13)
AREAS REQUIRING ACCESSIBILITY FOLLOWING AN ACCIDENT AREA LOCATION (Figure & Column Lines)
OCCUPANCY REQUIREMENTS MAXIMUM DOSE RATE (mrem/hour)
ACCEPTABLE REMARKS (EC-5000082374, R301)
- 1)
Control Room & T.S.C.
RAB EL. +46.00' Figure 12.3A-8 Continuous
<5 Rem post-LOCA, per Chapter 15.6 Yes (c)
Additional shielding has been added on north side of the control room.
(EC-5000082374, R301;)
(EC-30976, R307)
- 2)
Valve Bay RAB EL. -15.5' (9A and K)
Figure 12.3A-4 a) One man for 10 minutes for each of valves SI-125A, SI-125B, SI-412A, SI-412B.
Valve operation required prior to shutdown cooling b) One man for 10 minutes to close CS-117A and/or CS-117B and close SI-129A and/or SI-129B with supplemental air supply prior to initiating shutdown cooling.
5.1 x 104 2.3 x 104 No Yes Motor Operated Actuators have been provided Acceptable since total dose <
- 3)
Valve Station RAB Wing Area EL. -35.00' (3A and N) Figure 12.3A-2 Two men for 10 minutes to operate valve SI-135B. Valve operation required prior to shutdown cooling.
9.1 x 105 No Motor Operated Actuators have been provided
- 4)
Valve Station RAB Wing Area EL. -35.00' (11A and N) Figure 12.3A-2 Two men for 10 minutes to operate valve SI-135A. Valve operation required prior to shutdown cooling.
9.1 x 105 No Motor Operated Actuators have been provided (DRN 99-2362, R11)
- 5)
Sampling Room (DRN 99-2362, R11) a) Normal Operation RAB EL. -4.00' (11A and H) Figure 12.3A-5 Infrequent access 2.2 x 106 No (d) b) Post-Accident RAB Wing Area EL. +21.00' (10AZ and N) Figure 12.3A-3 Infrequent access 1.63 x 104 Yes Access for 18 minutes is acceptable since this would result in a dose of 5 Rem.
- 6)
Radiochemistry Lab/
Counting Room RAB EL. -4.00' (10A and H)
Figure 12.3A-5 Infrequent access 9.7 Yes (c)
(DRN 99-2362, R11)
- 7)
Electrical Equipment Area RAB Wing EL. +21.00' east side Figure 12.3A-7 One man-hour immediately, post-LOCA 1.63 x 104 Yes Access to Switchgear Room "B" for 18 minutes is acceptable since this would result in a dose of 5 Rem.
Other Electrical Equipment Areas are not in the RAB wing Area and are continuously accessible; for these other areas, the dose rate is 2.62 mrem/hour when averaged over the 30 days.
(DRN 99-2362, R11)
WSES FSAR UNIT 3 TABLE 12.3A 9 (Sheet 2 of 2)
Revision 14 (12/05)
AREAS REQUIRING ACCESSIBILITY FOLLOWING AN ACCIDENT AREA LOCATION (Figure & Column Lines)
OCCUPANCY REQUIREMENTS MAXIMUM DOSE LEVEL (mrem/hour)
ACCEPTABLE REMARKS
- 8)
Security Room Proprietary Proprietary Proprietary Yes
(DRN 99-2362, R11)
- 9) 4 Access Control Point Offices RAB EL. 4.00' (11A and J) Figure 12.3A 5 From limited to continuous 1.3 x 104 No A new location or additional shielding will be provided
(DRN 99-2362, R11)
(DRN 05-144, R14)
- 10)
Diesel Generator Rooms RAB EL. +21.00' (4A and J)
Figure 12.3A 7 When diesel generators are operating, access required for 15 minutes every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for operating parameter review. When diesel generators are not operating, units must be test started within 30 days.
(One man for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> each unit.)
240.0 Yes (e)
(DRN 05-144, R14)
- 11)
Waste Management System &
Boron Management System Control Panels RAB EL. 4.00' (7A and J)
Figure 12.3A 5 One man for 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to run systems in "cold" state prior to introduction of radioactive fluid.
One man for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to align system for radioactive fluid processing.
One man for 30 minutes every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for equipment surveillance, local sampling, etc.
8.1 Yes (c)
Note (c):
Indicates that the dose in this area is acceptable for continuous operation.
(DRN 99-2362, R11;03-114, R12-B)
Note (d): This is the postulated worst case dose rate assuming samples are obtained from the normal primary sample system. This area may be accessible if the conditions permit access based upon radiation surveys and access would result in a dose of 5 REM.
(DRN 99-2362, R11;03-114, R12-B)
(DRN 05-144, R14)
Note (e): Acceptable indicates that operators could make the required rounds to diesel generators, as described in Section 12.3A.3.A without exceeding 5 REM.
(DRN 05-144, R14)
WSES-FSAR-UNIT-3 12.4-1 Revision 12 (10/02) 12.4 DOSE ASSESSMENT 12.4.1 ANTICIPATED DOSE RATES Peak external dose rates expected are those given as the maximum dose rates present in a designated access zone area shown in Figures 12.3-1 through 12.3-8. The access area zones have been designated by either anticipated radiation level of the equipment in the area or the desired radiation levels to be achieved through shielding. The peak external doses are not expected to occur during normal operation but could occur due to infrequent anticipated operational occurrences during which times the plant might be operating with maximum coolant activities corresponding to one percent defective fuel cladding. The annual average isotopic concentrations of fission products are expected to be much less than the design maximum, and therefore the actual dose rates in a given zone are expected to be significantly less (approximately one eighth of the design maximum) than the maximum calculated dose rate in that zone. Another source of radiation exposure within the plant comes from airborne radionuclides. Occupational exposure to the source is usually insignificant in comparison to direct radiation exposure. However under certain circumstances of coolant leakage in containment coupled with high coolant activity, doses from airborne activity could become a major fraction of the allowed limits. Proper purging however should maintain airborne exposures so low that their contribution to the total man-rem is insignificant in comparison with exposure to direct radiation.
Radiation exposures from direct radiation at locations outside the plant structures are insignificant since there are no potential sources of radiation outside the plant structures. Therefore the total exposure from the plant is given essentially by the occupational exposure to direct radiation within plant structures. Inside equipment compartments or adjacent to equipment carrying radioactive material, the anticipated dose rates will result from the function of the equipment. The highest doses in the plant occur in Zone V areas, such as in rooms containing equipment and piping handling highly radioactive fluids, or in the containment. Again it is emphasized that experience in the design of power plants shows that the measured radiation levels are usually less than those used as shielding design objectives for controlling the radiation doses. The annual doses received by the plant personnel will therefore be well below the limits of 10CFR20 since the plant shielding and access control are designed based on radiation levels from maximum coolant activities.
¨ (DRN 99-2362)
The main control room and office areas in the Reactor Auxiliary Building will be Zone I areas and hence will have a maximum allowable dose rate of 0.25 mrem/hr as in Table 12.3.1. Therefore, annual doses in these areas, considering occupancy factor, will be well within the limits of IOCFR20, particularly since the expected dose rates in those areas will be well below 0.25 mrem/hr. PWR operating experience confirms the above assertions.
(DRN 99-2362) 12.4.2 ESTIMATES OF PLANT PERSONNEL EXPOSURES (Direct) 12.4.2.1 General The annual direct exposure that could be received by plant personnel during reactor operations and surveillance, routine and special maintenance, in-service inspection, radwaste handling and refueling has been estimated to verify that personnel exposures would be less than 10CFR20 limits.
¨(DRN 02-110)
The expected doses to individuals at or beyond the site boundary are shown in Table 11.3-7.
WSES-FSAR-UNIT-3 12.4-2 Revision 12 (10/02)
For the purposes of the estimate, reactor operations and surveillance, routine and special maintenance, in-service inspection, radwaste handling and refueling have been defined as follows:
a)
Reactor Operations and Surveillance: Include main control room work, local operations such as valve alignment, starting and stopping pumps, operation of radwaste system (but not the actual handling of the solid wastes), periodic jobs such as sampling, radiation and contamination surveys, occasional lubrication, etc.
b)
Routine and Special Maintenance: Routine maintenance includes periodic preventive maintenance operations such as oil changes, controls calibration, performance tests, and overhauls with possible preventive repairs. Special maintenance include unplanned repairs, or infrequent major jobs such as steam generator tube plugging, etc.
c)
In-service Inspection: Primarily inspection for pressure boundary integrity. As such the testing includes removal of insulation, testing of the component installation of the insulation, and possibly the removal of the reactor vessel fuel and internals.
Exposures from routine operational inspections are not included in in-service inspection, nor are those associated with work on steam generation parts. The former are included in a) and the latter in b).
¨ (DRN 99-2362) d)
Radwaste Handling: Includes actual handling operations of the plant wastes such as resin and concentrates handling, filter handling, and handling of low activity wastes such as rags, plastics, etc.
(DRN 99-2362) e)
Refueling: Include exposures associated with all phases of refueling operations, but does not include exposures from surveillance or maintenance which for convenience may take place concurrently with refueling operations, but which are not part of refueling.
The assumptions in the estimate are as follows:
a)
The power plant is staffed by the personnel shown in the organizational chart of Chapter 13.
b)
The plant is operated in continuous shifts. Each shift should have as a minimum four members who hold an NRC license.
In addition two more auxiliary operators will be part of the shift. At all times at least two and often three of the above will be in the main control room. Two or three will be able to perform routine surveillance and inspection functions in other areas of the plant.
c)
Maintenance personnel are normally available during the day shift only. On Shift Maintenance personnel cover all shifts normally performing preventive maintenance activities.
d)
Shift Technical Advisors are on continuous shifts.
¨(DRN 02-110)
From an operational basis, the Annual Radioactive Effluent Release Report provides an assessment of radiation doses to the public due to liquid and gaseous radioactive effluents.
WSES-FSAR-UNIT-3 12.4-3 Revision 13 (04/04)
¨(DRN 03-1135, R13)
The technical staff will spend most or all of their time in uncontrolled areas where the radiation levels are expected to be less than 0.03 mr/hr. For instance the Engineering Personnel will be involved essentially in office work except during refueling. Likewise the Radiation Protection Manager, the Chemistry Superintendent and their Associates and Technicians will spend a considerable part of their shift in office work. The Radiation Protection Technicians however, will be engaged in periodic plant surveys and Health Physics Technicians will be present during many maintenance operations.
The technicians may spend approximately one third of their shifts in laboratory work and collecting certain samples.
(DRN 03-1135, R13)
¨(DRN 02-110, R12)
Actual annual exposure received during the plant operation years has been less than the estimated exposure listed in Table 12.4-1a. ALARA concepts and gained operation experiences with strong management support contributed to the exposure drop during the past years.
(DRN 02-110, R12) 12.4.2.2 Reactor Operations and Surveillance Exposures for reactor operations and surveillance are primarily incurred by operation personnel which patrol the plant during all shifts and operates it. The technical staff performs the majority of their work in offices and laboratories except for the surveys of the plant performed by Health Physics personnel and the collection of local samples by chemistry personnel.
The operations people, primarily the auxilary operators, patrol the plant every shift, check for leaks, inspect equipment, and during the day shift operate the blowdown, boric acid, and waste management systems.
They also perform minor lubrication Jobs.
The exposures listed in Table 12.4-la are predicated on surveillance of the plant with inspection for checkup of every active equipment, and daily operation of the blowdown, boric acid, and waste management system panels.
The total exposures predicted agree reasonably well with the average exposures for reactor operation and surveillance deduced from Reference 1.
¨(DRN 02-110, R12)
Actual annual exposure received during the plant operation years has been less than the estimated exposure listed in Table 12.4-1a. ALARA concepts and gained operation experiences with strong management support contributed to the exposure drop during the past years.
(DRN 02-110, R12) 12.4.2.3 Maintenance Two kinds of maintenance are performed in a power plant: routine maintenance and special maintenance.
Exposures resulting from the former can be estimated by examining each individual maintenance item, identifying the personnel required, the time involved, and computing the resulting exposures. This maintenance includes the testing and check out of equipment, lubrication, etc, which are normally performed by the operators; tasting and calibration of controls, valve maintenance, overhaul, sump cleaning and so on, which are performed by maintenance personnel.
Special maintenance includes jobs of unanticipated occurrence and frequency. Estimates of the exposures from this type of maintenance are therefore, based entirely on past experience (References 1 and 2).
Table 12.4-lb lists a breakdown of the activities estimated for maintenance at the Waterford 3 Plant.
The exposures associated with each activity have been derived considering the time and number of personnel required to perform routine maintenance of pumps, compressors, and selected instrumentation.
WSES-FSAR-UNIT-3 12.4-4 Revision 12 (10/02)
The maintenance involves oil changing, greasing, disassembly, including "cutting and rewelding, removal to hot machine or instrumentation shop, reassembly, etc. For general valve maintenance and instrumentation and control work, exposures have been derived from Reference 2. The distribution of these exposures is allocated to the individual disciplines in the table on the basis of the efforts normally required by the disciplines to perform an individual valve maintenance.
Exposures resulting from special maintenance have also been derived from Reference 2. Guidance from that document and operating utilities reports have been used to distribute the exposures for special maintenance among the worker's categories.
It has been assumed that major repair work would be handled by outside contractors. From Reference 2, the average amount of exposure in manrem for major equipment failure type work has been derived to be approximately 30 man-rems.
¨(DRN 02-110)
Actual annual exposure received during the plant operation years has been less than the estimated exposure listed in Table 12.4-1a. ALARA concepts and gained operation experiences with strong management support contributed to the exposure drop during the past years.
(DRN 02-110) 12.4.2.4 In-service Inspection
¨ (DRN 99-0823)
It is expected that exposures from in-service inspection activities will be borne primarily by outside contractors, site non-desctructive examination, and operation personnel with some exposure received by Waterford 3 supervisory personnel.
Exposures due to inservice inspection activities have averaged approximately 8.2 man-rem per outage. This average is based on the exposure received during outages contained in the first inservice inspection interval.
Therefore, in Table 12.4-1, the total exposure for inservice inspection is estimated also as 5.5 man-rem per year.
(DRN 99-0823) 12.4.2.5 Waste Processing Exposures resulting from waste handling evolutions are primarily confined to operating personnel and Radwaste personnel who operate the solid waste handling and solidification area, perform the spent resin transfers, perform the periodic filter replacements, and package the low activity radioactive wastes. Some additional exposure will also be received by Health Physics and Chemistry personnel who support the sampling and radiological monitoring activities affiliated with waste processing activities.
¨ (DRN 99-2362)
A large fraction of the exposure will result from filter handling. The exposures for this operation have been estimated by assuming a frequency of filter change based on plant experience for that type of filter and a sequence of operation. The sequence of operation involves the installation of the filter transfer shield container, unbolting and swinging the filter head, withdrawal of the filter, removal of the shield container and transport to the filter packaging area. insertion of new filter and repetition of the opening phases in reverse.
¨(DRN 02-110)
Actual annual exposure received during the plant operation years has been less than the estimated exposure listed in Table 12.4-1a. ALARA concepts and gained operation experiences with strong management support contributed to the exposure drop during the past years.
WSES-FSAR-UNIT-3 12.4-5 Revision 12 (10/02) 12.4.2.6 Refueling Operations Exposures during refueling have been estimated on the basis of known tasks and estimated time required for the performance of these tasks. The forecasted exposures are tabulated by personnel categories in Table 12.4 - 1 a&b. These exposures do not include exposures resulting from maintenance (either special or routine) or in-service inspection activities which are likely to be conducted during the refueling outage. Of the total exposure, approximately nine man-rem are associated with the actual fuel handling operations. 18 man-rem are associated with health physics surveys and building decontamination with the remaining 35 man-rem being due to head removal and reinstallation.
It must be emphasized that exposures during refueling are difficult to evaluate on the basis of predicted) dose rates, since the number of persons involved will vary.
Experience (3) indicates that approximately 29 manrem occur at PWR plants per removal and reinstallation of the head. These exposures appear to be primarily due to activation products accumulation in the control rod drive mechanism. The predicted exposures are then in good agreement with those experienced in the past. The exposures accrued during the actual fuel handling may vary between six and 33 man-rem/yr.
This range appears to be caused by random difficulties encountered in the removal and replacement of fuel assemblies. The nine man-rem of this estimate considers such difficulties only to the extent of employing two full weeks for the fuel shuffling operation. Since these difficulties are encountered commonly, the personnel dose could be higher.
¨(DRN 02-110)
Actual annual exposure received during the plant operation years has been less than the estimated exposure listed in Table 12.4-1a. ALARA concepts and gained operation experiences with strong management support contributed to the exposure drop during the past years.
(DRN 02-110) 12.4.2.7 Comparison With Exposure Experience at Other PWR Plants Table 12.4-2 presents data on total number of personnel and total annual dose in operating PWRs. These data are taken from References 1 and 3 and is further supported by data given in Reference 2. Data from this table is reduced to a weighted average expressing the average man-rem for operating power plants in Table 12.4-3, which also shows the good agreement between this average and the exposure estimated for Waterford 3.
Table 12.4-4 derived from Reference 1, lists the distribution of man-rem doses for various functions of operating light water reactors (including BWRs) and compares it with that estimated for Waterford 3.
WSES-FSAR-UNIT-3 12.4-6 SECTION 12.4: REFERENCES 1.
NUREG-0322. Ninth Annual Occupational Radiation Exposure Report. 1976. USNRC Office of Management. Information and Program Control. Oct. 1977 2.
National Enviromental Studies Project. Compilation and Analyses of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants SAI Services. Sept. 1974 3.
Murphy, T D et al. NUREG-75/032. Occupational Radiation Exposure at Light Water Cooled Power Reactors 1969-1974. USNRC Radiological Assessment Branch. June, 1975
WSES-FSAR-UNIT-3 TABLE 12.4-1a Revision 11 (05/01)
ESTIMATE OF PERSONNEL EXPOSURE (MAN-REM)
Maintenance Tech Staff Others Activity Total Man-Rem/Yr.
Operation &
Maintenance Supv./
Control & Maint Op/
Maintenance Op.
Mech.
Elect I & C Nucl.
Eng./Asso.
Nuc. Eng.
Health Phys/
Engr. Tech/
Tech Chem Eng/
Asso. Eng/
Engr. Tech/
Reactor Operations
& Surveillance 17.5 13.0 0.115 2.25 2.25 Maintenance 291 35.0 120 3.0 7.0 25.0 1
20 80
In-service Insp.
5.5 2.0 1.5 2.0
Waste Processing 5.5 2.3 1.15 1.1 1.0 Refueling 51 7.0 17.8 0.2 1.9 1.8 1.67 1.0 2.5 1.2
Total 370.5 59.3 138 3.0 9
2.0 46 4
1 24 84.2
WSES-FSAR-UNIT-3 TABLE 12.4-1b (Sheet 1 of 2)
Revision 11-A (02/02)
ESTIMATE OF PERSONNEL EXPOSURE DURING MAINTENANCE (MAN-REM)
Maintenance Tech Staff Others Activity Operation &
Maintenance Supv./
Control & Maint Op/
Maintenance Op.
Mech. Eng.
Elect I & C Nucl.
Eng./Asso.
Nuc. Eng.
Health Phys/
Engr. Tech/
Tech Chem Eng/
Asso. Eng/
Engr. Tech/
Tech QA NSSS Outside Contractor ROUTINE MAINT.
a) Waste Gas Compressors
.085 0.56 0.26 0.16 b) All Valve Work 2.4 8.2 0.4 0.4 1.7 c) Inst. & Controls 5.8 3.6 d) Charging Pumps
.09 2.16 0.22 0.27 0.48
.04 e) BA Condensate Tanks & Pumps
.02 0.37 0.04 0.16 f) Hold-Up Pumps
.03 0.55 0.06 0.24 g) BA Make-up Tanks & Pumps
.02 1.27 0.55 0.175 h) Sump Pumps & Sumps 0.12 6.44 0.48 3.92 i) Waste Condensate Tanks, Laundry Tanks & Pumps 0.04 1.48 0.16 0.64 j) Waste Tanks & Pumps 0.06 1.44 0.08 0.12 0.32 k) Shutdown HX Room 0.06 0.03 0.015 l) Safety Inj. Pumps 0.27 8.0 0.48 0.81 1.92 0.32 m) BA & Waste Concentratr.
0.10 0.20 0.05 0.10 0.10 n) Process Monitors 0.18 1.8 0.18
¨(DRN 00-1046) o) Hot Tool Room 0.10 0.20 0.05 0.10 0.10 (DRN 00-1046) p) Concentrate Tank 0.45 0.07 0.07 q) Containment Work 1.0 1.0 1.0
WSES-FSAR-UNIT-3 TABLE 12.4-1b (Sheet 2 of 2)
ESTIMATE OF PERSONNEL EXPOSURE DURING MAINTENANCE (MAN-REM)
Maintenance Tech Staff Others Activity Operation &
Maintenance Supv./
Control & Maint Op/
Maintenance Op.
Mech. Eng.
Elect I & C Nucl.
Eng./Asso.
Nuc. Eng.
Health Phys/
Engr. Tech/
Tech Chem Eng/
Asso. Eng/
Engr. Tech/
Tech QA NSSS Outside Contractor SPECIAL MAINT.
a) Major Equipment Failure 30 b) S.G. Repair 11 45 3
0.2 10 45 c) Main Coolant Loops 5
11 2
0.2 1
d) R.C. Pump 5
7 2.5 0.2 2
e) Fuel Pool Decon. Weld 2
4 1
f) Maint. of Loose Parts Mon.
Syst.
1 1
g) Chem & Vol. Control Syst.
1 2
.5 1
h) Pressurizer 3.5 9
1.0 i) Clean Up Containment 2
9 1.0 j) Constr. Work on Plant 5
k) Work Inside Reactor Vessel 5
GRAND TOTAL TOTALS ROUNDOFF 35.6 35 1.20 1.20 2.3 3.0 7.6 7.0 25.3 25.0
.96 1.0 20 20 81 80
WSES-FSAR-UNIT-3 TABLE 12.4-2 (Sheet 1 of 2)
DATA FROM OPERATING PWR PLANTS (a)
Year Plant Designed Power Level (MWe)
Total No. of Personnel Total Annual Dose (Man-Rem) 1970 Connecticut Yankee San Onofre - Unit 1 575 450 734 251 689 155 1971 Connecticut Yankee Ginna San Onofre - Unit 575 490 450 289 340 121 342 430 50 1972 Connecticut Yankee Ginna Point Beach - Unit 1 Robinson San Onofre - Unit 1 575 490 497 707 450 355 677 NA 245 326 325 1,032 580 215 256 1973 Connecticut Yankee Ginna Palisades Point Beach - Units 1
& 2 (2nd Unit 4/73)
Robinson San Onofre - Unit 1 575 490 821 497, 497 707 450 841 421 901 729 831 878 673 244 1,109 570 695 329 1974 Connecticut Yankee Fort Calhoun Ginna Haddam Neck Maine Yankee Oconee - Unit 1 Palisades Point Beach - Units 1 & 2 Robinson San Onofre Surry - Units 1 & 2 (Unit 2 5/73)
Turkey Point - Units 3 & 4 (Units 4 9/73) 575 457 490 575 790 886 821 497, 497 707 450 823, 823 745, 745 550 327 884 550 610 844 774 400 853 219 1,715 794 201 71 1,225 201 420 517 627 295 672 71 884 454 a.
These are taken from data for operating PWR plants given in Reference 1 and 3. In compiling this table, generally data from the first year of plant operation has not been considered. Only data from those PWR plants that are designed to operate at power levels greater than or equal to 450 MWe were chose.
NA:
Not available
WSES-FSAR-UNIT-3 TABLE 12.4-2 (Sheet 2 of 2)
Year Plant Designed Power Level (MWe)
Total No. of Personnel Total Annual Dose (Man-Rem) 1975 Arkansas One Unit 1 Calvert Cliffs 1 Fort Calhoun Haddam Neck Kewaunee Maine Yankee Oconee Unit 1 Palisades Point Beach 1 & 2 Robinson San Onofre Surry 1 & 2 Turkey Point 3 & 4 850 1,065 457 575 560 790 886 821 497, 497 707 450 823, 823 745, 745 147 783 469 685 104 440 829 495 339 849 424 1,948 1,176 21 77 294 538 28 319 497 306 459 1,142 292 1,649 876 1976 Arkansas One Unit 1 Calvert Cliffs D C Cook Fort Calhoun Haddam Neck Kewaunee Maine Yankee Oconee 12 Palisades Point Beach 1 & 2 Prairie Island 1 & 2 Robinson San Onofre Surry 1 & 2 Three Mile Island Turkey Point 3 & 4 Zion 1 & 2 850 1,065 1,090 457 575 560 790 886 821 447 NA 707 450 823, 823 819 745, 745 1,015, 1,015 476 507 395 516 758 381 244 1,215 742 313 818 597 1,330 2,753 819 1,647 774 289 74 116 313 636 270 85 1,026 696 459 447 715 880 3,165 286 1,184 571
WSES-FSAR-UNIT-3 TABLE 12.4-3 YEARLY AVERAGES AND GRAND AVERAGE FOR NUMBER OF PERSONNEL AND MAN-REM DOSES FOR OPERATING PWR PLANTS (a)
Year No. of Units Total No.of Personnel Total Man-Rem Dose Average No. of Personnel Average Man-Rem Dose 1970 2
985 844 493 422 1971 3
750 822 250 274 1972 5
1,603(b) 2,408 401 482 1973 7
4,601 3,620 657 517 1974 15 8,529 5,538 568 369 1975 16 9,483 7,794 593 487 1976 23 15,416 12,031 670 523 1970 -
1976 71 41,367(c) 33,057 583 466 (a)
This table is based on the data given in Table 12.4-2.
(b)
The entry corresponds to four plants only, since no information on personnel is available for Point Beach, Units 1.
(c)
The entry corresponds to a total of 70 plants only.
WSES-FSAR-UNIT-3 TABLE 12.4-4 DISTRIBUTION MAN-REM DOSES FOR VARIOUS FUNCTIONS Operating Light Water Reactors*
(Includes BWRs)
Waterford 3 Percentage of Percentage of Total Man-Rem Total Man-Rem Reactor Operation
& Surveillance 10.4 4.2 Maintenance (Routine & Special) 71.2 70.5 In-Service Inspection 5.7 11.8 Waste Processing 4.8 1.3 Refueling 7.9 12.2
- Reference 1 Table 5 1976 data
WSES-FSAR-UNIT-3 12.5-1 Revision 307 (07/13) 12.5 HEALTH PHYSICS PROGRAM 12.5.1 ORGANIZATION 12.5.1.1 Program and Staff Organization (DRN 02-110, R12)
The health physics program specifies guidelines for handling all radioactive materials at Waterford 3, including those being received and those in preparation for shipment offsite. Guidelines cover special nuclear, source, and byproduct materials. The program assures that the station operation meets the radiation protection and training requirements of 10CFR19, 10CFR20, 10CFR50 Appendix I, and NRC Regulatory Guides 8.2, February 1973; 8.3, February 1973; 8.4, February 1973; 8.7, June 1992; 8.8, March 1977; 8.9, September 1973; 8.10, September 1975; 8.15, October 1976; 1.16, August 1975, and 1.39, September 1977. The program assures that radiation protection training is provided to workers, that personnel and inplant radiation monitoring is performed, and records of training, exposure, and surveys are maintained. It also assures a commitment to maintain exposures as low as reasonably achievable (ALARA) is fulfilled.
(DRN 02-110, R12)
(DRN 03-1135, R13; LBDCR 13-005, R307)
The Waterford 3 organization, including the health physics organization, is discussed in Section 13.1. The General Manager-Plant Operations is responsible for the overall performance of the health physics program. He delegates the administration of the program to the Radiation Protection Manager. The Radiation Protection Manager is equivalent to the radiation protection manager referred to in NRC regulatory guides. He is responsible for administering the station radiation protection program, with support from the Chemistry Superintendent who is responsible for certain aspects of the health physics program including radioactive effluent releases, radiological environmental monitoring, and some radioactivity measurements in support of plant health physics activities.
(LBDCR 13-005, R307)
Reporting to the Radiation Protection Manager are personnel functionally responsible for the areas of personnel dosimetry, ALARA, HP job coverage, instrument and respiratory protection program and radiological engineering.
(DRN 03-1135, R13)
Reporting to the Chemistry Superintendent are supervisory personnel functionally responsible for the areas of primary chemistry, secondary/auxiliary chemistry, effluent releases and environmental monitoring. Also reporting to the Chemistry Superintendent are technical personnel providing expertise in the areas of radiochemistry/chemical engineering.
Personnel assigned to the health physics organization, as designated by administrative procedures, will perform various radiation protection activities. They observe work in progress and ensure that radiation safety guidelines are followed.
12.5.1.2 Program Objectives The objectives of the health physics program are:
a) to provide radiation protection controls for personnel and operations onsite, b) to ensure that personnel exposures to radiation and radioactive materials are within the guidelines of 10CFR20 and that such exposure is kept ALARA, c) to ensure that all radioactive effluent releases and waste shipments meet guidelines established in plant procedures, and d) to ensure that radioactive effluent releases are within Offsite Dose Calculation and Technical Requirements Manual requirements and kept ALARA.
WSES-FSAR-UNIT-3 12.5-2 Revision 305 (11/11) 12.5.1.3 Health Physics Program The station health physics program was implemented when radioactive material under the Waterford 3 license was initially brought onsite, and will be maintained throughout the life of the plant. The program includes management and worker philosophies, practices, guidelines, and procedures to ensure that the program objectives stated above are fulfilled in a reasonable manner.
The health physics program ensures that:
a) all radiation workers receive radiation protection training commensurate with their respective responsibilities, (DRN 99-2362, R11) b)
respiratory protection equipment training is provided to radiation workers who may use the equipment, (DRN 99-2362, R11) c)
emergency plan training is provided as necessary for personnel who may be assigned to radiation emergency teams, d) appropriate personnel dosimetry is available, e) internal and external dose assessment is provided for monitored workers, f) personnel contamination monitoring equipment is used to assess personnel contaminations, g) respiratory protection equipment is provided if necessary to keep internal exposure ALARA, (DRN 99-2362, R11; EC-27665, R305) h)
Radiologically Controlled Areas (RCA) are segregated to control potential radiological exposures, (EC-27665, R305) i)
access to radiologically controlled areas is proceduralized to control potential radiological exposures, j) radiological instrumentation is provided and maintained to assess potential exposure, (DRN 99-2362, R11) k)
incoming shipments of radioactive material are received and surveyed properly, l) outgoing shipments of radioactive material are packaged, surveyed, and labeled properly, and m) necessary measures are taken and guidelines followed to keep exposures and effluents ALARA while safely supplying a reliable source of power to the public.
WSES-FSAR-UNIT-3 12.5-3 Revision 305 (11/11) 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 12.5.2.1 Health Physics Facilities (EC-27665, R305)
Access control to the Radiologically Controlled Area (RCA) may be through the Westside Facility or the
-4 elevation of the Reactor Auxiliary Building. The preferred entry point is the Westside Access. Other control points are established as necessary. Access control points contain necessary equipment (i.e.,
access and egress terminals, personnel and tool contamination monitors, etc.) to prevent the spread of contamination.
(EC-27665, R305)
Space is reserved for personnel at the -4 elevation control point for health physics/chemistry personnel.
It provides for equipment, records and supply storage. Space for instrumentation issue and storage may be provided near the Westside Access or the -4 Control Point.
The counting room located in the -4 elevation RAB control point area provides facilities for analysis of samples. Portable sample counting instrumentation will be utilized at other control points as necessary.
Limited personnel decontaminations may be performed in the Westside Access. The -4 control point decon facility will be used on a case basis, for example in the event personnel showering is needed.
Contaminated equipment will normally be decontaminated in the decontamination facility located on the
+21.00 ft. west wing area. Other equipment decon locations may be approved by health physics. The decontamination facility is described in Subsection 12.1.2.
(EC-27665, R305)
Personnel contamination monitors, tool and personnel contamination friskers are located at the -4 and Westside access control points. All personnel exiting the RCA will be monitored. Respiratory protection equipment is available and will be maintained, inspected, and used in accordance with Regulatory Guide 8.15, October 1976. Portal monitors are also located at the Primary Access Point in the administration building.
(EC-27665, R305)
(DRN 99-2362, R11)
Other control points inside the plant may be established as applicable to ensure positive radiation control and provide protective equipment and supplies.
(DRN 99-2362, R11)
(DRN 99-1051, R11; 03-1429, R13)
Storage space in the Radioactive Material Storage Building (RMSB) is provided to allow the storage of reusable radioactive material such as tools, previously contaminated PCs, etc.
(DRN 99-1051, R11)
(DRN 99-1697, R11-A)
The Radioactive Material Storage Building (RMSB) is a pre-engineered, prefabricated sheet metal building. The entire building is constructed on a reinforced concrete slab with curbing such that water is kept from entering the building. The RMSB facility will be used for storage of radiation protection consumables and for storage and maintenance of refueling tools.
(DRN 99-1697, R11-A; 03-1429, R13) 12.5.2.2 Health Physics Instrumentation 12.5.2.2.1 Laboratory Instrumentation Laboratory instrumentation allows plant personnel to ascertain the radioactive material present in survey samples. Typical samples would be contaminated survey smears, airborne survey filter, and charcoal cartridges; tritium surveys and other samples may be processed using radiochemistry counting room
WSES-FSAR-UNIT-3 12.5-4 Revision 305 (11/11) equipment. The health physics counting instrumentation is listed in Table 12.5-1. Instrumentation with equivalent or better sensitivity may be used in lieu of those listed. The criteria for selection of these various counters was to obtain instrumentation that could reliably and quickly count samples; could provide the necessary low backgrounds and sensitivities and that could, to some extent, analyze the counting data to provide information in a more easily used form. Each laboratory counting system is checked and calibrated at regular intervals with standard radioactive sources traceable to a National Institute of Standards & Technology (NIST) source. Counting efficiency, background count rates, and high voltage settings are checked by plant personnel in accordance with plant procedures.
12.5.2.2.2 Portable Survey Instrumentation (DRN 99-2362, R11)
Portable survey instruments are located near the access control points and other control points inside the plant. This equipment will allow plant personnel to perform alpha, beta, gamma, and neutron surveys for radiation, airborne, and surface contamination control.
(DRN 99-2362, R11)
The criteria for selection of these instruments was to obtain accurate and reliable instrumentation that could be easily serviced and that would cover the entire spectrum of radiation measurements expected to be made at the station during normal operation, shutdown, and accident conditions.
(EC-27665, R305)
Each portable survey instrument will be calibrated, when in use. Dose rate meters will be calibrated using a multi source gamma calibrator as described in section 12.5.2.2.4 or in accordance with the instrument manufactures approved calibration method. Instruments will be source checked to verify proper operation in accordance with plant procedures. Sufficient quantities of each type of instrument will be available to permit calibration, maintenance, and repair without causing a shortage in operational instrumentation. Portable radiological survey instrumentation is listed in Table 12.5-2. Instrumentation with equivalent or better sensitivity may be used in lieu of those listed.
(EC-27665, R305) 12.5.2.2.3 Personnel Monitoring Instruments (DRN 06-625, R15; EC-14865, R303, EC-27665, R305)
Personnel monitoring will be provided by use of dosimeters of legal record (DLRs), direct-reading pocket dosimeters and survey instrumentation. The criteria for selection of the DLR for a dose measuring device was to have a device accepted as a legal record and that could be evaluated within a reasonable time by dosimetry personnel. The criteria for selection of (direct reading dosimeters) was to have a device that could be read immediately by exposed personnel. All personnel, with the exception of visitors that do not enter a high radiation, contaminated or airborne radioactivity area, entering the Radiologically Control Area (RCA) will be issued DLR badges which will be used to measure exposure to beta-gamma and Neutron radiation, or to measure only beta-gamma radiation when neutron radiation is not expected.
These badges contain suitable filters to allow determination of penetrating vs. non-penetrating radiation.
DLRs under normal conditions, will be processed periodically based on the fade test results. This will provide the official record of personnel exposure to beta, neutron and gamma radiation.
(DRN 06-625, R15, EC-27665, R305)
Direct reading dosimeters will be worn by personnel in the Radiologically Control Area (RCA). These dosimeters will provide a day-to-day estimate of personnel exposure.
(EC-14865, R303, EC-27665, R305)
(DRN 06-625, R15, EC-27665, R305)
Direct-reading dosimeters will be tested for proper response. DLR chips will be checked for matched response periodically. Quality control of DLR performance is proceduralized.
(DRN 06-625, R15)
Personnel survey instrumentation will consist of Personnel Contamination Monitors (PCM), G-M countrate meters (contamination friskers), portal monitors, and whole body counting capability. The criteria for selection of external contamination measuring equipment was to have devices available at checkpoints and other areas that could be used to determine the location of contamination (friskers), and to have devices that require minimum action by personnel being checked (PCM's). The criteria for selection of the whole body counting system was to have a system readily available to quickly supply information concerning internal contamination levels.
WSES-FSAR-UNIT-3 12.5-5 Revision 305 (11/11)
These instruments will be calibrated and source checked at a frequency specified by plant procedures.
Personnel monitoring instrumentation is listed in Table 12.5-3. Instrumentation with equivalent or better sensitivity may be used in lieu of those listed.
12.5.2.2.4 Health Physics Equipment (EC-27665, R305)
Portable air samplers are used to survey airborne radioactive material concentrations. Mass flow calibrations for air samples are performed when in use. Surveys may be performed for radioactive particulate and radioiodine airborne concentrations. Portable continuous air monitors may be used to monitor airborne concentrations at specific work locations. Local indication will be provided as well as trend information. Alarm setpoints are variable and visual, and audible alarms are provided.
(DRN 99-2362, R11)
Respiratory protection equipment must be available and can be easily accessed. Self-contained breathing apparatus are available at the emergency equipment storage lockers. Equipment will be maintained in accordance with Regulatory Guide 8.15, October 1976.
(DRN 99-2362, R11; EC-27665, R305)
An instrument calibrator will be used for calibrating gamma dose rate instrumentation. This will be a self-contained, heavily shielded, multiple source calibrator or an open air calibration source. Beta and alpha radiation sources will also be available for instrument calibration. Sources are traceable to a NIST source. Neutron sources will be available for proper response checks with actual calibration of neutron instrumentation performed by an outside vendor.
Protective clothing will be supplied for personnel working in contamination areas. The clothing required for a particular instance will be prescribed by health physics personnel on a radiation work permit (see Subsection 12-5.3.4), based on actual or potential radiological conditions.
An adequate inventory of protective clothing will be maintained on hand as necessary to support plant operations and maintenance activities.
Additional contamination control consumables will be available to assist in identifying, deconning and barricading contamination areas.
A listing of health physics equipment is shown in Table 12.5-1 12.5.2.2.5 Other Health Physics Instrumentation The Area Radiation Monitoring System will be installed in areas where it is desirable to have constant dose rate information. Monitors will indicate dose rate locally and/or in the main control room. Fixed continuous airborne radioactivity monitors are also provided at strategic locations where personnel exposure to airborne radionuclides is likely. More information on these fixed instruments is given in Subsection 12.3.4.
WSES-FSAR-UNIT-3 12.5-6 Revision 309 (06/16) 12.5.3 PROCEDURES (DRN 99-2362, R11)
Procedures will be developed to cover all necessary areas of plant operations and maintenance activities and to control potential exposures. ALARA considerations will be embodied in applicable procedures, as Section 12.1 herein describes. In addition, certain methods that will be proceduralized to maintain radiological control over plant operations and maintenance activities are discussed below.
(DRN 99-2362, R11)
The Waterford 3 commitment to regulatory guides will be incorporated into procedures as appropriate.
12.5.3.1 Radiation Surveys Health physics and health physics qualified personnel normally perform radiation surveys, the techniques are delineated in plant health physics procedures. Surveys are performed on frequencies that vary with the potential radiological hazards associated with a given area. Frequencies are also delineated in plant health physics procedures.
(LBDCR 16-016, R309)
Surveys are normally performed to ascertain radiation/contamination levels and for airborne radionuclide concentration determination. Records of all surveys are maintained. Current survey information for some areas within the Radiologically Controlled area are normally posted. Survey information is factored into exposure stay time determination and radiation work permit specifications (see Subsection 12.5.3.4).
(LBDCR 16-016, R309)
Radiation surveys may be performed for gamma, beta, and neutron exposure. Contamination surveys are normally performed to establish gross beta-gamma contamination level, but may be processed for specific types of radiation (beta-alpha-gamma) or specific radionuclides (via gamma spectroscopy). Air samples are normally taken to establish airborne concentrations of particulates and/or radioiodine, but specific nuclide information may also be obtained. Availability of current survey information will aid in keeping exposures ALARA.
12.5.3.2 Additional Methods To Maintain Exposures ALARA The Waterford 3 ALARA policy is delineated in Section 12.1. ALARA considerations are incorporated into various plant and health physics procedures. In addition, various methods are used to maintain exposures ALARA.
12.5.3.2.1 Refueling Some examples of methods of maintaining exposures ALARA during refueling are:
a)
Refueling cavity water is filtered to remove radioactive material.
b)
Refueling cavity water is maintained 140F, and surface ventilation is provided to minimize airborne radioactive material.
c)
Prior to removing the vessel head, the primary system is degassed and sampled to minimize expected airborne levels when the head is removed.
WSES-FSAR-UNIT-3 12.5-7 Revision 305 (11/11) d)
Movement of irradiated fuel assemblies will be accomplished with the assembly maintained under water.
(EC-27665, R305) e)
Work performed in the RCA is staged, i.e., workers are briefed on assignments and familiarized with procedures and equipment needed to complete assignments.
(EC-27665, R305) f)
Current survey information is used.
g)
Ventilation is provided to minimize airborne radioactive material.
h)
The radiation work permit system is used to maintain positive radiological control over work in progress.
12.5.3.2.2 In-service Inspection Some examples of methods of maintaining exposures ALARA during in-service inspections are:
a)
Equipment is calibrated and checked prior to entry into the radiation area.
b)
Portable shielding is used where practicable.
(EC-27665, R305) c)
Work performed in the RCA is staged, i.e., workers are briefed on assignments and familiarized with procedures and equipment needed to complete assignments.
(EC-27665, R305) d)
Current survey information is used.
e)
Ventilation is provided to minimize airborne radioactive material.
f)
The radiation work permit system is used to maintain positive radiological control over work in progress.
12.5.3.2.3 Radwaste Handling Some examples of methods of maintaining exposures ALARA during radwaste handling are:
a)
The volume of radwaste generation has been minimized by station design.
b)
Radwaste systems are heavily shielded and remotely located so that operator and other personnel exposure is minimized.
c)
The spent resin collection and transfer system has been redesigned and modified to further reduce personnel exposure during normal operations.
d)
Filter changeout will utilize a remotely operated extraction device.
e)
Portable shielding is available for use when necessary.
WSES-FSAR-UNIT-3 12.5-8 Revision 306 (05/12) f)
Ventilation is provided, where appropriate, to minimize airborne radioactive material during waste handling operations.
g)
Extension reach rods will be used, where appropriate for hard to reach valves in high radiation areas.
h)
Administrative controls will be instituted in radioactive material storage areas to maximize the use of self shielding effects for packaged low-level waste containers such that personnel exposures will be minimized.
i)
Current radiological survey information will be used.
(EC-14275, R306) 12.5.3.2.4 Spent Fuel Handling, Loading, and Unloading (EC-14275, R306)
Some examples of methods of maintaining exposures ALARA during spent fuel handling are:
a)
The spent fuel pool water is filtered to remove radioactive material.
b)
The spent fuel pool water is cooled and surface air ventilation is provided to minimize airborne radioactive material.
(EC-14275, R306) c)
Loading of the canister in the transfer cask is performed under water.
d)
Fuel handling cranes and extension tools are used to handle transfer casks, fuel assemblies, and inserts.
(EC-14275, R306) e)
Movement of irradiated fuel assemblies not contained in a transfer cask will be accomplished with the assembly maintained under water.
(EC-27665, R305) f)
Work performed in the RCA is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.
(EC-27665, R305) g)
Current survey information is used.
h)
The radiation work permit system is used to maintain positive radiological control over work in progress.
i)
Ventilation is provided to minimize airborne radioactive material.
(EC-14275, R306) j)
After the transfer cask is loaded, it is decontaminated using a pressurized water washing device.
k)
Use of temporary shielding.
(EC-14275, R306) 12.5.3.2.5 Normal Operation Some examples of methods of maintaining exposures ALARA during normal operation are:
a)
The station is designed so that significant radiation sources are minimized and shielded.
WSES-FSAR-UNIT-3 12.5-9 Revision 305 (11/11) b)
An area radiation monitoring system is available and provides indication of radiation levels and, as applicable, local and/or remote alarms.
(EC-27665, R305) c)
Work performed in the RCA is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.
(EC-27665, R305) d)
Current survey information is used.
e)
Ventilation is provided to minimize airborne radioactive material.
f)
The radiation work permit system is used to maintain positive radiological control over work in progress.
g)
During initial start-up, neutron and gamma dose rate surveys are performed to verify shielding adequacy.
h)
Areas are conspicuously posted in accordance with 10CFR20.
(DRN 99-2362, R11) i)
Standby low radiation areas are designated for ALARA purpose to minimize the radiation workers exposure.
(DRN 99-2362, R11) 12.5.3.2.6 Maintenance Some examples of methods of maintaining exposures ALARA during maintenance are:
a)
Equipment is moved to areas with lower radiation and contamination levels for maintenance when practicable.
b)
Extension tools are used when practical.
c)
Portable shielding is used as practical.
(EC-27665, R305) d)
Work performed in the RCA is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.
(EC-27665, R305) e)
Current survey information is used.
f)
The radiation work permit system is used to maintain positive radiological control over work in progress.
g)
Routine maintenance is proceduralized and precautions specified.
h)
Required tools are specifically listed in procedures where practical.
12.5.3.2.7 Sampling Some examples of methods of maintaining exposures ALARA during sampling are:
a)
Sampling hoods are provided in the radiochemistry laboratory. Ventilation minimizes airborne radioactive material. The sampling hoods are located to reduce the exposure from sampling of radioactive liquids and gases.
WSES-FSAR-UNIT-3 12.5-10 Revision 305 (11/11) b)
Procedures specify proper sampling methods.
c)
Radiation levels of samples are checked.
d)
Extension tools are used when practicable.
e)
The Radiation Work Permit System is used to maintain positive radiological control over work in progress.
12.5.3.2.8 Calibration Some examples of methods of maintaining exposures ALARA during calibration are:
a)
The instrument calibrator is heavily shielded.
b)
An interlock is provided so that the calibrator door cannot be opened while sources are exposed.
c)
Portable sources of a significant hazard used to calibrate fixed instruments are transported and maintained in shielded containers.
d)
The radiation work permit system is used, where applicable, to maintain positive radiological control over calibration.
12.5.3.3 Access Control (EC-27665, R305)
Access to the RCA may be through the Westside Access or the -4 RAB control point. These control points provide positive access control over personnel entering controlled areas.
(EC-27665, R305)
High radiation areas (as defined in Technical Specifications) also have access control features. Controls will be established (barricades, flashing lights, signs, etc.). Key control for such areas is delineated in administrative procedures. Access control to high radiation areas is provided through administrative and physical control features as delineated through plant procedures. These procedures adequately address the section of Technical Specifications concerning high radiation areas. Key control for such areas is delineated in administrative procedures.
12.5.3.4 Radiation Work Permit All work in radiation, high radiation and other radiological areas (as determined by health physics) requires a Radiation Work Permit. The Radiation Work Permit establishes the minimum radiological requirements for tasks to be performed safely and efficiently. Violations of the Radiaiton Work Permits instructions should be documented in accordance with station administrative procedures.
WSES-FSAR-UNIT-3 12.5-11 Revision 310 (12/17) 12.5.3.5 Contamination Control Contamination limits for personnel, equipment, and areas are delineated in plant procedures. Surveys for contamination control are performed on a routine basis at various locations in the plant. Non-routine surveys are performed in areas whenever a change in contamination levels is likely and may be important for radiation protection. Areas found contaminated are posted, isolated (with ropes, barriers, etc.) and decontaminated as practical. Since the complete removal of surface contamination from some plant areas is not practical, these areas may be designated as contaminated areas. The level of contamination and number of such areas is minimized. Entrance to such an area normally requires authorization of, and adherence to the specifications of a radiation work permit.
(EC-27665, R305)
Tools and equipment used in contaminated areas are surveyed for removable contamination and contaminated tools are bagged for transportation. If tools or equipment do not meet the clean area limits, they are decontaminated before leaving the RCA or released for restricted use only. Some tools and equipment are for use only in a controlled area. These items are surveyed and decontaminated as appropriate.
(EC-27665, R305)
Personnel are protected from contamination by the protective clothing and equipment specified in radiation work permits. Personnel survey themselves for contamination upon exiting a contamination area (if practical). In addition, when personnel pass through the access control point (see Subsection 12.5.2), they pass through a Personnel Contamination Monitor. Contaminated personnel are decontaminated at the decontamination facility.
12.5.3.6 Radiation Protection Training (LBDCR 16-062, R310)
Plant personnel, both permanent and temporary, whose duties require such training, will be instructed in the fundamentals of radiation protection. Radiation protection training will be given as part of general employee retraining. Training is commensurate with the degree of hazard associated with personnel work assignments. Personnel must be acceptably cognizant of fundamentals presented in training to enter the radiologically controlled areas unescorted.
(LBDCR 16-062, R 310)
Training topics will include: instructions in applicable station and NRC exposure limits, station procedures, instructions to women concerning prenatal exposure, properties of radiation and radioactivity, biological effects of exposure, techniques of radiation protection, ALARA, emergency and fire alarm response, and other topics as pertinent. More detail on the Waterford 3 training program is given in Section 13.2.
Additional training is given to plant personnel whose duties involve greater degrees of radiological hazard, such as health physics personnel and operators.
12.5.3.7 Personnel Monitoring 12.5.3.7.1 External Radiation Exposure (DRN 06-625, R15, EC-27665, R305)
All personnel with the exception of visitors that do not enter a high radiation, contaminated or airborne radioactivity area, who enter the RCA will wear a DLR badge and a direct-reading dosimeter.
(DRN 06-625, R15, (EC-27665, R305)
WSES-FSAR-UNIT-3 12.5-12 Revision 305 (11/11)
(DRN 06-625, R15)
Any individual or group of individuals who enter high radiation areas shall wear DLR badges and a direct-reading dosimeter and shall be provided with or accompanied by one or more of the following:
(DRN 06-625, R15) a)
A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b)
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.
c)
A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveys.
(DRN 06-625, R15, EC-27665, R305)
DLR badges are processed periodically. A permanent exposure record is kept for all badged personnel, in accordance with Regulatory Guide 8.7, June, 1992, and 10CFR20.2106. Direct-reading dosimeters provide a daily estimate of personnel exposure.
(DRN 06-625, R15; EC-27665, R305)
(EC-14865, R303)
If a high dosimeter reading indicates potential unexpected exposure or overexposure, the individual's DLR badge may be processed to verify exposure.
(EC-14865, R303) 12.5.3.7.2 Internal Exposure (DRN 03-1135, R13)
A bioassay program will be performed in accordance with Regulatory Guide 8.9, and the portions of ANSI N343-1978 directly applicable to nuclear power plants. The Health Physics procedures are the vehicle by which the bioassay program is implemented. All personnel who may regularly enter an airborne radioactivity area and any other area where unencapsulated radioactive material is present in a form and quantity such that the area has a significant potential for becoming an airborne radioactivity area will be included in the bioassay program. The need for bioassays for some individuals will be based on the Radiation Protection Manager's judgement as to the probability and potential magnitude of internal exposure. An excreta bioassay will be performed as deemed appropriate by the Radiation Protection Manager. Internal exposure assessments and results will be recorded in accordance with Regulatory Guide 8.7, June, 1992.
(DRN 03-1135, R13) 12.5.3.8 Airborne Radionuclide Control, Assessment, and Personnel The plant ventilation systems (refer to Subsection 6.5.1 and Section 9.4) provides the means for removing airborne radioactive material from the in-plant atmosphere. Airborne radionuclide concentrations are controlled by minimizing loose surface contamination levels and providing containment of sources.
WSES-FSAR-UNIT-3 12.5-13 Revision 305 (11/11)
Concentrations of airborne radionuclides are routinely assessed by fixed and portable continuous air monitors and air sample surveys. Air sample surveys are taken routinely at specified frequencies and nonroutinely when the potential for personnel exposure exists (as determined by health physics personnel).
Radiation work permits may specify air sampling prior to the start of work in a given area. Continuous air monitors alarm when airborne radio-nuclide concentrations exceed preset values in a given area.
Internal exposures are minimized by this assessment and follow up control.
There may exist areas in which airborne radionuclide concentrations cannot be maintained below applicable station limits (normally, these are the Derived Air Concentration limits, as found in 10CFR20, Appendix B). Controls are established in plant procedures to maintain exposures ALARA if personnel entry into those areas is required.
(EC-27665, R305)
Respiratory protection equipment is available and can be easily accessed. Equipment will be maintained, inspected and used in accordance with Regulatory Guide 8.15, October 1976.
(EC-27665, R305)
To assure an adequate program for respiratory protection, the following controls are incorporated into the program:
a)
Each respirator user is advised that he may leave an airborne radioactivity area for psychological or physical relief from respirator use. Each user must leave the area in the case of respirator malfunction or any other condition that might cause reduction in the protection afforded the user.
b)
Sufficient air samples and surveys are made to identify the various radionuclides present and to estimate the individual exposures so that selection of appropriate respiratory equipment can be made in accordance with 10CFR20.1703.
c)
Procedures are established to assure correct fitting, use, maintenance, and cleaning of the various types of respiratory equipment.
d)
Bioassays, will be performed in accordance with plant procedures and, as required, to evaluate individual internal intake of radionuclides and to assess the overall effectiveness of the respiratory protection program.
12.5.3.9 Radioactive Material Safety Program Radioactive material may be used by station personnel for calibration and other purposes. This will include both sealed sources and unsealed materials (gaseous, liquid, or solid). Calibration of radiochemistry counting, fixed monitoring, and portable survey instrumentation is the most common use of such material. Exempt quantities or exempt concentrations of radioactive material do not require special handling for radiation protection purposes.
WSES-FSAR-UNIT-3 12.5-14 Revision 11 (05/01)
Recognized methods for the safe handling of radioactive materials, such as those recommended by the National Council of Radiation Protection and Measurement, will be proceduralized to ensure proper usage.
Procedures specify handling techniques, storage, and other safety considerations, as listed below:
a) proper labeling of all radioactive material (per 10CFR20),
b) inventorying of licensed sealed radioactive sources in accordance with plant procedures, c) leak testing of sealed sources at six month intervals in accordance with license conditions, and d) monitoring of all packages received containing radioactive material in accordance with 10CFR20.1906.
WSES-FSAR-UNIT-3 TABLE 12.5-1 Revision 305 (11/11)
COUNTING INSTRUMENTATION Instrument Sensitivity Range Quantity Remarks (DRN 99-1034, R11, EC-27665, R305)
Alpha Counter
- 0.4 pCi 0-107cpm 1
For contamination levels on survey samples.
(EC-27665, R305)
GM Counter 200 pCi 0-105cpm 2
May be portable and used at inplant control points.
Liquid Scintil-lation Counter
~1x10-6Ci/ml 0-107cpm 1
For low energy counting.
Maintenance by Chemistry Department and kept in chemistry counting area.
(EC-27665, R305)
(EC-27665, R305)
Ge (Li) Detector Particulate
~1x10-11Ci/ml liquid
~1x10-7Ci/ml N/A 1
With associated electronics and spectrum-stripping computer for isotopic analysis.
(EC-27665, R305)
(EC-27665, R305)
(EC-27665, R305)
Note: Instrument accuracies, ranges, and quantities may vary depending upon station and ERO specific needs. Radiological instrumentation that is equivalent or better may be substituted to meet the specific monitoring function.
(DRN 99-1034, R11, EC-27665, R305)
WSES-FSAR-UNIT-3 TABLE 12.5-2 Revision 305 (11/11)
PORTABLE RADIOLOGICAL SURVEY INSTRUMENTATION Instrument Accuracy Range Quantity Remarks (DRN 99-1034, R11)
Alpha Survey Meter 10%
0-105 counts 2
Scintillation (EC-27665, R305)
Neutron Survey Meter 15%
0-100 Rem/hr 2
Capable of detecting neutron over the range of
.025 eV to 10 MeV.
(EC-27665, R305)
GM Survey Meter 10%
0-103 Rem/hr 2 E-Plan 3 Normal ops Telescoping probe.
GM Survey Meter 10%
0-200 mRem/hr 2 E-Plan 2 Normal ops Energy Compensated hand held probe.
(EC-27665, R305)
GM Survey Meter 10%
0-105 cpm 20 E-Plan 20 Normal ops Pancake probes.
Ion Chamber Survey Meter 10%
0-5000 mRem/hr 4 E-Plan 10 Normal ops Dose rate air filled chamber capable of detecting both and radiation.
Ion Chamber Survey Meter 10%
0-50,000 mRem/hr 6 E-Plan 10 Normal ops Dose rate air filled chamber capable of detecting both and radiation.
(EC-27665, R305)
Ion Chamber Survey Meter 20%
0-104 Rem/hr 1
High range, remote probe.
Energy compensated.
(DRN 99-1034, R11)
(EC-27665, R305)
Note: Instrument accuracies, ranges, and quantities may vary depending upon station and ERO specific needs. Radiological instrumentation that is equivalent or better may be substituted to meet the specified monitoring function.
(EC-27665, R305)
WSES-FSAR-UNIT-3 TABLE 12.5-3 Revision 309 (06/16)
PERSONNEL MONITORING INSTRUMENTATION Instrument Sensitivity Range Quantity Remarks (DRN 99-1034, R11)
(DRN 99-1034, R11)
Portal Monitors 1.0 ci Cs-137 NA 2
Scintillation type (DRN 99-1034, R11; EC-14865, R303)
Whole Body Counter
.2% of most nuclide ALI 0-several nCi ALI 2
NaI Detector system (EC-14865, R303)
(EC-27665, R305; LBDCR 16-016, R309)
Direct Reading Dosimeters 50mr 0-1500 mr 40 E-Plan (EC-27665, R305; LBDCR 16-016, R309)
Direct Reading Dosimeters 500 mr 0-10,000 mr 10 Normal ops 27 E-Plan (EC-27665, R305)
Direct Reading Dosimeters 10 mr 0-200 mr 300 E-Plan (EC-27665, R305)
Dosimeter Chargers 4
(DRN 99-1034, R11)
(EC-27665, R305)
Note: Instrument accuracies, ranges, and quantities may vary depending upon station and ERO specific needs. Radiological instrumentation that is equivalent or better may be substituted to meet the specified monitoring function.
(EC-27665, R305)
WSES-FSAR-UNIT-3 TABLE 12.5-4 Revision 305 (11/11)
HEALTH PHYSICS EQUIPMENT Equipment Quantity Range Remarks (DRN 99-1034, R11; EC-27665, R305)
High Volume Air Sampler 6 E-Plan 5 Normal ops 1 to 5 ft3/min Used for rapid assessment of airborne levels.
(EC-27665, R305)
Low Volume Air Sampler 1 E-Plan 15 Normal ops 10 to 100 lpm Used for long duration sampling and trending.
(EC-27665, R305)
Air-Purifying Respirators 67 E-Plan 25 Normal ops N/A Full face, negative pressure.
(EC-27665, R305)
Airline Respirators 19 E-Plan N/A Full face constant flow Self-Contained Breathing Apparatus E-Plan 18 TSC 15 Security OSC (10)
Fire Brigade 5 Locker #1 5 Locker #2 5 Locker #3 6 Normal ops 5 Training N/A N/A N/A N/A Full face pressure demand Full face pressure demand Full face pressure demand Full face pressure demand Portable Continuous Air Monitors 1 E-Plan 5 Normal ops 0 - 100,000 cpm Monitoring of work areas.
Instrument Calibrator 1
.002-500 rem/hr Multiple source shielded self-contained calibrator.
(DRN 99-1034, R11)
(EC-27665, R305)
Note: Instrument accuracies, ranges, and quantities may vary depending upon station and ERO specific needs. Radiological instrumentation that is equivalent or better may be substituted to meet the specified monitoring function.
(EC-27665, R305)