ML22278A289
| ML22278A289 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 09/19/2022 |
| From: | Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22278A223 | List:
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| References | |
| Download: ML22278A289 (81) | |
Text
12.3 RADIATION PROTECTION DESIGN FEATURES
12.3.1 FACILITY DESIGN FEATURES
Specific design features for maintaining personnel exposures ALARA are discussed in this section.
These are used in addition to, and in conjunction with, the more generalized design features described in Section 12.1.2.
12.3.1.1 Common Equipment and Component Designs for ALARA
This section describes the design features used for several general classes of equipment and components. Since these classes of equipment are common to many of the plant systems, the features employed in each system to minimize exposures are similar and can be discussed generically by equipment type.
Filters: Filters that accumulate radioactivity are supplied with the means either to backflush/recharge the filter remotely or to perform cartridge replacement with semiremote tools (i.e., long-handled tools). For cartridge filters, adequate space is provided to allow removing, loading, and transporting the cartridge to the solid radwaste storage area.
Demineralizers: Demineralizers in radioactive systems are designed so that spent resin can be remotely transferred to spent resin tanks and fresh resin can be remotely loaded into the demineralizers. The demineralizers and piping can be flushed with condensate or demineralized water to remove any accumulations of resin. The equipment and floor drain demineralizers are typical examples; these are shown on drawings M-62 and M-63, respectively.
Evaporators: System abandoned (Section 11.2.2.1.3).
Pumps: Wherever practicable, pumps in radioactive areas are provided with mechanical seals to reduce seal servicing time. Pumps and associated piping are arranged to provide adequate space and access for maintenance. Small pumps are installed in a manner that allows easy removal if necessary. All pumps in radioactive waste systems are provided with flanged connections for ease of removal. Pump casings have connections for draining the pump prior to maintenance. The use of base plates with drains connected to the floor drain system minimizes the spread of contamination resulting from leakage of continuously operating pumps.
Tanks: Whenever practicable, tanks are provided with sloped bottoms and bottom outlet connections. Overflow lines are lower than vent lines and are directed to the waste collection system to prevent an overflow from spreading contamination within plant structures. For tanks containing radioactive material, the tank and associated discharge piping can be flushed to reduce radiation levels if required for entry into the tank cells. Access is provided for removal, maintenance, or inspection of tank motor agitators and also for cleaning operations involving tank internals. The solid radwaste collection system (drawing M-66 and Figure 12.3-2) provides examples of these features.
Heat Exchangers: Heat exchangers are provided with corrosion-resistant tubes of stainless steel or other suitable materials with tube-to-tube sheet joints usually welded to minimize leakage.
Impact baffles are provided and tube side and shell side velocities are limited to minimize erosive effects. Space is reserved for tube bundle removal, and provisions for flushing are supplied.
CHAPTER 12 12.3-1 REV. 20, SEPTEMBER 2020 LGS UFSAR
Instruments: Instrument devices are located in low radiation zones and away from radiation sources whenever practicable to reduce personnel exposure during maintenance. Primary instrument devices, which for functional reasons are located in high radiation zones, are designed for easy removal to a lower radiation zone for maintenance and calibration. Some instruments located in high radiation zones, such as thermocouples, are provided in duplicate to preclude the need for immediate entry in case of instrument failure and allow maintenance to be performed at a later time when radiation levels may be lower. Transmitters and readout devices are located in low radiation zones. Instrument and sensing lines are provided with flushing capability and are routed to minimize radioactive gas buildup. Backflushing capability exists for reactor vessel sensing lines.
Tanks containing two-phase fluids are fitted with probe-type instruments.
Whenever practicable, diaphragm seals are provided on instrument sensing lines on process piping that contains highly radioactive solids to reduce radiation exposures during servicing of the instrument. Instrument and sensing line connections are typically located in or above the piping midplane to avoid corrosion product buildup.
Valves: To minimize personnel exposures from valve operations and maintenance, motor-operated, air-operated, or other remotely actuated valves are used to the maximum extent practicable where the reduction in operational exposure is greater than the expected increase in maintenance exposure associated with the remote operator.
Whenever practicable, valves are located in valve galleries and are shielded separately from the major components that accumulate radioactivity. Long runs of exposed radioactive piping are minimized in valve galleries. In areas where manual valves are used on frequently operated radioactive process lines, either reach rods or radiation shielding is provided to minimize personnel exposure.
For equipment located in high radiation zones, remote actuators are provided for frequently operated valves associated with system operation. Examples of remote actuators can be seen on drawing M-66. All other valve operations are either infrequent or performed when the equipment is not operating. To the maximum extent practicable, these valves are provided with straight reach rods to allow operators to feel whether or not the valves are tightly closed. Valves with reach rods are installed with their stems located in a horizontal position wherever possible, so that the reach rods are also horizontal but above head level to prevent restriction of access during maintenance.
Provisions are made in many radiation areas to drain adjacent radioactive components when maintenance is required on valves.
Wherever practicable, valves for clean, nonradioactive systems are separated from radioactive sources and are located in readily accessible areas. Vent, drain, and instrument root isolation valves on radioactive systems are located close to the process piping or equipment with which they are associated. This minimizes the lengths of piping carrying process fluids when these valves are closed.
Manually operated valves required for normal operation and shutdown are not located in filter and demineralizer valve compartments.
For large valves (21/2 inches and larger) originally supplied and installed in lines carrying radioactive fluids, a double set of packing with a lantern ring is usually provided. Such valves, except those having a packing gland which constitutes a direct communication path between primary containment atmosphere and the reactor building, are being systematically provided with an
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improved packing arrangement that eliminates use of the second packing set. Power-actuated valves so changed are also being fitted with automatic gland adjustment capability. A stuffing box is also provided with a leak-off connection that may be piped to a drain header. Flow from the leak-off is stopped by a plug or a manual stop valve. Full ported valves are used in systems containing radioactive solids.
Valve designs with minimum internal crevices are used where crud trapping could become a problem, such as in piping carrying spent resin. Globe valves two inches and smaller in radioactive systems are Y-pattern-type to facilitate rodding if plugged.
Piping: The piping in pipe chases is designed for the lifetime of the unit. The number of valves or instruments in the pipe chases has been reduced to the maximum extent practicable. Piping layout is discussed in Section 12.3.1.2.
Floor Drains: Floor drains and sloped floors are provided for each room or cubicle that has serviceable components containing radioactive liquids. Whenever practicable, drain lines are embedded in concrete floors which provide shielding. If a radioactive drain line must pass through a radiation zone lower than that at which it will terminate, proper shielding is provided. Local gas traps or porous seals are not used on radwaste floor drains. Gas traps are provided at the common sump or tank. Wherever practicable, provisions exist for removing plugging in drain piping. Floor drains are designed to handle potential backflooding.
Lighting: Multiple electric lights are provided for each cell or room containing highly radioactive components so that the burnout of a single lamp does not require entry and immediate replacement of the defective lamp; sufficient illumination still remains available. Lighting in a radioactive area is actuated from outside the area and long-life bulbs are used. Section 9.5.3 describes the lighting system.
HVAC: The HVAC system is designed to minimize radioactive buildup and provides for easy access and fast replacement of the filter elements. Filter banks and components are separated from adjacent banks and components. Section 12.3.3 provides additional description of the HVAC radiation protection design features.
Sample Stations : Sample stations for routine sampling of process fluids are located in accessible areas. Shielding is provided at the local sample stations as required to maintain radiation zoning in proximate areas and minimize personnel exposure during sampling. The counting room and laboratory facilities are described in Section 12.5.
Clean Services : Wherever practicable, clean services such as compressed air piping, clean wate r piping, ventilation ducts, and cable trays are not routed through radioactive pipe-ways. In addition, active components of these clean services are located outside high radiation areas wherever possible to minimize any radiation exposure associated with maintenance of clean systems.
12.3.1.2 Common Facility and Layout Designs for ALARA
This section describes the design features used for standard -type plant processes and layout situations for radioactive systems and for potentially radioactive systems. These features are used in conjunction with the general equipment designs described in Section 12.3.1.1 and include the details discussed in the following paragraphs:
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Valve Galleries: Valve galleries are provided with shielded entrances for personnel protection. In many cases, the valve galleries are divided by shielding or distance into subcompartments that service only two or three components so that personnel are only exposed to the valves and piping associated with a few components at any given location. Process isolation valves are located close to wall penetrations. Floor drains are provided to control radioactive leakage, and curbs are supplied as necessary. To facilitate decontamination in valve galleries, concrete surfaces are covered with a smooth surfaced coating that allows easy decontamination.
Piping: Each piping run is analyzed to determine the potential radioactivity level and maximum expected surface dose rate. Radioactive pipes are routed separately from nonradioactive pipes to minimize personnel exposure. Pipes carrying radioactive materials are routed through controlled access areas zoned for a corresponding level of activity. Where radioactive piping must be routed through corridors or other low radiation areas, shielded pipe-ways are provided. Valves and instruments are not normally placed in radioactive pipe-ways. Wherever practicable, each equipment compartment is used as a pipe-way only for those pipes associated with equipment in the compartment. Doing so minimizes exposure due to the operation of one system while maintenance is being performed on another system that is shut down.
Piping is designed to minimize low points and dead legs. Drains are provided on piping where low points and dead legs cannot be eliminated. Where possible, thermal expansion loops are raised rather than dropped. In radioactive systems, the use of nonremovable backing rings in the piping joints is minimized to eliminate a potential crud trap for radioactive materials. Wherever possible, branch lines having little or no flow during normal operation are connected above the horizontal midplane of the main pipe. Line size changes are typically made by eccentric reducers. Orifices are placed in vertical lines wherever possible.
Piping carrying resin slurries is run with large radius bends wherever possible instead of elbows, and horizontal runs are minimized. To prevent possible crud buildup, flow control valves and orifices are not normally used unless they are required for system operation. Large diameter piping is typically used with a minimum number of pipe fittings to reduce crud accumulation.
Field Run Piping: All routing of radioactive process piping, large and small, is reviewed by the engineering office to ensure that the radiation zone routing is proper and that the above principles are being employed.
Penetrations: Penetrations are normally located with an offset between the source and the accessible areas to minimize radiation streaming. If offsets are not practicable, penetrations are located as far as possible above the floor elevation to reduce the exposure to personnel. If neither of these two methods is used, then alternative means are employed, such as using baffle shield walls or radiation shielding in the area around the penetration.
Contamination Control: Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. Equipment vents and drains from radioactive systems are normally piped directly to the collection system instead of allowing any contaminated fluid to flow across to the floor drain. All-welded piping systems are used for radioactive systems to the maximum extent practicable to reduce system leakage and crud buildup at joints. The valves in some radioactive systems are provided with leak-off connections piped directly to the collection system.
Decontamination of potentially contaminated areas within the plant is facilitated by the application of suitable smooth surfaced coatings to the concrete floors and walls.
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Floor drains and sloping floors are provided in all potentially radioactive areas of the plant. In addition, radioactive and potentially radioactive drains are separated from nonradioactive drains.
Systems that become highly radioactive, such as the radwaste slurry transport system, are provided with flush and drain connections. Certain systems have provisions for chemical cleaning prior to maintenance.
Equipment Layout: In systems where process equipment is a major radiation source (such as fue l pool cleanup, radwaste, condensate demineralizer, etc.), pumps, valves, and instruments are separated from the process component. This allows servicing and maintenance of items in reduced radiation zones. Control panels are located in the lowest dose r adiation zones.
Redundant equipment can be separated from each other, and shielding is provided between the equipment to allow maintenance concurrent with system operation.
Except for HVAC components, major components (such as tanks, demineralizers, and filters) in radioactive systems are located in individual shielded compartments. For highly radioactive components (such as filters and demineralizers), completely enclosed shielded compartments with hatch openings are provided. Provision is made for some major plant components for removal to lower dose radiation zones for maintenance. Large HVAC filter plenums with multiple filter cartridges are individually shielded.
Labyrinth entrance -way shields or shielding doors are provided for each compartment or plenum from which radiation could stream to access areas and exceed the radiation dose limits for those areas. Adequate space for removal of components is provided.
Wherever practicable, lubrication of equipment in radiation areas is achieved with the us e of tube-type extensions to reduce exposure during maintenance.
Figures 12.3 -1 to 12.3-7 provide typical layout arrangements for demineralizers, liquid and particulate filters, waste sludge tanks, offgas recombiners, sample stations, charcoal beds, and their associated valve compartments or galleries.
Exposure from routine in -plant inspection is controlled by locating inspection points in shielded low background radiation areas wherever possible. Radioactive and nonradioactive systems are normally separ ated to limit radiation exposure from routine inspection of nonradioactive systems.
For radioactive systems, emphasis is placed on adequate space and ease of motion in a shielded inspection area. Where longer times for routine inspection are required and permanent shielding is not feasible, sufficient space for portable shielding is normally provided. In high radiation areas where routine surveillance is required, remote viewing devices are provided as needed. Typically, equipment and valves in high rad iation areas are made easily accessible by providing permanent access platforms, easily removable insulation, etc. Equipment man -ways are readily accessible.
Equipment lay -down area requirements are considered in the layout, and adequate space is provided where necessary.
Facilities for Handling Sealed and Unsealed Radioactive Material : As discussed in Section 12.2.1.8, special materials used in the radiochemistry laboratory require the design of special handling equipment. For unsealed materials, the f ollowing are provided:
- a. Exhaust hoods that exhaust to the ventilation system are located in areas such as sample stations and the radiochemistry laboratory.
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- b. Decontamination facilities, the radiochemical laboratory, controlled zone shop, and instrument repair shops are situated at various locations in the plant and are described in Section 12.5.
- c. An area for the repair and maintenance of removed CRDs is provided in the reactor enclosure in close proximity to the CRD removal hatch.
12.3.1.3 Radiation Zoning and Access Control
Access to areas inside the plant structures and plant yards is regulated and controlled. Each high radiation area (as defined in 10CFR20) is provided with a personnel alert barrier. Each high radiation area with a dose rate greater than 1 Rem/hour and very high radiation areas (as defined in 10CFR20) require additional controls per Health Physics Supervision. Section 12.5 describes the control of ingress and egress of plant operating personnel to controlled access areas and the procedures employed to ensure that personnel exposure is within the limits prescribed by 10CFR20.
All plant areas are categorized into radiation zones according to expected radiation levels. Each radiation zone defines either the highest component dose rate in the area or the radiation level to which the aggregate of all contributing sources must be attenuated by shielding, whichever is higher. Each room, corridor, and pipe -way of every plant structure is evaluated for potential radiation sources during normal operation, including anticipated operational occurrences and shutdown. The radiation zone categories used, and their descriptions, are given in Table 12.3 -1, and the specific zoning for each plant area is shown in drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130 N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-139, N-140, N-141, N-142, and N-143. All frequently used areas such as corridors are shielded for Zone I or Zone II access.
The locations of airborne radioactivity and area radiation monitors are described in Section 12.3.4.
Unit 1 and common area radiation monitors are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130 N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-139, N-140, N-141, N-142, and N-143, and Table 12.3-7. Unit 2 area radiation monitors are shown on Table 12.3-7.
12.3.1.4 Control of Activated Corrosion Products
To minimize the radiation exposure associated with the deposition of activated corrosion products in reactor coolant and auxiliary systems, the following steps have been taken:
- a. The reactor coolant system consist s mainly of austenitic stainless steel, carbon steel, and low alloy steel components. The nickel content of these materials is low, and it is controlled in accordance with applicable ASME material specifications. A small amount of nickel base material (Inconel 600) is employed in the reactor vessel internal components. Inconel 600 is required where components are attached to the reactor vessel shell, and the coefficient of expansion must match the thermal expansion characteristics of the low alloy vessel steel. Inconel 600 is selected because it provides the proper thermal expansion characteristics and adequate corrosion resistance, and it can be readily fabricated and welded.
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- b. Materials employed in the reactor coolant system are purchased to ASME ma terial specification requirements. No special controls on levels of cobalt impurities are specified.
- c. Hard-facing and wear materials having a high percentage of cobalt are restricted to applications where no satisfactory alternative materials are available.
- d. A high temperature filtration system is not employed in the RWCU. The reasons for this decision include:
- 1. Lack of quantitative data on the removal efficiency for insoluble cobalt by the high temperature filter.
- 2. Uncertainty in the deposition model, including the relative effectiveness of cobalt removal on deposition rate.
- 3. Doubtful cost -effectiveness in an area where other methods under study (such as decontamination) may prove better at reducing dose rates w hile also being more cost -effective.
- e. Items a, b, and c above also apply to valve materials in contact with reactor coolant.
Valve packing materials are selected primarily for their properties in the particular environment.
- f. Sections 12.1.2.2, 12. 3.1.1, and 12.3.1.2 describe the various flushing, draining, testing, and chemical addition connections that have been incorporated into the design of piping and equipment that handle radioactive materials. If decontamination is to be performed, these co nnections are used for that purpose.
- g. The plant is designed with a powdered resin, pressure precoat cleanup system for the primary coolant in the reactor, and a full flow condensate cleanup system for the feedwater. These systems are described in Secti ons 5.4.8 and 10.4.6, respectively.
- h. A chemistry control program to reduce crud buildup has been developed based upon studies performed by GE. This program will be implemented at LGS.
12.3.2 SHIELDING
In this section the bases for radiation shielding and shielding configurations are discussed.
12.3.2.1 Design Objectives
The basic objective of the plant radiation shielding is to reduce personnel exposures, in conjunction with a program of controlled personnel access to and occupancy of radiation a reas, to levels that are within the dose regulations of 10CFR50 and are ALARA within the dose regulations of 10CFR20. Shielding and equipment layout and design are considered in ensuring that exposures are kept ALARA during all anticipated personnel activi ties in all areas of the plant containing radioactive materials.
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The plant conditions considered in radiation shielding design are normal operation including anticipated operational occurrences at full power, and plant shutdown. The shielding design objectives are as follows:
- a. To ensure that radiation exposure to plant operating personnel, contractors, administrators, visitors, and proximate exclusion area boundary occupants are ALARA and within the limits of 10CFR20.
- b. To ensure sufficient personnel access and occupancy time to allow normal anticipated maintenance, inspection, and safety-related operations required for each plant equipment and instrumentation area.
- c. To reduce potential equipment neutron activation and mitigate radiat ion damage to materials.
The control room is sufficiently shielded so that the direct dose plus the inhalation dose (calculated in Chapter 15) in the event of DBAs does not exceed the limits of GDC 19.
12.3.2.2 General Shielding Design
Shielding is provided to attenuate direct radiation through walls and penetrations, and to attenuate scattered radiation to less than the upper limit of the radiation zone for each area shown in drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130 N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143, which show the minimum shielding requirements for all plant areas. Shielding design criteria and shielding design source terms for specific plant equipment are presented in Section 12.2.
The material used for most of the plant shielding is ordinary concrete with a minimum bulk density of 140 lb/ft3. Whenever poured-in-place concrete has been replaced by concrete blocks or other material, design ensures protection on an equivalent shielding basis as determined by the shielding characteristics of the concrete block and associated grout fill material selected. As discussed in Section 1.8, Regulatory Guide 1.69 is not employed for LGS. Further discussion of concrete standards is contained in Section 3.8. Water is used as the primary shield material for areas above the spent fuel transfer and storage areas.
The guidance provided in Regulatory Guide 8.8 was followed for LGS. ALARA design features are discussed in Sections 12.1.2 and 12.3.1. Specific features pertaining to shielding are described below.
12.3.2.3 Shielding Calculational Methods
The shielding thicknesses provided to ensure compliance with plant radiation zoning and to minimize plant personnel exposure are based on maximum equipment activities under the plant operating conditions, as described in Section 12.2. The thickness of each shield wall surrounding radioactive equipment is determined by approximating the actual geometry and physical condition of the source or sources. The isotopic concentrations are converted to gamma ray sources using data from standard References 12.3 -1 through 12.3-5.
The geometric model assumed for shielding evalua tion of pipes, tanks, heat exchangers, filters, demineralizers, and evaporators (abandoned) is a finite cylindrical volume source. In cases where
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radioactivity is deposited on surfaces, such as inside pipe, the source is treated as a cylindrical annulus. Typical computer codes that are used for original design shielding analysis are listed in Table 12.3-2, which includes References 12.3-6 through 12.3-15. Shielding attenuation data used in those codes include gamma mass attenuation coefficients (Reference 12.3-16), gamma buildup factors (Reference 12.3-17), neutron/gamma multigroup cross-sections (Reference 12.3-18), and albedos (Reference 12.3-19). Additional sources of information pertaining to shielding techniques include References 12.3-20 through 12.3-27. This list of publications and computer programs are as applied in the original plant design and are not meant to be all-inclusive.
Shielding design for plant changes use programs and references which are industry standards at the time of the change. Examples of standards include those developed by the American Nuclear Societys ANS-6 Radiation Protection and Shielding Division, and the International Commission on Radiological Protection.
Computer programs used include those distributed by the Oak Ridge National Radiation Safety Information Computational Center, or commercially developed programs implementing established methods and parameters in industry standards.
The shielding thicknesses are selected to reduce the aggregate computed radiation level from all contributing sources below the upper limit of the radiation zone specified for each plant area.
Shielding requirements are evaluated at the point of maximum radiation dose through an y wall.
Therefore, the actual anticipated radiation levels in the greater region of each plant area are below this maximum dose and therefore below the radiation zone upper limit.
Where shielded entry-ways to compartments containing high radiation sources are necessary, labyrinths or mazes are designed so that the scattered dose rate plus the direct dose rate through the shield wall from all contributing sources is below the upper limit of the radiation zone specified for each plant area.
12.3.2.4 Turbine Enclosure Shielding Design
Radiation shielding is provided as necessary for the following systems in the turbine enclosure and control structure, excluding the control room, to ensure that zone access requirements (drawings N-110, N-111, N-112, N-113, N-115, N-125, N-126, N-127, N-128, and N-130) are met for the surrounding areas:
- a. Main steam system
- b. Condensate system
- c. Feedwater system
- d. Air removal system
- e. Sealing steam system
- f. Condensate cleanup system
- g. Gaseous radwaste reco mbiner system
- h. Turbine enclosure HVAC system
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- i. Control room HVAC system.
12.3.2.5 Control Room Shielding Design
Drawings N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-128, N-130 N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-139, N-140, N-141, N-142, N-143 and Figure 12.3-26 represent layout and isometric drawings of the control room, showing its relationship to the reactor enclosure.
The design basis LOCA dictates the shielding requirements for the control room. Shielding is provided to permit access and occupancy of the control room under LOCA conditions with radiation doses limited to 5 rem whole body from all contributing modes of exposure for the duration of the accident, in accordance with GDC 19.
The design basis LOCA is described in Section 15.6.5. The direct radiation from airborne fission products inside the reactor enclosu re contributes less than 50 mRem to personnel inside the control room for the 30 day period following a LOCA, based on the radioactivity sources described in Section 12.2.
The assumptions used to determine control room habitability are listed in Regulatory Guide 1.3 and are discussed in Section 15.0.
For isotopes that escape from the drywell to the reactor enclosure, credit is taken for shielding by the internal structures in the reactor enclosure. Shielding credit is taken for the reactor enclosure and control structure walls. For all isotopes that remain within the drywell, shielding credit is taken for the drywell wall.
12.3.2.6 Reactor Enclosure and Refueling Area Shielding Design
During reactor operation, the steel-lined, reinforced concrete drywell wall and the reactor enclosure walls protect personnel occupying adjacent plant structures and yard areas from radiation originating in the reactor vessel and in associated equipment within the reactor enclosure. The reactor vessel shield wall, drywell wall, and various equipment compartment walls, together with the reactor enclosure walls, reduce the radiation levels in the yard area outside the reactor enclosure to less than Zone I maximum dose rates, for those areas that are normally accessible.
Where personnel and equipment removal hatch openings or penetrations pass through the drywell wall, additional shielding, such as labyrinths or doors, is provided to attenuate the radiation to below the required level defined by the radiation zone outside the drywell wall.
Inside Drywell Structure: Areas within the drywell are Zone V and are normally inaccessible during plant operation. The reactor vessel shield provides shielding for access in the drywell during shutdown and reduces the activation of and radiation damage to drywell equipment and materials.
Outside Drywell Structure : The drywell wall reduces radiation levels in accessible areas of the reactor enclosure from sources within the drywell to below the maximum level for Zone II. Shielding requirements for the personnel access and equipment removal and CRD removal hatch openings are shown on drawings N-119 and N-134 in the areas numbered 401, 405, and 409, respectively.
Shielding for these areas includes high density concrete, because neutron shielding must be considered for these large openings in the drywell wall. Drywell piping and electrical penetrations are shielded by providing either lock shields within the penetration assembly or a shielded penetration room. Shielded piping penetration room locations and bulk shielding requirements are
CHAPTER 12 12.3-10 REV. 20, SEPTEMBER 2020 LGS UFSAR
shown on drawings N-118, N-119, N-120, N-133, N-134, and N-135. These rooms, numbered 306, 307, 309, 501, 510, 522, 523, and 599 are designated radiation Zone V during reactor power operation and are provided with personnel access controls.
Radiation shielding is provided as necessary for the following systems in the reactor enclosure and refueling area to ensure that the radiation zone and access requirements (drawings N-116, N-122, N-131, N-137, and N-139) are met for surrounding areas:
- a. RWCU system
- b. FPCC system
- c. Neutron monitoring system
- d. HPCI system
- e. RCIC system
- f. RHR system
- g. Core spray system
- h. Reactor enclosure and refueling area HVAC systems.
Main steam lines are located within shielded structures from the drywell wall to the reactor enclosure wall.
Spent fuel is a primary source of radiation during refueling. Because of the extremely high activity of the fission products contained in the spent fuel assemblies and the proximity of Zone II areas, extensive shielding is provided for areas surrounding the fuel transfer canal and pool to ensure that radiation levels remain below zone levels specified for adjacent areas.
After reactor shutdown, RHR system pumps and heat exchangers are in operation to remove heat from the reactor water. The radiation levels in the vicinity of this equipment will temporarily reach Zone V levels due to corrosion and fission products in the reactor water. Shielding is provided to attenuate radiation from RHR equipment during shutdown cooling operations to levels consistent with the radiation zoning requirements of adjacent areas. Adequate shielding is also provided to maintain radiation zoning requirements during the hot standby operation of the RHR system.
The concrete shield walls surrounding the spent fuel cask loading, storage, and transfer areas, as well as the shield walls surrounding the fuel transfer and storage areas, are sufficient to limit radiation levels outside the shield walls in all accessible areas to below Zone II maximum dose rates.
Water in the spent fuel pool provides shielding above the spent fuel transfer and storage areas.
Radiation levels at the fuel handling equipment are calculated to be below Zone II maximum dose rates during normal operations.
Water is also used as shielding material above the steam dryer and separator storage area.
Concrete walls and water in the pool are designed to provide Zone II dose rates in adjacent accessible areas during storage of the dryer and separator.
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A portable refueling shield is provided to reduce radiation dose rates in the drywell that are due to the transfer of spent fuel assemblies from the reactor vessel to the spent fuel pool. During refueling, the lead and steel shield is located in the reactor well, between the reactor vessel and the spent fuel pool, which permits continuous personnel occupancy of the drywell.
In addition, a temporary local personnel alarming rate meter is expected to be used in the drywell, with local alarms. Access controls to the upper reaches of the drywell aid in ensuring satisfactory protection for personnel.
12.3.2.7 Radwaste Enclosure Shielding Design
Radiation shielding is provided as necessary for the following systems in the radwaste enclosure to ensure that the radiation zone and access requirements (drawings N-140 through N-143) are met for surrounding areas:
- a. Liquid radwaste equipment drain subsystem
- b. Liquid radwaste floor drain subsystem
- c. Liquid radwaste chemical waste subsystem
- d. Liquid radwaste laundry drain subsystem
- e. Solid radwaste system
12.3.2.8 Offgas Enclosure Shielding Design
Radiation shielding is provided as necessary for the gaseous radwaste system in the offgas enclosure to ensure that zone access requirements ( drawings N-140 through N-142) are met for surrounding areas.
12.3.2.9 Diesel Generator Enclosure Shielding Design
There are no radiation sources in the diesel generator enclosure; therefore, no radiation shielding is required for the enclosure.
12.3.2.10 Miscellaneous Plant Areas and Plant Yard Areas
Radiation shielding is provided for all radiation sources located in plant enclosures so that radiation levels at accessible areas outside are maintained below Zone I levels. Plant yard areas that are frequently occupied by plant personnel are accessible during normal operation and shutdown.
These areas are surrounded by a security fence and closed off from areas accessible to the general public.
12.3.3 VENTILATION
The plant ventilation system provides a suitable environment for personnel and equipment during normal operation and anticipated operational occurrences. Detailed HVAC system descriptions are provided in Section 9.4. Control room habitability is discussed in Section 6.4.
CHAPTER 12 12.3-12 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.3.3.1 Design Objectives
The systems are designed to operate so that the in-plant airborne activity levels for normal operation (including anticipated operational occurrences) in the general personnel access areas are within the limits of 10CFR20. The systems operate to reduce the spread of airborne radioactivity during normal and anticipated abnormal operating conditions.
During postaccident conditions, the ventilation system for the plant control room provides a suitable environment for personnel and equipment and ensures continuous occupancy in this area. The plant ventilation systems are designed to comply with the airborne radioactivity release limits for offsite areas during normal operation.
12.3.3.2 Design Criteria
Design criteria for the plant HVAC systems include the following:
- a. During normal operation and anticipated operational occurrences, the average and maximum airborne radioactivity levels to which plant personnel are exposed in restricted areas (as defined in pre -1994 10CFR20) of the plant are ALARA and within the limits specified in 10CFR20. The average and maximum airborne radioactivity levels in unrestricted areas (as defined in pre-1994 10CFR20) of the plant during normal operation and anticipated operational occurrences will be ALARA and within the limits of 10CFR20.
- b. During normal operation and anticipated operational occurrences, the dose from concentrations of airborne radioactive material in unrestricted areas beyon d the exclusion area boundary will be ALARA and within the limits specified in 10CFR20 and 10CFR50.
- c. The dose limits of 10CFR50.67 will be satisfied following those hypothetical accidents, described in Chapter 15, that involve a release of radioactivity from the plant.
- d. The dose to control room personnel shall not exceed the limits specified in GDC 19 and 10CFR50.67 following those hypothetical accidents, described in Chapter 15, that involve a release of radioactivity from the plant.
12.3.3.3 Design Guidelines
To accomplish the design objectives, the following guidelines are followed whenever practicable.
12.3.3.3.1 Guidelines to Minimize Airborne Radioactivity
The following design guidelines describe equipment and layout features that minimize the formation of airborne radioactivity:
- a. Equipment vents and drains are piped directly to a collection device connected to the collection system instead of allowing any radioactive fluid to flow across the floor to the floor drain.
CHAPTER 12 12.3-13 REV. 20, SEPTEMBER 2020 LGS UFSAR
- b. All-welded piping systems are used on radioactive systems to the maximum extent practicable to reduce system leakage.
- c. Decontaminatable coatings are applied to the concrete floors and walls of potentially radioactive areas to facilitate decontamination.
- d. Radioactive equipment has design features that minimize the potential for airborne radioactive contamination during maintenance operations. These features include flush connections on pump casings for draining and flushing the pump before maintenance and flush connections on piping systems that could handle radioactive fluids.
- e. Exhaust hoods are used in the laboratories, work areas, and sample stations to facilitate processing of radioactive samples by forcing contaminants away from the personnel breat hing areas and into the ventilation and filtering systems.
- f. Equipment decontamination facilities are ventilated to ensure control of released radioactivity and prevent the spread of radioactive contamination.
- g. The valves in some systems are provide d with leak-off connections piped directly to the collection system.
- h. To minimize the amount of airborne radioactivity that results from valve packing leakage, most larger valves (21/2 inches and larger) are supplied and installed with a double set of pa cking and lantern ring in lines carrying radioactive fluids. The stuffing box is provided with a leak-off connection which is usually plugged, but may be piped through a manual stop valve to a drain header. An improved packing arrangement is being system atically provided for such valves and eliminates use of the second packing set. The exceptions for those valves that have a packing gland, which constitutes a direct communication path between primary containment atmosphere and the reactor building. These valves are being fitted with an improved packing arrangement that uses two packing sets and a lantern ring to maintain packing gland testability. Power -actuated valves, as part of the improved packing arrangement, are fitted with automatic gland adjustm ent capability.
12.3.3.3.2 Guidelines to Control Airborne Radioactivity
- a. The airflow is directed from areas with lesser potential for radioactive contamination to areas with greater potential for radioactive contamination.
- b. In building compartmen ts with a potential for radioactive contamination, a greater volumetric flow is exhausted from the area than is supplied to the area to minimize the amount of uncontrolled exfiltration from the area, or makeup air for the exhaust from the area is infiltrated from surrounding less contaminated areas with no direct supply to the area provided.
- c. Floor and equipment drain collection tank vents are piped to a collection header and processed by the ventilation system.
- d. Air is supplied to each principal en closure via separate supply intakes and duct systems.
CHAPTER 12 12.3-14 REV. 20, SEPTEMBER 2020 LGS UFSAR
- e. Air being discharged from potentially contaminated areas of the reactor and turbine enclosures is passed through prefilters, HEPA filters, and charcoal filters before release (drawings M-75 and M-76). Air being discharged from the radwaste enclosure is passed through prefilters and HEPA filters as shown in Figure 9.4 -3. In addition, to aid in minimizing the quantity of radioactive material discharged, means are provided to isolate the reactor enclosure and the control room upon indication of high airborne activity levels.
- f. Suitable primary containment isolation valves are installed in accordance with GDC 54 and GDC 56, including valve controls, to ensure that the containment integrity is maintained.
- g. Only exhaust duct -work serves potentially contaminated equipment rooms. The duct-work is at negative pressure so that any leakage is directed into the duct.
12.3.3.3.3 Guidelines to Minimize Personnel Exposure from HVAC Equipment
- a. Ventilation ducts are designed to minimize the leakage of radioactive contamination into or out of ducts (as applicable) and the buildup of radioactive contamination within the ducts. Internal obstructions are avoided wherever practicable. Within the reactor enclosure and control structure, welded construction of duct sections is used, and flanged and gasketed joints are used to join duct-work segments.
Seismic class II duct-work in the turbine and radwaste enclosures is constructed to Sheet Metal and Ai r Conditioning Contractors National Association standards modified to allow no more than 4% leakage. All duct -work is 100% leak tested after installation.
- b. Access and service of ventilation systems in potentially radioactive areas are improved by comp onent location to minimize operator exposure during maintenance, inspection, and testing, as follows:
- 1. The outside air supply units and building exhaust system components are enclosed in ventilation equipment rooms. These equipment rooms are located in radiation Zone II areas and are accessible to the operators. Work space is provided around each unit for anticipated maintenance, testing, and inspection. Filter adsorber units comply with the access and service requirements of Regulatory Guide 1.52 a nd Regulatory Guide 1.140, to the extent discussed in Sections 6.5.1 and 9.4, respectively.
- 2. Local cooling equipment servicing the normal building requirements is located in low radiation areas whenever practicable.
- c. All filter systems in which radioactive materials could accumulate to produce significant radiation fields external to the duct-work are appropriately located and shielded to reduce exposure to personnel.
- d. The HVAC system is designed to allow fast replacement of components. To simplify element handling, access to active elements is direct from working platforms. Space is provided on the platforms for accommodating safe personnel
CHAPTER 12 12.3-15 REV. 20, SEPTEMBER 2020 LGS UFSAR
movement during replacement of components, including the use of necessary material handling facilities, and during any inplace testing operations.
- e. The prefilters and HEPA filters are designed with replaceable 2 foot by 2 foot units that are clamped in place against compression seals. The filter housings are designed, tested, and proven to be airtight with bulkhead -type doors that are closed against compression seals. There are two types of charcoal filters: tray -type and rechargeable -type. The tray-type charcoal filter units are designed with replaceable trays clamped in place against compression se als. There are three trays for each 2 foot by 2 foot unit. The rechargeable charcoal filter units are designed with perforated metal beds filled with bulk charcoal. The bulk charcoal in the beds can be drained and refilled.
- f. While most of the activi ty in the filter train is eliminated by simply removing the contaminated filters, further decontamination of the internal structure is facilitated by the proximity of electrical outlets for operation of decontamination equipment, and water supply for washd own of the interior, if necessary. Drains are provided on the filter housing for removal of contaminated water.
- g. Filters in all systems are changed based on the airflow and the pressure drop across the filter bank. Charcoal adsorbers are changed based on the residual adsorption capacity of the bed as measured by the testing of carbon samples taken from the removable canisters located in the carbon bed. The testing of the carbon adsorbers and all other components is described in Sections 6.5 and 9.4.
12.3.3.4 Design Description
Portions of the ventilation systems serving the following enclosures are assumed to be potentially radioactive and are discussed in detail in Section 9.4.
- a. Turbine enclosure
- b. Control structure
- d. Radwaste enclosure
- e. Chemistry Laboratory Expansion
- f. Offgas enclosure
Although the control room is considered a nonradioactive area, radiation protection is provided to ensure habitability (Section 6.4).
Ventilation system des ign parameters are given in Tables 12.3 -3 through 12.3 -6.
A typical layout of a potentially radioactive filter unit is given on Figure 12.3-7.
12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION
CHAPTER 12 12.3-16 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.3.4.1 Area Radiation Monitors
Fixed area radiation monitors are mounted throughout the plant at selected locations. Each location contains a gamma-sensitive detector, local indicator, and local alarm (visual and audio).
Indicators, alarms, and recorders are located in the control room, except for the nine local area monitors (Section 12.3.4.1.3). Alarm setpoints are adjusted from the auxiliary equipment room/control room except for local area monitors.
The location and range for each monitor are given in Table 12.3-7. Ranges specified are consistent with the potential dose rates for a given area. Selection of location is based on the equipment in the area and the need for personnel access. In some locations, the detector is mounted immediately inside a cavity while associated local indicators and alarms are mounted just outside the access door. Personnel are thereby alerted to unusual radiation levels within the cavity before initiating access. Experience at PBAPS Units 2 and 3 was used in determining locations.
Area monitors located at fuel storage areas comply with the requirements of 10CFR70.24. These and area monitors in the radwaste enclosure comply with requirements of GDC 63. The area radiation monitors, with associated alarms, indicators and recorders in the main control room provide the capability to alert supervisors and personnel in that area of abnormally high and unexpected radiation levels. Subsequent notification to appropriate plant personnel could avoid inadvertent unnecessary exposure. Unexpected increases in radiation levels may be due to changes in operations, deposition of crud, or transport of radioactive material (e.g., sources) within the plant. Recorded area radiation monitor readings serve to define trends that may result from build-up, spills, or contamination of a process fluid.
Control and calibration of radiation monitoring (fixed and portable) are provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program described in Section 17.2.
12.3.4.1.1 Design Bases
The purpose of the area radiation monitoring system is to provide personnel protection in accordance with the guidelines of 10CFR20, 10CFR50, 10CFR70, and Regulatory Guide 8.8. The area radiation monitoring system has no function related to the safe shutdown of the plant or to the quantitative monitoring of the release of radioactive material to the environment. Consistent with this purpose, the area radiation monitoring system is designed to provide the following functions:
- a. Warn of excessive gamma radiation levels in areas where nuclear fuel is stored or handled, in accordance with 10CFR70.24(a)(2).
- b. Provide operating personnel with a record and indication in the control room of gamma radiation levels at selected locations within the various plant structures, in accordance with Regulatory Guide 8.8.
- c. Supplement other systems in detecting abnormal migrations of radioactive material
in or from process streams.
- d. Provide local alarms at strategic locations throughout the plant, in accordance with Regulatory Guide 8.8, where a substantial change in radiation levels or loss of sensing capability might be of immediate danger to personnel in the area.
CHAPTER 12 12.3-17 REV. 20, SEPTEMBER 2020 LGS UFSAR
- e. Contribute supervisory information to the control room so that correct decisions can be made with respect to deployment of personnel in the event of a radiation accident.
- f. Assist in the detection of unauthorized or inadvertent movement of radioactive
material in the plant including the radwaste enclosure.
- g. Furnish information for conducting radiation surveys.
- h. Area Radiation Monitors may be removed from service for maintenance, testing, or known evolutions which would cause an ARM to alarm. Additionally an ARM may be removed from service to clear contro l room annunciators so other ARM alarms will not be masked. ARM's may be removed from service if the appropriate compensatory actions, if required, are taken (i.e. HP surveys, access control) when the ARM is removed from service.
The above functions are performed under the following design conditions:
- h. Environmental parameters shown in Table 12.3 -8 are applicable to the design of the area radiation monitoring equipment except for the monitors located inside primary containment, where the detectors are designed for a normal operating temperature range of 64 F to 151F (up to 340 F for accident conditions) and a pressure range of atmospheric to 2 psig.
- i. Noise from any source in the operating environment should not disturb the meter indication by more than +/-2% of equivalent full-scale.
- j. The detector/indicator and trip unit should be responsive to gamma radiation over an energy range of 0.08 MeV to 7 MeV. The energy dependence should not exceed +/-20% of the indicated scale reading for a dose rate of approximately 50 mRem/hr resulting from 0.1 MeV to 3 MeV gammas.
- k. At the control room, the reading should be reproducible within +/-10% of the local indicated point, and drift should not exceed +/-0.2% of equivalent linear full-scale for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or +/-2% for a 30 day period.
- l. The range of the monitors is shown in Table 12.3 -7. The ranges selected ensure readout of both the highest and lowest anticipated radiation levels, in accordance with Regulatory Guide 8.8.
- m. Operate without effective change in performance when the ac supply voltage changes over a range of +/-10% from nominal value or has a frequency variation of
+/-5% from nominal value.
- n. Operate so that interruption or failure of the ac power supply will result in actuation of the trip circuit to produce an alarm, in accordance with Regulatory Guide 8.8.
12.3.4.1.2 Design Details
The area radiation monitoring system is shown in the typical functional block diagram of Figure 12.3-27. Each channel consists of a combined sensor/converter unit, a local auxiliary unit (readout with visual and audible alarm), a combined indicator/trip unit, a shared power supply, and a shared
CHAPTER 12 12.3-18 REV. 20, SEPTEMBER 2020 LGS UFSAR
multipoint recorder. The locations of the radiation sensors are indicated in Table 12.3 -7. Further design details of the area radiation monitoring system are as follows:
- a. Each indicator and trip unit is provided with one upscale trip continuously adjustable over the entire scale and one downscale trip adjustable over the lower decade.
Provision is ma de to permit the upscale trip to be set and checked for accuracy with respect to the indicator.
- b. Detectors are wall -mounted and suitable for operation in the anticipated plant environment with no additional protection.
- c. Panel locations of the remot e equipment are shown in Table 12.3 -9.
- d. The following features are provided for components located in the auxiliary equipment room:
- 1. Radiation level indicator (meter)
- 2. High radiation alarm light (amber)
- 3. Downscale alarm light (white)
- 4. Alarm reset (push button)
- 5. Meter zero adjust (on the amplifier)
- 6. Alarm level adjust
- 7. Trip check push button
- 8. Power supply switch and "power -on" light (clear lens)
- 9. Indicators to show power supply voltages
- 10. Annunciator outputs
- 11. Recorder outputs
- e. The radiation monitors are calibrated at regular time intervals in accordance with station procedures. Calibration methods are covered in detail in the equipment procedures manual.
- f. The following annunci ators are located in the control room to alert the operator:
- 1. Reactor enclosure area, high radiation (Units 1 and 2)
- 2. Refueling floor area, high radiation (Units 1 and 2)
- 3. Turbine enclosure area, high radiation (Units 1 and 2)
- 4. Turbine enclosure common area, high radiation
- 5. Radwaste enclosure common area, high radiation
CHAPTER 12 12.3-19 REV. 20, SEPTEMBER 2020 LGS UFSAR
- 6. Reactor enclosure common area, high radiation
- 7. Admin. bldg. area, high radiation
- 8. Unitized area, Rad monitors downscale
- 9. Common area, Rad monitors downscale
12.3.4.1.3 Local Area Monitors
In addition to the area radiation monitors described above, ten local area monitors are provided, located on each of the two refueling bridges, the three turbine enclosure crane cabs, the east and west unit 1 and 2 turbine enclosure deep bed vessel areas, and the health physics and chemistry source storage and calibration room. The essential differences between these monitors and those area monitors described above are as follows:
- a. No outputs to the control room are provided.
- b. Alarms are local only.
- c. No recorders are provided.
- d. Local power (battery) packs are provided in the event of external power cutoff, except for the local area radiation monitors in the health physics an d chemistry source storage and calibration room, and east and west Unit 1 and 2 turbine enclosure deep bed vessel areas.
The power for the turbine enclosure cranes is turned off except when the cranes are being used or tested. The battery backup for the crane monitors is not available when the crane power is off.
A portable alarming rate meter will be used in the drywell to warn plant personnel if a core component is dropped during fuel transfer operations. The meter will have local alarms to ensure adequate personnel warning. Administrative controls will be used to prevent access to the upper drywell and unnecessary high exposures to personnel.
12.3.4.1.4 Postaccident Area Radiation Monitors
In the event of an accident, access will be required to certain locations outside secondary containment to perform sampling, analysis, and monitoring tasks. The identified areas include the counting room and chemistry laboratory, the main control room, the turbine enclosure near the control room exit, the north stack instrument room, the postaccident sampling station. The operational support center utilizes a portable arm. The area monitors associated with these locations have been designated as postaccident area radiation monitors. In addition to having dose rates available through the shared multipoint recorder in the control room, instantaneous and stored data are available on demand for display and trending in the control room, TSC via the RMMS computer system. Data from PMS is transmitted to the EOF via the emergency plan display system (EPDS). The ranges of these area monitors are specified in Table 12.3 -7 and incorporate the highest anticipated postaccident dose rates for these locations. The design details and operating conditions are as described in Sections 12.3.4.1 and 12.3.4.2. Compliance with Regulatory Guide 1.97 (Rev 2) is discussed in Section 7.5.
CHAPTER 12 12.3-20 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.3.4.2 Airborne Monitoring
Airborne radioactivity monitoring is accomplished by use of the following:
- a. Monitoring ventilation ducts in key locations throughout the plant.
- b. Continuous air monitors which are portable, cart -mounted, and monitor particulate activity.
- c. High and low volume portable samplers capable of attaching filters and charcoal cartridges for particulate and iodine monitoring.
The ventilation system monitors are located at positions which provide representative air concentrations and a rapid indication of abnormal conditions. Those systems which require HEPA filtration have monitors upstream of the filters. Both the in-line GM tube and beta scintillator, and off-line particulate, iodine, and noble gas monitoring configurations can be utilized. Readout and annunciation are provided in the main control room. The off -line monitors provide the capability to obtain particulate and iodine samples for isotopic analysis. Emergency dc power is provided in the event of a LOOP. The detectors are calibrated routinely and after any maintenance work is performed on the detector.
CAMs are located in freely accessible areas wh ere airborne radioactivity is most likely to exist.
These CAMs are mobile and can be moved from area to area as deemed necessary by plant conditions or maintenance operations. CAMs incorporate either fixed or movable filters for the collection of particulate activity, which is monitored directly by a detector. Readout is recorded in counts per minute. The filters can be removed for further analysis using counting room instrumentation. Audible and visual alarms indicate when setpoint levels have been ex ceeded.
The detectors are calibrated routinely and after any maintenance work is performed on the detector.
The CAM's primary function is to indicate trends and sudden changes in airborne activity. Typical locations are solid waste handling areas, spent fuel pool areas, and the reactor operating floor and turbine building. The monitoring system is capable of detecting particulate, iodine and noble gas radioactivity. A flexible hose can be attached to the monitor intake and inserted into a cavity or work area to detect the presence of localized airborne activity. Conformance to Regulatory Guide 8.2 is discussed in Section 12.5.1.
Ventilation monitors and CAMs are used as trending devices and will indicate areas and times needing special samples taken.
Alarm setpoints are set at low levels to ensure close respiratory controls. CAMs, however, cannot account for inversion conditions or properly identify isotopic content of the air. When a setpoint is reached, the monitor alarms. The reason for the alarm is evaluated and when necessary, grab air samples are taken and analyzed in the counting lab. The DAC hours are isotopically calculated by the computer program or are done manually. Appropriate actions can then be taken based on accurate data.
Potentially airborne accessible areas are air sampled at regular intervals. The survey/sampling frequency is designed for that area by the routine survey program.
CHAPTER 12 12.3-21 REV. 20, SEPTEMBER 2020 LGS UFSAR
Low and high volume samples with filter paper or charcoal cartridges are used. These are described in more detail in Section 12.5.3.1.3.
Airborne monitoring provides the information necessary to determine stay times in given areas and applicable respiratory equipment. The information is also of value in identifying process system leakage. Such monitoring is conducted in accordance with the guidelines of Regulatory Guide 1.21.
Control and calibration of radioactivity monitoring (fixed and portable) are provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program described in Section 17.2.
Qualification and training of health physics and chemistry personnel will follow ANSI/ANS 3.1-2014 guidance.
12.
3.5 REFERENCES
12.3-1 J.J. Martin and P.H. Blichert-Toft, "Radioactive Atoms, Auger Electrons, and X-Ray Data," Nuclear Data Tables, Academic Press, (October 1970).
12.3-2 J.J. Martin, "Radioactive Atoms Supplement 1", ORNL 4923, (August 1973).
12.3-3 W.W. Bowman and K.W. MacMurdo, "Radioactive Decays Ordered by Energy and Nuclide," Atomic Data and Nuclear Data Tables, Academic Press, (February 1970).
12.3-4 M.E. Meek and R.S. Gilbert, "Summary of X -Ray and Gamma Ray Energy and Intensity Data," NEDO -12037, (January 1970).
12.3-5 C.M. Lederer, et al., "Table of Isotopes", Lawrence Radiation Laboratory, University of California, (March 1968).
12.3-6 D.S. Duncan and A.B. Spear, "Grace 1 - An IBM 704 -709 Program Design for Computing Gamma Ray Attenuation and Heating in Reactor Shields", Atomics International, (June 1959).
12.3-7 D.S. Duncan and A.B. Spear, "Grace 2 - An IBM 709 Program for Computing Gamma Ray Attenuation and Heating in Cylindrical and Spherical Geometries",
Atomics International, (November 1959).
12.3-8 W.W. Engle, Jr., "A User's Manual for ANISN: A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," Report No. K -1693, Union Carbide Corporation, (1967).
12.3-9 E.D. Arnold and B.F. Maskewitz, "SDC, A Shielding Design Calculation for Fuel Handling Facilities," ORNL -3041, (March 196 6).
12.3-10 R.E. Malenfant, "QAD, A Series of Point-Kernel General Purpose Shielding Programs," LA -3573, Los Alamos Scientific Laboratory, (October 1966).
12.3-11 D.A. Klopp, "NAP - Multigroup Time -Dependent Neutron Activation Predication Code", IITRI-A6088-21, (January 1966).
CHAPTER 12 12.3-22 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.3-12 E.A. Straker, P.N. Stevens, D.C. Irving, and V.R. Cain, "MORSE - A Multigroup Neutron and Gamma Ray Monte Carlo Transport Code", ORNL -4585, (September 1970).
12.3-13 W.A. Rhoades and F.R. Mynatt, "The DOT 3 Two -Dimensional Discrete Ordinates Transport Code".
12.3-14 M.J. Bell, "ORIGEN - The ORNL Generation and Depletion Code," Oak Ridge National Laboratory, ORNL -4628, (May 1973).
12.3-15 R.E. Malenfant, "G -33: A General Purpose Gamma Ray Scattering Program," LA -
5176, Los Alamos Scientific Laboratory, (June 1973).
12.3-16 G.W. Goldstein, "X -ray Attenuation Coefficients from 10 keV to 100 MeV", NBS Circular 583, (April 30, 1957).
12.3-17 D.K. Trubey, "A Survey of Empirical Functions Used to Fit Gamma Ray Buildup Factors," ORNL -RSIC-10, (February 1966).
12.3-18 ORNL RSIC Computer Code Collection DLC -23, "CASK - 40 Group Neutron and Gamma-Ray Cross-Section Data".
12.3-19 W.E. Selph, "Neutron and Gamma Ray Albedos," ORNL -RSIC-21, (February 1968).
12.5-20 "Reactor Physics Constants", Argonne National Laboratory, ANL -5800, (July 1963).
12.3-21 J.F. Kircher and R.E. Bowman, "Effects of Radiation on Materials and Components", (March 1964).
12.3-22 T. Rockwell, "Reactor Shielding Design Manual", D. Van Nostrand Co., Ne w York, (1956).
12.3-23 C.R. Tipton, Jr., "Reactor Handbook", Vol. I, Materials, second edition, (1962).
12.3-24 H. Soodak, "Reactor Handbook", Vol. III, Part A, Physics, second edition, (1962).
12.3-25 E.P. Blizzard and L.S. Abbott, "Reactor Handbook", Vol. III, Part B, Shielding, second edition, (1962).
12.3-26 N.M. Schaeffer, "Reactor Shielding for Nuclear Engineers", TID-25951, (1973).
12.3-27 R.G. Jaegar, et al., "Engineering Compendium on Radiation Shielding, Volume 1:
Shielding Fundam entals and Methods", Springer-Verlag, New York, (1968).
CHAPTER 12 12.3-23 REV. 20, SEPTEMBER 2020 LGS UFSAR
Table 12.3-1
PLANT RADIATION ZONE DESCRIPTION
MAXIMUM DESIGN DOSE RATE DESIGNATION (mRem/hr) DESCRIPTION
I 0.5 No radiation sources, no radiological control required
II 2.5 Low radiation sources, radiological control required
III 15 Low to moderate radiation sources, radiological control required
IV < 100 Moderate radiation sources, radiological control required
V 100 High radiation sources, radiological control required
CHAPTER 12 12.3-24 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-2
LIST OF COMPUTER CODES USED IN SHIELDING DESIGN CALCULATIONS
GRACE 1 Multigroup, multiregion, gamma ray attenuation code used to compute gamma heating and gamma dose rates in slab geometry (Reference 12.3-6).
GRACE 2 Multigroup, multiregion, gamma ray attenuation code used to compute the dose rate or heat generation rate for a spherical or a cylindrical source with slab or concentric shields (Reference 12.3-7).
ANISN Multigroup, multiregion code solving the Boltzmann transport equation for neutrons and gamma rays in one-dimensional slab, cylindrical, or spherical geometry (Reference 12.3-8).
SDC Multigroup, multiregion, gamma ray attenuation code that calculates dose rates for 13 geometry options (Reference 12.3-9).
QAD Multigroup, multiregion, three-dimensional, point kernel code that calculates fast neutron and gamma ray dose and heat generation rates (Reference 12.3-10).
NAP Determines neutron activation and gamma emission source strengths as a function of neutron exposure and decay time (Reference 12.3-11).
MORSE-CG Three-dimensional Monte Carlo neutron and gamma ray general transport code (Reference 12.3 -12).
DOT 3 Two-dimensional neutron, gamma ray, discrete ordinate, transport code (Reference 12.3-13).
ORIGEN Isotopic generation and depletion code that solves equations of radioactive growth and decay for isotopes of arbitrary coupling (Reference 12.3-14).
G-33 A general purpose gamma ray scattering code (Reference 12.3 -15).
These programs are as used in original plant design. Updated versions or alternative programs applying the same basic industry standard analysis methodology may be used in evaluation of plant design changes.
CHAPTER 12 12.3-25 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-3
TURBINE ENCLOSURE VENTILATION SYSTEM DESIGN FEATURES
SUPPLY/EXHAUST RADIOLOGICAL AIR FLOW RATE AREA SAFETY FEATURES (cfm)
General Personnel Three 50% supply 250,000/200,000 Access Areas fans, three 50%
exhaust fans
Equipment Areas Two 100% 0/63,000 exhaust fans, continuously filtered
CHAPTER 12 12.3-26 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-4
CONTROL STRUCTURE VENTILATION SYSTEM DESIGN FEATURES
AIR EXPOSURE TO OPERATION RADIOLOGICAL FLOW RATE AIRBORNE MODE SAFETY FEATURES (cfm) CONCENTRATIONS
Normal Two 100% supply 26,200 Background and return fans
Accident Two 100% supply 3,000 Less than fans. Automatic/ maximum allowable limits manual switch to set in 10CFR20 emergency intake and filtering and recirculation system on high activity signal; recirculation and filtering
CHAPTER 12 12.3-27 REV. 21, SEPTEMBER 2022 LGS UFSAR
Table 12.3-5
REACTOR ENCLOSURE AND REFUELING AREA VENTILATION SYSTEMS DESIGN FEATURES
SUPPLY/EXHAUST RADIOLOGICAL AIR FLOW RATE AREA SAFETY FEATURES (cfm)
Reactor Enclosure Three 50% supply 180,000/140,000 General Personnel fans, three 50%
Access Areas exhaust fans. Emergency:
No or low 60,000 activity exhaust recirculation is not filtered; enclosure isolation 3,000 maximum SGTS on high activity signal. Automatic switch to recirculation and SGTS.
Reactor Enclosure Two 100% exhaust 0/40,000 (1)
Equipment Areas fans, two 100%
filters. Automatic switch to recirculation and SGTS.
Refueling Area Three 50% supply 54,000/54,000 General Personnel fans, three 50%
Access Areas exhaust fans.
No or low activity exhaust is not filtered; area isolation 3,000 maximum SGTS on high activity signal. Automatic switch to SGTS.
(1) Transfer air flow from personnel access areas
CHAPTER 12 12.3-28 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-6
RADWASTE CHEMISTRY LABORATORY AND OFFGAS ENCLOSURES VENTILATION SYSTEM DESIGN FEATURES
SUPPLY/EXHAUST RADIOLOGICAL AIR FLOW RATE AREA SAFETY FEATURES (cfm)
Equipment Areas Two 100% supply fan 0/53,000 cabinets, two 100%
supply fans, and two 100% exhaust fans; all exhaust are passed through HEPA filters.
Service and Control Same as equipment 0/2,850 Areas areas.
Fume Hoods *Two 100% supply 5510/11200 fan cabinets and two 100% exhaust fans; exhausts are passed through HEPA filters.
General Personnel Two 100% supply fan 73,500/16,490 Access Areas cabinets, two 100%
supply fans, and two 100% exhaust fans.
Common Tank Vent One 100% exhaust fan. 0/385
- Located in the Penthouse of the Chemistry Lab Building
CHAPTER 12 12.3-29 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-7
AREA RADIATION MONITORING SYSTEM EQUIPMENT
CHANNEL ELEVATION RANGE NUMBER(4) SENSOR NUMBER DESCRIPTION ENCLOSURE(1) LOCATION(2) (feet) (mRem/hr)
01 RE01-M1-1N001 RCIC pump compartment RE E-14 177.01-104 01 RE01-M1-2N001 RCIC pump compartment RE E-31 177.01-104 02 RE02-M1-1N001 HPCI pump compartment RE F-14 177.01-104 02 RE02-M1-2N001 HPCI pump compartment RE F-31 177.01-104 03 RE03-M1-1N001 Sump compartment RE H-22 177.01-104 03 RE03-M1-2N001 Sump compartment RE H-23 177.01-104 04 RE04-M1-1N001 CRD pumps area TE M-8 200.01-104 04 RE04-M1-2N001 CRD pumps area TE M-38 200.01-104 05 RE05-M1-1N001 Turbine auxiliary bay hallway TE M-12 200.01-104 05 RE05-M1-2N001 Turbine auxiliary bay hallway TE M-34 200.01-104 06 RE06-M1-1N001 Isolation valve compartment RE J-17 201.01-104 06 RE06-M1-2N001 Isolation valve compartment RE J-26 201.01-104 07 RE07-M1-1N001 Condensate pump compartment TE Q-19 189.01-104 07 RE07-M1-2N001 Condensate pump compartment TE Q-27 189.01-104 08 RE08-M1-1N001 RHR division I compartment RE D-17 201.01-104 08 RE08-M1-2N001 RHR division I compartment RE D-25 201.01-104 09 RE09-M1-1N001 RHR division II compartment RE D-20 201.01-104 09 RE09-M1-2N001 RHR division II compartment RE D-30 201.01-104 10 RE10-M1-1N001 Steam vent area stairwell RE E-21 217.01-104 10 RE10-M1-2N001 Steam vent area stairwell RE E-25 217.01-104 11 RE11-M1-1N001 Railroad access airlock RA F-23 217.01-104 11 RE11-M1-2N001 Railroad access airlock RA F-23 217.01-104 12 RE12-M1-1N001 Hallway, condensate filter/demineralizers TE J-13 217.01-104 12 RE12-M1-2N001 Hallway, condensate filter/demineralizers TE J-33 217.01-104 13 RE13-M1-1N002 Condenser area TE N-8 217 1.0-106 13 RE13-M1-2N002 Condenser area TE N-38 217 1.0-106 14 RE14-M1-1N002 Reactor drywell RE F-20 253 1.0-106 14 RE14-M1-2N002 Reactor drywell RE F-29 253 1.0-106 15 RE15-M1-1N001 CRD HCU area east RE G-23 253.01-104 15 RE15-M1-2N001 CRD HCU area east RE G-32 253.01-104 16 RE16-M1-1N001 CRD HCU area, west RE G-14 253.01-104 16 RE16-M1-2N001 CRD HCU area, west RE G-23 253.01-104 17 RE17-M1-1N001 Neutron monitoring system area RE J-21 253.01-104 17 RE17-M1-2N001 Neutron monitoring system area RE J-30 253.01-104 18 RE18-M1-1N001 Neutron monitoring drive mechanism RE H-22 253.01-104 18 RE18-M1-2N001 Neutron monitoring drive mechanism RE J-31 253.01-104 19 RE19-M1-1N002 Turbine auxiliary bay hallway east TE N-17 239 1.0-106 19 RE19-M1-2N002 Turbine auxiliary bay hallway east TE N-36 239 1.0-106
CHAPTER 12 12.3-30 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-7 (Cont'd)
CHANNEL ELEVATION RANGE NUMBER(4) SENSOR NUMBER DESCRIPTION ENCLOSURE(1) LOCATION(2) (feet) (mRem/hr)
20 RE20-M1-1N002 Turbine auxiliary bay hallway west TE N-10 239 1.0-106 20 RE20-M1-2N002 Turbine auxiliary bay hallway west TE N-30 239 1.0-106 21 RE21-M1-1N001 Reactor water cleanup pump area RE F-14 283.01-104 21 RE21-M1-2N001 Reactor water cleanup pump area RE F-31 283.01-104 22 RE22-M1-1N001 Reactor water cleanup heat exchanger area RE E-15 283.01-104 22 RE22-M1-2N001 Reactor water cleanup heat exchanger area RE E-31 283.01-104 23 RE23-M1-1N001 Standby liquid control system area RE F-21 283.01-104 23 RE23-M1-2N001 Standby liquid control system area RE F-24 283.01-104 24 RE24-M1-1N001 RWCU instrument rack area RE H-15 283.01-104 24 RE24-M1-2N001 RWCU instrument rack area RE H-31 283.01-104 25 RE25-M1-1N001 Turbine auxiliary bay TE M-12 269.01-104 25 RE25-M1-2N001 Turbine auxiliary bay TE M-34 269.01-104 26 RE26-M1-1N001 Washdown area TE Q-10 269.01-104 26 RE26-M1-2N001 Washdown area TE Q-37 269.01-104 27 RE27-M1-1N001 Reactor water cleanup filter area RE H-15 313.01-104 27 RE27-M1-2N001 Reactor water cleanup filter area RE H-31 313.01-104 28 RE28-M1-1N001 Equipment compartment exhaust filters area TE K-12 302.01-104 28 RE28-M1-2N001 Equipment compartment exhaust filters area TE K-34 302.01-104 29 RE29-M1-1N001 Drywell head lay-down area RA D-18 352.01-104 29 RE29-M1-2N001 Drywell head lay-down area RA D-28 352.01-104 30 RE30-M1-1N001 Dryer/separator area RA G-14 352.01-104 30 RE30-M1-2N001 Dryer/separator area RA G-32 352.01-104 31 RE31-M1-1N001 Spent fuel pool RA J-21 352.01-104 31 RE31-M1-2N001 Spent fuel pool RA J-25 352.01-104 32 RE32-M1-1N001 Fuel storage vault RA H-22 350.01-104 32 RE32-M1-2N001 Fuel storage vault RA H-24 350.01-104 33 RE33-M1-1N001 Pool plug lay-down area RA J-18 352.01-104 33 RE33-M1-2N001 Pool plug lay-down area RA J-28 352.01-104 34 RE34-M1-1N001 Hydrogen/oxygen analyzers area CS K-23 200.01-104 34 RE34-M1-2N001 Hydrogen/oxygen analyzers area CS K-24 200.01-104 35 RE35-M1-1N001 Gaseous radwaste recombiner hallway CS K-23 180.01-104 35 RE35-M1-2N001 Gaseous radwaste recombiner hallway CS K-23 180.01-104 36 RE36-M1-0N001 Unit 1 deep bed demin area TE N-9 217.01-104 41 RE41-M1-0N001 RWCU Sludge discharge mixing pump-room RWE B-11 162.01-104 42 RE42-M1-0N001 Radwaste enclosure hallway RWE D-12 162.01-104 43 RE43-M1-0N001 Concentrate storage discharge pump-room RWE F-10 191.01-104 44 RE44-M1-0N001 Laundry drain processing room RWE H-10 191.01-104 45 RE45-M1-0N001 Floor drain filter holding pump-room RWE F-12 191.01-104 46 RE46-M1-0N001 Fuel pool holding pump-room RWE G-12 191.01-104 47 RE47-M1-0N001 Precoat tank & pump-room RWE H-14 191.01-104 48 RE48-M1-0N001 Remote shutdown control area CS M-20 289.01-104
CHAPTER 12 12.3-31 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-7 (Cont'd)
CHANNEL ELEVATION RANGE NUMBER(4) SENSOR NUMBER DESCRIPTION ENCLOSURE(1) LOCATION(2) (feet) (mRem/hr)
49 RE49-M1-0N001 Radwaste cask loading area RWE C-11 217.01-104 50 RE50-M1-0N001 Railroad car airlock RA D-23 217 01-104 51 RE51-M1-0N001 Radwaste enclosure hallway RWE G-11 217.01-104 52 RE52-M1-0N001 Hot maintenance shop AB J-34 217.01-104 53 RE53-M1-0N001 Entrance, turbine enclosure railroad TE R-23 217.01-104 54 RE54-M1-0N001 Radwaste building, el 239' RWE F-10 239.01-104 55 RE55-M1-0N001 Radwaste exhaust fan area RWE G-10 257.01-104 56 RE56-M1-0N001 Control room CS J-23 269.01-104 57 RE57-M1-0N001 Turbine area operating floor/operational TE N-23 269.01-104 support center 58 RE58-M1-0N001 Standby gas treatment filter room CS M-25 332.01-104 60 RE60-M1-0N001 North stack instrument room RE H-21 411.01-104
- RIAH-TA-025(3) (Local) Source storage & calibration room RWE D-10 191.1-104
- RIAH-TA-001(3) (Local) A turbine enclosure crane TE variable 310.01-104
- RIAH-TA-002(3) (Local) B turbine enclosure crane TE variable 310.01-104
- RIAH-TA-101(3) (Local) refueling platform RA variable 352.01-104
- RIAH-TA-201(3) (Local) refueling platform RA variable 352.01-104
- RIAH-TA-102(3) (Local) turbine deep bed vsl area TE M-7 217.01-102
- RIAH-TA-103(3) (Local) turbine deep bed vsl area TE M-11 217.01-102
- RIAH-TA-202(3) (Local) turbine deep bed vsl area TE M-38 217.01-102
- RIAH-TA-203(3) (Local) turbine deep bed vsl area TE M-35 217.01-102
(1) RE Reactor enclosure RA Refueling area TE Turbine enclosure RWE Radwaste enclosure CS Control structure AB Administration building (2) Locations indicate east-west and north-south column lines (Drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143).
(3) Local monitor only - requires local power supply.
(4) When the same channel number is given twice, the sensor number ending in 1N001 refers to a Unit 1 sensor, and 2N001 refers to a Unit 2 sensor.
CHAPTER 12 12.3-32 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-8
AREA RADIATION MONITORING ENVIRONMENTAL PARAMETERS
DETECTOR, ENVIRONMENTAL DESIGN MONITOR PREAMPLIFIER PARAMETER CENTER RANGE RANGE
Temperature 77F 41F-122F 32F-140F
Relative Humidity 50% 20%-90% 20%-100%
Pressure Atmospheric Atmospheric Atmospheric
CHAPTER 12 12.3-33 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.3-9
REMOTE EQUIPMENT PANEL LOCATIONS
EQUIPMENT PANEL NOS.
Auxiliary Equipment Room Control Room
Common equipment:
Indicator and trip units 00C643 -
Recorders - 00C624
Annunciators - 00C824
Unitized equipment:
Indicator and trip units 10C605, 20C605
Recorders - 10C600, 20C600
Annunciators - 10C800, 20C800
CHAPTER 12 12.3-34 REV. 13, SEPTEMBER 2006 LGS UFSAR
12.4 DOSE ASSESSMENT
This section discusses the estimated radiation exposures both in-plant and at locations outside the plant structures. Sections 12.4.1 and 12.4.2 discuss direct radiation and airborne radiation exposures within the plant; Section 12.4.3 is concerned with exposures outside the plant structures; and Section 12.4.4 estimates the exposure to Unit 2 construction workers during the operation of Unit 1. The ISFSI radiation exposures are not considered here, as they are addressed in a separate 10CFR72.212 document.
Not all the methods for occupational radiation dose assessment discussed in Regulatory Guide 8.19 are employed for LGS. Alternative methods are described in Section 12.4.1.
12.4.1 Direct Radiation Dose Estimates For Exposures Within The Plant
To estimate the total annual man -rem dose from direct radiation to person nel within the plant, seven broad categories or job functions were defined, and the annual man -rem dose for each category was evaluated. Where the functions and expected radiation levels are predictable or clearly defined, analytical methods were employed for the man-rem estimates. In other cases, the estimate basis is historical exposure data from operating BWR power plants. Section 12.4.1.1 provides the definitions and components of each of the seven broad categories, and Section 12.4.1.2 describes briefly the estimation techniques used.
The resultant dose estimates are contained in Section 12.4.1.3, along with further discussion of the factors involved and the methodology used for each category and its related components.
12.4.1.1 Definition of Categories Used in Exposure Estimates
Seven broad categories were used in estimating the total annual man -rem dose. These categories are:
Routine Operations : This category is composed of the following three components or subcategories:
- a. Routine patrols and surveillances of the reactor enclosure, turbine enclosure, control structure, and radwaste and offgas enclosures.
- b. Periodic tests and checks in the reactor enclosure, turbine enclosure, control structure, and radwaste and offgas enclosures.
- c. Control room operations; specifically, the dose received by operators in the control room and radwaste control room.
Routine Maintenance : This includes all scheduled maintenance. This does not imply that a particular date has been established, but rather t hat the maintenance is planned and occurs at least annually. This category also includes the preventative maintenance performed in the radiation areas of the reactor, turbine, radwaste, and offgas enclosures.
Inservice Inspections: These are inspections normally performed by NDT personnel and outside contractors. Such inspections normally occur during outages on piping and systems that cannot be checked while at power.
CHAPTER 12 12.4-1 REV. 16, SEPTEMBER 2012 LGS UFSAR
Special Maintenance: All maintenance that is not scheduled. This maintenance is not planned in advance and normally cannot be predicted.
Radwaste Processing: This includes any work with solid or liquid radwaste: movement of casks and liners; radwaste, condensate system, or fuel pool filter changes; resin moving; baling of low level trash, etc. Maintenance of radwaste equipment is covered by the maintenance categories and is not included in this job function.
Refueling: This is all work with fuel or reactor components performed in the reactor and pool area.
Health Physics: This covers all health physics activities and includes chemistry operations and sample collection.
12.4.1.2 Exposure Estimate Methodology
Analytical Method: The analytical method used for man-rem estimation is based upon the product of estimated exposure time and estimated ambient dose rate. Estimates of the occupancy time requirements for operations associated with equipment in plant radiation areas (e.g., maintenance time or surveillance time) are first developed. An applicable frequency of occurrence is the n factored in to determine the annual exposure time for each operation. Areas with no significant radiation sources are not included in the exposure estimate. Where radiation sources are present, the design maximum dose rate of 2.5 mRem/hr is assumed for some Zone II areas (radiation zones are defined in Table 12.3 -1). Similarly, 15.0 mRem/hr is assumed for some Zone III areas.
All other estimated dose rates are based on either calculations or actual radiation levels encountered at operating plants. Radiation levels encountered or estimated are adjusted in proportion to power. The analytical method was used in determining the exposure estimates for the routine maintenance and routine operations categories.
Historical Method: In the historical method, the annual man -rem is estimated from the exposures received at operating BWR power plants. This method was used for all the remaining categories (special maintenance, inservice inspection, radwaste processing, refueling, and health physics).
The data sources are the annual and semiannual BWR operating reports and plant correspondence with regulatory agencies. Included are a total of 61 reactor years of operation for 16 nuclear units, which are listed in Table 12.4-1. The average licensed power level of these units is 747 MWe with the smallest rated at 514 Mwe. The data were collected and assembled using the following guidelines:
- a. No data before the first calendar year which contained at least nine months of commercial operation are used.
- b. In multiple unit plants, each unit is assumed to contribute equally to the annual exposures.
- c. If exposure contributions from two or more job functions cannot be separated, a conservative approach is taken by assigning all the exposure to one function and by making no entry in the data base for the other.
Table 12.4-2 contains the results of the historical data compilation and includes both the number of reactor years contributing to and the standard deviations associated with each job function. The large standard deviations, which range from about 60% to 160% of the mean values, are indicative of the wide spread of data that have been reported within each exposure category.
CHAPTER 12 12.4-2 REV. 16, SEPTEMBER 2012 LGS UFSAR
12.4.1.3 Results of Annual Direct Radiation Dose Estimates
The annual man-rem estimates for each category and subcategory are detailed below in Sections 12.4.1.3.1 through 12.4.1.3.7. The methods used in their determination are as described previously, with any additional assumptions or required information included below.
In each of the following sections, the annual exposure estimates are reported for two plant configurations: single-unit operational, and two units operational. In general, the "two-unit dose" is twice the "single-unit dose"; however, the exposures associated with certain job functions are assumed to be independent of the number of units in operation since the functions are performed regardless of whether one or two units are operational. These specific job functions are:
- a. Control room operations
- b. Radwaste control room operations
- c. Radwaste and offgas enclosure routine surveillances
- d. Radwaste and offgas enclosure periodic testing
- e. Radwaste and offgas enclosure routine maintenance
For these estimates, the single-unit dose is assumed to be the same as the two -unit dose.
A summary of the direct radiation dose estimates is given in Section 12.4.1.3.8.
12.4.1.3.1 Routine Operations Dose Estimate
During normal operations, routine patrols and surveillances are performed by plant operators. The majority of items checked are rotating equipment (pumps, fans, etc), and each is viewed to verify the absence of leaks, excessive vibrations, or other abnormal conditions. For the surveillance man-rem exposure estimation, the following assumptions were made:
- a. Dose rates are estimated as outlined in Section 12.4.1.2. Additionally, because of the high potential dose rates associated with certain equipment, routine surveillances of such equipment are performed from a remote location (such as the equipment cell doorway) and credit is taken for the lower ambient radiation level at that point.
- b. Each patrol consists of only one person.
The results of the routine surveillance exposure estimate are contained in Tables 12.4 -3 through 12.4-5.
Similarly, the details and results of the exposure estimate for the periodic testing subcategory are also contained in Tables 12.4-3 through 12.4 -5. Since periodic testing is assumed to occur during equipment shutdown, the estimated shutdown dose rates ar e therefore used for determining the associated exposures.
CHAPTER 12 12.4-3 REV. 16, SEPTEMBER 2012 LGS UFSAR
The remaining subcategory is control room operations exposures. Exposures have been estimated from the anticipated control room radiation levels and the staffing requirements for the main and radwaste control rooms. It is assumed that the staffing levels of both control rooms are identical for either one or two units operational. Table 12.4-6 contains the details of the control room operations exposure estimate.
The total annual exposure estimate for the routine operations category is the sum of the three subcategory annual exposures as shown in Table 12.4-7 and summarized below:
Annual Exposure Estimate: Routine Operations
42.6 man-rem (single-unit operational)
68.9 man-rem (two units operational)
12.4.1.3.2 Routine Maintenance Dose Estimate
A detailed review of plant radiation areas was performed to produce a listing of the types and quantities of selected equipment present in each area. Next, total annual maintenance man-hours were estimated for each equipment type identified based on a combination of operating experience and engineering judgement. Total estimated man-hours are shown in Table 12.4-8 and are intended to include all expected routine activities for each equipment type such as valve repacking, valve relapping, pump seal replacement, fan overhaul, etc.
In any area, the total annual man-hours for routine maintenance is then the summation of the quantity-man-hour products for all equipment types found in the area. Multiplying the area's annual maintenance man-hours by the anticipated area dose rate produces the estimated man-rem by area. As with periodic testing, the estimated shutdown dose rate was used for estimating maintenance exposures.
A "total annual" maintenance approach was used for each component since currently available data generally does not contain sufficient information to provide a basis for man-hour breakdowns by maintenance activity. In addition, the area-by-area methodology employed makes estimate compilations by system unnecessary, since locations where high man-rem expenditures are expected are clearly indicated. Tables 12.4 -3 through 12.4-5 contain the details of the routine maintenance exposure estimate for each enclosure. The sum from the three enclosures is presented below:
Annual Exposure Estimate: Routine Maintenance
232.5 man -rem (single-unit operational)
427.7 man -rem (two units operational)
12.4.1.3.3 Inservice Inspection Dose Estimate
The annual exposure estimate for inservice inspection is based upon the data from operating BWRs given in Table 12.4 -2, and is:
Annual Exposure Estimate: Inservice Inspection
CHAPTER 12 12.4-4 REV. 16, SEPTEMBER 2012 LGS UFSAR
27.5 man-rem (single-unit operational)
55.0 man-rem (two units operational)
12.4.1.3.4 Special Maintenance Dose Estimate
The annual exposure estimate for special maintenance is based upon the data from operating BWRs given in Table 12.4-2, and is:
Annual Exposure Estimate: Special Maintenance
273.1 man-rem (single-unit operational)
546.2 man-rem (two units operational)
12.4.1.3.5 Radwaste Processing Dose Estimate
Most of the operations in the plant associated with the radwaste processing category are performed remotely and are therefore not suitable for evaluation by the analytical estimation technique. Consequently, the annual man-rem estimate for radwaste processing is more properly taken from the historical BWR operating data of Table 12.4-2, since this provides a conservative estimate of the anticipated exposure.
Annual Exposure Estimate: Radwaste Processing
37.0 man-rem (single-unit operational)
74.0 man-rem (two units operational)
12.4.1.3.6 Refueling Dose Estimate
The annual exposure estimate for refueling is based upon the data from operating BWRs given in Table 12.4-2, and is:
Annual Exposure Estimate: Refueling
19.2 man -rem (single-unit operational)
38.4 man -rem (two units operational)
12.4.1.3.7 Health Physics Dose Estimate
The annual exposure estimate for health physics monitoring is based upon the data from operating BWRs given in Table 12.4 -2, and is:
Annual Exposure Estimate: Health Physics
29.3 man -rem (single-unit operational)
58.6 man -rem (two units operational)
CHAPTER 12 12.4-5 REV. 16, SEPTEMBER 2012 LGS UFSAR
12.4.1.3.8 Summary of Direct Radiation Dose Estimates
The annual dose estimates in the preceding seven sections are summarized and totaled in Table 12.4-9. As shown in this table, the estimate of total annual in-plant exposure from direct radiation is:
Annual Exposure Estimate: Total
661.2 man-rem (single-unit operational)
1,268.7 man-rem (two units operational)
Dose estimates for inservice inspection, special maintenance, radwaste processing, refueling, and health physics are based on historical data from operating facilities. Any further breakdown of the dose estimate (such as was made for routine operations and routine maintenance) for these types of activities would still rely primarily on the historical information available. The resultant dose estimate would therefore not be any more precise than an estimate based solely on reported radiation exposures.
In all five areas where historical data is used in the dose estimate, the LGS design includes design features which will reduce actual exposures received by plant personnel. However, due to the lack of sufficiently detailed information to allow the precise quantification of the dose reduction, the calculation of the reduction was not attempted.
As an example of design features which will result in dose reduction, the following design features have been incorporated to facilitate inservice inspection:
- a. Quick removal insulation around the RPV nozzles
12.4.2 AIRBORNE RADIOACTIVITY DOSE ESTIMATES FOR EXPOSURES WITHIN THE PLANT
The estimated exposures to plant personnel from airborne radioactivity are based upon the source distributions and radionuclide concentrations presented in Section 12.2.2 and Tables 12.2 -93 through 12.2-101. Because of the limited geometry afforded by the finite compartment sizes within the plant, personnel exposures due to noble gas immersion are ex pected to be insignificant when compared to inhalation exposures and therefore have not been estimated.
In order to determine whether exposure contributions from airborne radioactive particulates are significant, an evaluation was made in each area of the ratio of total particulate maximum permissible concentration fractions to total radioiodine MPC fractions (which is equivalent to the ratio of particulate MPC-hours to iodine MPC -hours). For the turbine and reactor enclosure areas, the particulate-to-iodine ratios are approximately 0.02 and 0.05, respectively, indicating that the particulate inhalation exposures are not significant in those areas. In the radwaste enclosure areas, however, the particulate-to-iodine ratio is approximately 1.11. Since over 75% of the total particulate MPC fraction is attributable to Co-60, both the thyroid inhalation dose due to
CHAPTER 12 12.4-6 REV. 16, SEPTEMBER 2012 LGS UFSAR
radioiodines and the lung inhalation dose due to Co-60 were estimated for the radwaste enclosure (the thyroid and the lung are the critical organs for iodines and Co-60, respectively).
In addition to the inhalation exposures due to radioiodines (all enclosures) and to Co-60 (radwaste enclosure), whole body exposures due to airborne tritium were also estimated for those areas where tritium is assumed to be present (turbine and reactor enclosures).
Tables 12.4-10 through 12.4-12 contain the compilations of the estimated annual occupancy times and the estimated annual exposures for each of the areas identified in Section 12.2.2 as potential sources of airborne radioactivity. The occupancy times are based upon detailed reviews of each area and the determination of the operations which might occur in those areas. The exposures are based upon the estimated airborne concentrations in Tables 12.2-98 through 12.2-100, dose factors from table C-1 of Regulatory Guide 1.109, and an assumed breathing rate of 3.47x10-4 m3/sec.
12.4.3 Exposures At Locations Outside Plant Structures
The radiation exposures at locations outside the plant structures were estimated for two areas: the site boundary and the visitor's center. Section 12.4.3.1 discusses direct radiation exposure at these locations, and Section 12.4.3.2 deals with airborne exposures.
12.4.3.1 Direct Radiation Dose Estimates Outside Plant Enclosures
At locations outside plant enclosures, the direct radiation exposure has two principal components:
- a. Sources of activity stored outside the enclosures, specifically, the refueling water storage tanks and the CST.
- b. Turbine shine due to the N-16 present in the reactor steam.
Based on the calculated surface dose rates for the refueling water storage tanks and CST given in Section 12.2.1.7, the dose contribution at locations outside the plant enclosures due to these tanks is considered negligible.
The N-16 present in the reactor steam in the primary steam lines, turbines, and moisture separators provides a dose contribution to locations outside the plant enclosure as a result of the high energy gamma rays that it emits as it decays. To reduce the turbi ne shine doses, radiation shielding is provided around each turbine train.
Hydrogen injection by the HWC system causes the reactor water chemistry to become less oxidizing which results in a re-distribution of the N-16 normally produced by radiolysis in the reactor core. Under HWC conditions, more of the N -16 is carried over into the steam and less remains in the reactor water. Thus, areas where steam piping exists will experience a greater level of radiation due to the increased N -16. The increased exp osure is within the estimates provided in Chapter 11 and 12.
Exposure at locations outside plant structures will also increase due to direct radiation and turbine shine from the N -16, which is a high energy gamma emitter. In order to minimize these effec ts, additional radiation shielding has been installed around the high pressure turbine and combined intermediate valves.
CHAPTER 12 12.4-7 REV. 16, SEPTEMBER 2012 LGS UFSAR
The resultant annual exposure due to turbine shine was calculated with the SKYSHINE computer program (Reference 12.4-1). Point sources are used to represent the components on the turbine deck and the source strengths are given in Table 12.4-13.
With an assumed 100% occupancy factor and an 80% capacity factor, the maximum calculated dose rate occurs at the northeast site boundary (Figure 12.4-1 and Table 12.4-14) and is 5.5 mRem/year.
The dose rate in the visitor's center is calculated by the SKYSHINE program to be 1.53x10 -3 mRem/hr with two units operational. Assuming a visitor stays at the visitor's center one day a year for eight hours, the estimated dose for the visitor is 1.2x10-2 mRem/year, as shown in Table 12.4-14.
12.4.3.2 Airborne Radioactivity Dose Estimates Outside Plant Enclosures
Doses at the site boundary due to released radioactivity are given in Section 11.3.
12.4.4 EXPOSURES TO CONSTRUCTION WORKERS
12.4.4.1 Direct Radiation Dose Estimates
The estimated dose rates from direct radiation and turbine shine received by construction workers on Unit 2 due to the operation of Unit 1 are well within the limits of 10CFR20 for exposure to individuals in unrestricted areas. Access to Unit 1 for Unit 2 workers is restricted by permanent concrete walls, security fences or chains, etc.
The estimated dose rates are the sums of the direct radiation from the Unit 1 reactor enclosure, turbine enclosure, radwaste enclosure, and offgas enclosure; and the turbine shine doses resulting from the decay of N-16 in the steam lines and turbine equipment of Unit 1. As discussed in Section 12.4.3.1, dose contributions from outside storage tanks are considered negligible and are not included in the exposure estimate.
The annual dose from Unit 1 operation has been estimated for various points in the Unit 2 construction area. The results of this estimate are listed in Table 12.4-15, and their corresponding points are shown on Figure 12.4-1.
The doses from turbine shine were calculated with the SKYSHINE computer program in the manner described in Section 12.4.3.1. The resultant dose includes the direct as well as air-scattered contribution. Credit is taken for the shielding which is afforded by the Unit 2 walls and floor slabs. The radioactive wastes are processed and stored in the radwaste enclosure, where shielding is provided to ensure that the dose outside the enclosure is less than 0.5 mRem/hr. With an allowance for distance between the radwaste enclosure and the Unit 2 construction area, the estimated direct shine dose is less than 0.01 mrem/hr under normal operating conditions.
The exposure for Unit 2 construction workers has been estimated based on the following assumptions:
- a. The current construction schedule is met.
CHAPTER 12 12.4-8 REV. 16, SEPTEMBER 2012 LGS UFSAR
- b. Personnel assigned to the control structure or Unit 2 turbine enclosure are assumed to work in the turbine enclosure only. Personnel assigned to the Unit 2 reactor enclosure or drywell are assumed to work in the reactor enclosure only.
- c. The average dose rate in the yard areas is the average of the dose rates at points 1 through 5 of Figure 12.4 -1, 0.0335 mRem/hr. The average dose rate in the field office is 0.0215 mRem/hr. Each of these dose rates includes 0.01 mRem/hr for the direct shine contribution and is for 100% plant capacity.
- d. The capacity factor for Unit 1 in determining total estimated exposure is 80%.
- e. 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week per person a t the work site, 50 weeks per year.
Exposure to personnel in various categories and locations is summarized in Table 12.4 -15, which gives the total estimated exposure to Unit 2 construction workers as 41.16 man-rem.
10CFR20.202 specifies that personnel monitoring equipment is required if the maximum expected dose per calendar quarter for workers in an area exceeds one -fourth of the 1250 mRem/quarter limit. It is determined that, even in areas with the highest radiation levels (the turbine deck), no construction worker would receive a dose greater than this, so personnel monitoring equipment is not necessary. However, periodic radiation surveys are to be made by the radiation protection staff.
12.4.4.2 Exposures Due to Airborne Radioactivity
Based on expected annual releases of gaseous effluents (Table 11.3 -1), a worst location annual average atmospheric dispersion factor of 1.48x10 -5 sec/m3, and an annual occupancy of 2000 hr (40 hr/week for 50 weeks/year) the total body gamma, beta skin, an d thyroid gamma dose rates from airborne radionuclides to a construction worker are estimated to be 2.47 mRem/year, 4.77 mRem/year and 0.57 mRem/year, respectively, during the construction stage. Dose rates at specific receptor points are presented in Table 12.4-16. A comparison of these values with those in Table 12.4-14 shows that doses to construction workers resulting from direct shine are dominant.
12.
4.5 REFERENCES
12.4-1 M.G. Wells, D.G. Collins, R.B. Small and J.J. Newell, "SKYSHINE, A Computer Procedure For Evaluation Effect of the Structure Design on N -16 Gamma Ray Dose Rates", RRA-T7209, (November 1, 1972).
CHAPTER 12 12.4-9 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-1
SUMMARY
OF HISTORICAL DATA USED IN COMPILATION OF EXPOSURES RECEIVED AT OPERATING BOILING WATER REACTORS
NUMBER OF OPERATING YEARS ACTUALLY CONTRIBUTING TO DATA BASE DATE OF TOTAL NET COMMERCIAL OPERATING ROUTINE ROUTINE INSERVICE SPECIAL WASTE REFUELING HEALTH ANNUAL PLANT UNIT MWe(1) OPERATION(1) YEARS(2) OPERATIONS MAINT INSPECTION MAINT PROCESSING OPERATIONS PHYSICS TOTAL
Brunswick 2 821 11/75 1 1 1 1 1 1 1 1 1 Cooper - 778 7/74 2 2 2 2 2 2 2 2 2 Dresden 2 809 8/70 6 4 4 0 4 4 4 3 6 Dresden 3 809 10/71 5 4 4 0 4 4 4 3 5 Duane Arnold 1 538 1/75 2 1 1 1 1 1 1 1 2 Edwin I. Hatch 1 786 5/75 1 1 1 1 1 1 1 1 1 Millstone 1 690 3/71 6 3 3 1 3 1 1 2 6 Monticello - 545 6/71 5 4 4 2 4 3 4 2 5 Nine Mile Point 1 610 12/69 7 5 5 1 5 1 5 5 6 Oyster Creek 1 650 12/69 7 7 7 3 4 3 2 6 7 PBAPS 2 1065 7/74 2 2 1 1 1 1 2 2 2 PBAPS 3 1065 12/74 2 2 1 1 1 1 2 2 2 Pilgrim 1 655 12/72 4 3 3 2 3 2 3 2 4 Quad-Cities 1 809 8/72 4 4 4 0 4 4 4 3 4 Quad-Cities 2 809 10/72 4 4 4 0 4 4 4 3 4 Vermont Yankee - 514 11/72 4 2 2 2 2 2 2 2 4
Totals: 62 49 47 18 44 35 42 40 61
(1) Source - "U.S. Central Station Nuclear Electric Generating Units: Significant Milestones", ERDA 77-30/1, (January 1, 1977).
(2) Total number of operating years available to the data base through the end of 1976 beginning with the first calendar year which includes at least nine months of commercial operation
CHAPTER 12 12.4-10 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-2
OCCUPATIONAL EXPOSURES BY JOB FUNCTION FOR OPERATING BOILING WATER REACTORS
STANDARD NUMBER OF AVERAGE DEVIATION REACTOR YEARS MAN-REM PER (MAN-REM PER AVERAGED REACTOR YEAR REACTOR YEAR)
Job Functions:
Routine Operations 49 60.1 43.8 Routine Maintenance 47 110.8 103.4 Inservice Inspection 18 27.5 29.1 Special Maintenance 44 273.1 285.7 Waste Processing 35 37.0 38.0 Refueling 42 19.2 30.1 Health Physics 40 29.3 17.0
Total Reported(1)
Annual Exposure: 61 511.4 454.8
(1) Total exposure by job function differs from the annual reported total exposure due to conservatisms employed in compilation of job function exposures.
CHAPTER 12 12.4-11 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-3
EXPOSURE ESTIMATES FOR THE TURBINE ENCLOSURE AND CONTROL STRUCTURE(1)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(4) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
180'-0" 154 60 - - 24 1.440 203 12.180 161,162 2 - - 0 0.000 282 0.564 161,162 15(3) 30 0.450 - - - -
165 2.5 - - 23 0.058 288 0.720
200'-0" 249 2.3 - - 3 0.007 38 0.087 252A 2.5 - - 6 0.150 12 0.3 252B 3.5 - - 9 0.032 414 1.449 252C 2.5 - - 1.5 0.038 6.15 253 4 - - 15 0.060 745 2.980 254 4 - - 0 0.000 450 1.800 255 4 - - 35 0.140 532 2.128 255 50 30 1.500 - - - -
256 2 - - 92 0.184 1,481 2.962 256 60 91 5.460 - - - -
264 2.3 - - 96 0.221 1,909 4.391 264 15(3) 30 0.450 - - - -
270A 25.0 - - 6 0.150 12 0.300 270B 3.5 - - 9 0.032 414 1.449 270C 25.0 - - 1.5 0.038 6 0.150 299 2.5 - - 11 0.028 0 0.000
217'-0" 332 4 - - 158 0.632 804 3.216 332 50 20 1.000 - - - -
333,334 6 - - 0 0.000 1,286 7.716 333,334 5(3) 46 0.230 - - - -
337 2.5 13 0.033 15 0.038 422 1.055 340 1.6 - - 21 0.034 927 1.483 340 2.5(3) 20 0.050 - - - -
341-H 25 - - - - 216 5.4 342 2 - - 3 0.006 47 0.094 354 A-H 25 - - - - 216 5.4
239'-0" 407 29 - - 0 0.000 30 0.870 438 4 - - 78 0.312 67 0.268 438 50 4 0.200 - - - -
439-441 3 - - 36 0.108 1,041 3.123 439-441 10(3) 13 0.130 - - - -
499 4 - - 56 0.224 1,696 6.784 499 50 7 0.350 - - - -
CHAPTER 12 12.4-12 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-3 (Cont'd)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(4) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
269'-0" 518 29 - - 0 0.000 88 2.552 544 1.6 - - 0 0.000 3,568 5.709 544 15(3) 9 0.135 - - - -
545-547 3 - - 9 0.027 426 1.278 545-547 10(3) 13 0.130 - - - -
551 3 - - 8 0.024 376 1.128 551 10(3) 4 0.040 - - - -
302'-0" 621 15 13 0.195 0 0.000 76 1.140
332'-0" 624 2.5 - - 46 0.115 106 0.265 TOTALS: 343 10.353 746 3.788 17,536 71.792
(1) All values are on a per unit basis.
(2) The estimated exposures assume all man-hours are expended in the area in which they appear. Portions of these man -hours may actually be spent at lower radiation levels within the area. Components may also be removed to a lower background area for maintenance.
(3) Estimated dose rate is based upon surveillance being performed from a remote location such as the equipment cell doorway. Estimated and measured dose rates will increase approximately in proportion to power.
(4) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-13 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-4
EXPOSURE ESTIMATES FOR THE REACTOR ENCLOSURE(1)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE
ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(5) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
177'-0" 102 15 - - 66 0.990 752 11.280 102 2.5(3) 46 0.115 - - - -
103 15 - - 66 0.990 752 11.280 103 2.5(3) 46 0.115 - - - -
108 2.5 46 0.115 72 0.180 650 1.625 109 2.5 46 0.115 83 0.208 917 2.293 115 2.5 - - 23 0.058 288 0.720 115 15 46 0.690 - - - -
201'-0" 203 15 - - 11 0.165 422 6.330 204 15 - - 11 0.165 238 3.570 209 20 - - 0 0.000 98 1.960 297 2.5 - - 0 0.000 84 0.210 298 2.5 - - 0 0.000 26 0.065
217'-0" 309 15 - - 5 0.075 683 10.245
253'-0" 402A 2.5 137 0.343 0 0.000 600 1.500 403 13 - - 222(4) 2.886 0 0.000 406 15 - - 0 0.000 126 1.890 407A 29 - - 0 0.000 114 3.306
283'-0" 501 2.5 - - 39 0.098 44 0.110 503-505 28 - - 29 0.812 537 15.036 503-505 5(3) 13 0.065 - - - -
507-509 28 - - 33 0.924 360 10.080 507-509 5(3) 20 0.100 - - - -
510 28 - - 0 0.000 135 3.780 511 16 - - 66 1.056 570 9.120 511 5(3) 30 0.150 - - - -
518A 29 - - 0 0.000 74 2.146 522 28 - - 24 0.672 125 3.500 523 15 - - 0 0.000 95 1.425 599 2.5 - - 0 0.000 156 0.390
CHAPTER 12 12.4-14 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-4 (Cont'd)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE
ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(5) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
313'-0" 600 2.3 - - 6 0.014 519 1.194 600 15(3) 13 0.195 - - - -
616,617 15 13 0.195 0 0.000 40 0.600 618 15 13 0.195 3 0.045 40 0.600 352'-0" 700 1 - - 119 0.119 2,068 2.068 700 2.5 46 0.115 - - - -
Drywell 400 10 - - 0 0.000 191 1.910 400A 11 - - 0 0.000 442 4.862 400B 10 - - 0 0.000 532 5.320 400C 10 - - 15 0.150 184 1.840 400D 10 - - 0 0.000 303 3.030 400E 10 - - 0 0.000 11 0.110 TOTALS: 515 2.508 893 9.607 12,176 123.395
(1) All values are on a per unit basis.
(2) The estimated exposures assume all man-hours are expended in the area in which they appear. Portions of these man-hours may actually be spent at lower radiation levels within the area. Components may also be removed to a lower background area for maintenance.
(3) Estimated dose rate is based upon surveillance being performed from a remote location such as the equipment cell doorway. Estimated and measured dose rates will increase approximately in proportion to power.
(4) Control rod drive rebuild, repair, and testing.
(5) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-15 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-5
EXPOSURE ESTIMATES FOR THE RADWASTE AND OFFGAS ENCLOSURES(1)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE
ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(5) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
162'-0" 124 10 - - 6 0.060 212 2.120 124 5(3) 6 0.030 - - - -
126,130 3 - - 10 0.030 449 1.347 126,130 15 12 0.180 - - - -
127,129 10 - - 12 0.120 449 4.490 127,129 5(3) 12 0.060 - - - -
132 3 - - 5 0.015 165 0.495 132 5(3) 6 0.030 - - - -
133 3 - - 5 0.015 571 1.713 133 5(3) 6 0.030 - - - -
137 2.5 - - 2 0.005 168 0.420 137 15 6 0.090 - - - -
138 10 - - 3 0.030 146 1.460 138 5(3) 6 0.030 - - - -
139 10 - - 3 0.030 146 1.460 139 5(3) 6 0.030 - - - -
142 3 - - 17 0.051 109 0.327 142 5(3) 6 0.030 - - - -
144 10 - - 11 0.110 233 2.333 144 5(3) 6 0.030 - - - -
146 3 - - 21 0.063 288 0.864 146 15 6 0.090 - - - -
147 2.5 - - 3 0.008 287 0.718 147 2.5 6 0.015 - - - -
148 2.5 - - 6 0.015 0 0.000 148 2.5 6 0.015 - - - -
Offgas Areas 5 - - 0 0.000 780 3.900
191'-0" 228 2.5 6 0.015 59 0.148 404 1.010 231 2.5 - - 6 0.015 203 0.508 231 15 6 0.090 - - - -
232 2.5 - - 5 0.013 117 0.293 232 15 6 0.090 - - - -
233 2.6 - - 3 0.008 3 0.008
CHAPTER 12 12.4-16 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-5 (Cont'd)
ROUTINE SURVEILLANCE PERIODIC TESTING ROUTINE MAINTENANCE
ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ESTIMATED ROOM DOSE RATE ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ANNUAL ELEVATION NO.(5) (mRem/hr) MAN-HOURS MAN-REM MAN-HOURS MAN-REM MAN-HOURS MAN-REM
234 2.6 - - 2 0.005 122 0.317 234 15 6 0.090 - - - -
236 2.3 - - 72 0.166 549 1.263 236 5(3) 6 0.030 - - - -
237 2.3 - - 26 0.060 253 0.582 237 5(3) 6 0.030 - - - -
238 2.3 - - 17 0.039 300 0.690 238 5(3) 6 0.030 - - - -
242,243 20 - - 16 0.320 100 2.000 242,243 5(3) 12 0.060 - - - -
245 42 - - 0 0.000 78 3.276 247 3 - - 5 0.015 142 0.426 247 5(3) 6 0.030 - - - -
248 3 - - 5 0.015 142 0.426 248 5(3) 6 0.030 - - - -
217'-0" 422 2.5 - - 0 0.000 60 0.150 423 2.5 - - 0 0.000 40 0.100
237'-0" 470,471 30 - - 0 0.000 34 1.020 470,471 5(3) 61 0.305 - - - -
485,486 15 - - 0 0.000 40 0.600
257'-0" 515,516 30 - - 0 0.000 100 3.000 515,516 5(3) 30 0.150 - - - -
TOTAL: 247 1.610 320 1.356 6,690 37.316
(1) All values are on a per unit basis.
(2) The estimated exposures assume all man-hours are expended in the area in which they appear. Portions of these man-hours may actually be spent at lower radiation levels within the area. Components may also be removed to a lower background area for maintenance.
(3) Estimated dose rate is based upon surveillance being performed from a remote location such as the equipment cell doorway. Estimated and measured dose rates will increase approximately in proportion to power.
(4) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-17 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-6
ESTIMATED EXPOSURE FOR OPERATORS IN CONTROL ROOMS
ANNUAL ESTIMATED ANNUAL OPERATORS(1) SHIFTS MANNED(1) OPERATOR DOSE RATE OPERATOR PER SHIFT PER DAY MAN-HOURS(2) (mRem/hr) MAN-REM(1)
Control Room 5 3 43,800 0.25 11.0
Radwaste Control 1 3 8,760 0.25 2.2 Room ___ ______ ___
TOTAL 6 52,560 13.2
(1) Assumed to be independent of the number of units in operation (2) Based on each operator spending 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the control room and the control rooms being manned 365 days per year
CHAPTER 12 12.4-18 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-7
SUMMARY
OF ROUTINE OPERATIONS EXPOSURE ESTIMATE (3)
ANNUAL ESTIMATED MAN -REM SINGLE-UNIT TWO UNITS
Routine Surveillances:
Turbine enclosure(1) 10.4 20.7 Reactor enclosure 2.5 5.0 Radwaste enclosure(2) 1.6 1.6 14.5 27.3
Periodic Tests:
Turbine enclosure(1) 3.9 7.8 Reactor enclosure 9.6 19.2 Radwaste enclosure(2) 1.4 1.4 14.9 28.4
Control Room Operations:
Control room 11.0 11.0 Radwaste control room 2.2 2.2 13.2 13.2
TOTAL 42.6 68.9
(1) Includes control structure exposures (2) Includes offgas enclosure exposures (3) Estimated and measured dose rates will increase approximately in proportion to power.
CHAPTER 12 12.4-19 REV. 16, SEPTEMBER 2012 LGS UFSAR
Table 12.4-8
ESTIMATE OF EXPECTED ROUTINE MAINTENANCE REQUIREMENTS
ESTIMATED MAN-HOURS EQUIPMENT TYPE OR ACTIVITY PER YEAR
- 1. Valves
- a. Under 3 inches 3
- b. 3 to 6 inches 9
- c. 8 to 10 inches 12
- d. 12 to 16 inches 17
- e. Over 16 inches 23
- 2. Valve Operators
- a. Reach rod 1
- b. Air operator 2
- c. Motor operator 2
- 3. Pumps
- a. Reactor recirculation 150
- b. RHR or condensate 100
- c. Any other pump 50
- 4. Motors
- a. 200 hp or greater 20
- b. Less than 200 hp 13
- 5. Motor-generator 35
- 6. Main generator 1000
- 7. Turbines
- a. Main turbine 2400
- b. Any other turbine 240
- 8. Heat exchanger 30
- 9. Deleted 50
- 10. Chiller 5
- 11. Unit heater 5
- 12. Unit cooler 5
- 13. Compressor 24
CHAPTER 12 12.4-20 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-8 (Cont'd)
ESTIMATED MAN -HOURS EQUIPMENT TYPE OR ACTIVITY PER YEAR
- 14. Fan or blower25
- 15. Main condenser (per shell) 150
- 16. Hoist or crane 60
- 17. Instrument panel 150
- 18. Instrument rack 150
- 19. Motor control center 5
- 20. Switchgear 120
- 21. Centrifuge 37
- 22. Agitator 5
- 23. Transfer cart or conveyor 15
- 24. TIP drive 20
- 25. CRD hydraulic units (total per reactor) 600
- 27. Offgas system (holdup pipe 780 to stack - total for station)
CHAPTER 12 12.4-21 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-9
SUMMARY
OF IN-PLANT DIRECT RADIATION EXPOSURE ESTIMATES (1)
ANNUAL ESTIMATED MAN-REM CATEGORY SINGLE-UNIT TWO UNITS
Routine Operations 42.6 68.9
Routine Maintenance 232.5 427.7
Inservice Inspection 27.5 55.0
Special Maintenance 273.1 546.2
Radwaste Processing 37.0 74.0
Refueling 19.2 38.4
Health Physics 29.3 58.6
Total 661.2 1,268.8
(1) Estimated and measured do se rates will increase approximately in proportion to power.
CHAPTER 12 12.4-22 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-10
ESTIMATED TURBINE ENCLOSURE AND CONTROL STRUCTURE INHALATION EXPOSURES DUE TO AIRBORNE RADIOACTIVITY(1)
ESTIMATED ESTIMATED ROOM(4) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE(2) TRITIUM DOSE(3)
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
Condenser Areas
253 0 15 745 760 2.3 2.1x10-2 254 0 0 450 450 1.4 1.3x10-2 255 30 35 532 597 1.8 1.5x10-2 256 91 92 1481 1664 5.0 4.7x10-2 332 20 158 804 982 2.9 2.7x10-2 342 0 3 47 50 1.5x10-1 1.4x10-3 499 7 56 1696 1759 5.3 4.9x10-2
SJAE Areas
154 0 24 203 227 5.7x10-2 5.2x10-4 333,334 46 0 1286 133 3.3x10-1 3.1x10-3
Mechanical Vacuum Pump Areas
337 13 15 422 450 3.0 2.9x10-2
Turbine Hall Areas
544 9 0 3568 3577 6.4x10-1 6.1x10-3
CHAPTER 12 12.4-23 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-10 (Cont'd)
ESTIMATED ESTIMATED ROOM(4) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE(2 TRITIUM DOSE(3)
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
Other Equipment Areas
161,162 30 0 282 312 5.9x10-2 5.6x10-4 165 0 23 288 311 5.9x10-2 5.6x10-4 249 0 3 38 41 7.8x10-3 7.4x10-5 264 30 96 1909 2035 3.9x10-1 3.7x10-3 299 0 11 0 11 2.1x10-3 2.0x10-5 340 20 21 927 968 1.8x10-1 1.7x10-3 407 0 0 30 30 5.7x10-3 5.4x10-5 438 4 78 67 149 2.8x10-2 2.7x10-4 439,440,441 13 36 1041 1090 2.1x10-1 2.0x10-3 518 0 0 88 88 1.7x10-2 1.6x10-4 545,546,547 13 9 426 448 8.5x10-2 8.1x10-4 551 4 8 376 388 7.4x10-2 7.0x10-4 621 13 0 76 89 1.7x10-2 1.6x10-4 624 0 46 106 152 2.9x10-2 2.7x10-4
TOTALS 343 729 16,888 17,960 2.4x10+1 2.2x10-1
(1) All values in the table are on a per unit basis.
(2) Thyroid dose attributable only to inhalation of radioiodines.
(3) Uniform dose to the total body from uptake of tritium.
(4) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-24 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-11
ESTIMATED REACTOR ENCLOSURE INHALATION EXPOSURES DUE TO AIRBORNE RADIOACTIVITY(1)
ESTIMATED ESTIMATED ROOM(6) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE(4) TRITIUM DOSE(5)
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
RWCU Pump Areas 503,504,505 13 29 537 579 9.8x10-1 -
507,508,509 20 33 360 413 7.0x10-1 -
510 0 0 135 135 2.3x10-1 -
522 0 24 125 149 2.5x10-1 -
RWCU Filter/
Demineralizer Areas 600 13 6 519 538 1.7 -
Refueling Areas 700 46 119 4,233(2) 4,398 6.6x10-1 3.8x10-2
ECCS Areas 102 46 66 752 864 5.8x10-1 -
103 46 66 752 864 5.8x10-1 -
108 46 72 650 768 5.1x10-1 -
109 46 83 917 1,046 7.0x10-1 -
203 0 11 422 433 2.9x10-1 -
204 0 11 238 249 1.7x10-1 -
297 0 0 84 84 5.6x10-2 -
298 0 0 26 26 1.7x10-2 -
309 0 5 683 688 4.6x10-1 -
400 0 0 191 191 5.5 -
400A 0 0 442 442 1.3x10+ -
400B 0 0 532 532 1.5x10+1 -
400C 0 15 184 199 5.8 -
400D 0 0 303 303 8.8 -
400E 0 0 11 11 3.2x10-1 -
CHAPTER 12 12.4-25 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-11 (Cont'd)
ESTIMATED ESTIMATED ROOM(6) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE(4) TRITIUM DOSE(5)
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
Other Equipment Areas
115 46 23 288 357 1.9x10-1 -
209 0 0 98 98 5.1x10-2 -
402A 137 0 600 737 3.8x10-1 -
403 0 222 0 222 1.2x10-1 -
406 0 0 126 126 6.6x10-2 -
407A 0 0 114 114 5.9x10-2 -
501 0 39 44 83 4.3x10-2 -
511 30 66 570 666 3.5x10-1 -
518A 0 0 74 74 3.8x10-2 -
523 0 0 95 95 4.9x10-2 -
599 0 0 156 156 8.1x10-2 -
616,617 13 0 40 53 2.8x10-2 -
618 13 3 40 56 2.9x10-2
TOTALS 515 893 14,341 15,749 5.8x10+1(3) 3.8x10-2
(1) All values in the table are on a per unit basis.
(2) Maintenance man-hours include 2,165 man-hours for refueling operations.
(3) The total thyroid dose of 58 man-rem takes no credit for respiratory protection. If a protection factor as low as 5 is assumed in the drywell (which is estimated to produce over 80% of the reactor enclosure thyroid exposures), the total thyroid dose is reduced to below 20 man-rem per year.
(4) Thyroid dose attributable only to inhalation of radioiodines.
(5) Uniform dose to the total body from uptake of tritium.
(6) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-26 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-12
ESTIMATED RADWASTE AND OFFGAS ENCLOSURES INHALATION EXPOSURES DUE TO AIRBORNE RADIOACTIVITY(1)
ESTIMATED ESTIMATED ROOM(3) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE LUNG DOSE
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
Solid Radwaste Handling Areas
124 6 6 212 224 3.6x10-2 1.7x10-2 126,130 12 10 449 471 7.5x10-2 3.5x10-2 127,129 12 12 449 473 7.6x10-2 3.5x10-2 138 6 3 146 155 2.5x10-2 1.2x10-2 139 6 3 146 155 2.5x10-2 1.2x10-2 142 6 17 109 132 2.1x10-2 9.9x10-3 144 6 11 233 250 4.0x10-2 1.9x10-2 233 0 3 3 6 9.6x10-4 4.5x10-4 234 6 2 122 130 2.1x10-2 9.8x10-3 422 0 0 60 60 9.6x10-3 4.5x10-3 423 0 0 40 40 6.4x10-3 3.0x10-3 470,471 61 0 34 95 1.5x10-2 7.1x10-3 515,516 30 0 100 130 2.1x10-2 9.8x10-3 (2) 0 0 4000 4000 6.4x10-1 3.0x10-1
Liquid Radwaste Handling Areas
132 6 5 165 176 1.6x10-1 7.6x10-2 133 6 5 571 582 5.3x10-1 2.5x10-1 137 6 2 168 176 1.6x10-1 7.6x10-2 147 6 3 287 296 2.7x10-1 1.3x10-1 148 6 6 0 12 1.1x10-2 5.2x10-3 231 6 6 203 215 2.0x10-1 9.2x10-2 232 6 5 117 128 1.2x10-1 5.5x10-2
CHAPTER 12 12.4-27 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-12 (Cont'd)
ESTIMATED ESTIMATED ROOM(3) ESTIMATED ANNUAL MAN-HOURS THYROID DOSE LUNG DOSE
SURVEILLANCE TESTING MAINTENANCE TOTAL (MAN-REM/YR) (MAN-REM/YR)
Liquid Radwaste Handling Areas (continued)
237 6 26 253 285 2.6x10-1 1.2x10-1 238 6 17 300 323 2.9x10-1 1.4x10-1 242,243 12 16 100 128 1.2x10-1 5.5x10-2 245 0 0 78 78 7.1x10-2 3.4x10-2 247 6 5 142 153 1.4x10-1 6.6x10-2 248 6 5 142 153 1.4x10-1 6.6x10-2
Other Equipment Areas
146 6 21 288 315 1.2x10-1 5.7x10-2 228 6 59 404 469 1.8x10-1 8.4x10-2 236 6 72 549 627 2.4x10-1 1.1x10-1 485,486 0 0 40 40 1.5x10-2 7.2x10-3 OFFGAS 0 0 780 780 3.0x10-1 1.4x10-1
TOTALS 247 320 10,690 11,257 4.3 2.0
(1) All values are on a per unit basis.
(2) Solid radwaste processing, container handling, and radwaste shipping.
(3) Room numbers are shown on drawings N-110, N-111, N-112, N-113, N-115, N-116, N-117, N-118, N-119, N-120, N-121, N-122, N-124, N-125, N-126, N-127, N-128, N-130, N-131, N-132, N-133, N-134, N-135, N-136, N-137, N-140, N-141, N-142, and N-143.
CHAPTER 12 12.4-28 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-13
SKYSHINE SOURCE TERMS FOR ONE -UNIT OPERATING (1)
SOURCE DESCRIPTION (6.2 MeV)/sec
HP Turbine 4.209x109
LP Turbine A 3.525x109
LP Turbine B 3.405x109
LP Turbine C 4.038x109
28" HP Turbine Inlet Pipe (Two) 3.010x1010 (Each)
42" HP Turbine Exhaust Pipe (Two) 2.836x1010 (Each)
42" Cross-Around Pipe (Six) 1.866x1010 (Each)
Steam Seal Evaporator 1.239x1010
(1) These are all the exposed sources above the turbine operating deck, el 269'-0".
CHAPTER 12 12.4-29 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-14
DIRECT RADIATION DOSE RATES (1)
YARD AREA(2)
DOSE RATE(3)
RECEPTOR POINT (mRem/hr)
- 1. Field Office 2.30x10-2
- 2. Lay-down Area 1.47x10-2
- 3. Circulating Water Pump Structure 5.67x10-2
- 4. Sewage Treatment Plant 2.52x10-2
- 5. Access Road 5.97x10-2
SITE BOUNDARY AND VISITORS' CENTER (4)
DISTANCE FROM DOSE RATE RECEPTOR POINT ORIGIN (ft) (mRem/yr)
- 6. Visitor's Center 1724 1.18x10-2(5)
- 7. N 2835 4.3
- 8. NNE 2550 5.4
- 9. NE 2560 5.5
- 10. ENE 2579 3.9
- 11. E 2581 3.6
- 12. ESE 2579 1.9
- 13. SE 2570 1.5
- 14. SSE 3247 0.6
- 15. S 2504 3.7
- 16. SSW 2549 2.9
- 17. SW 2942 0.7
- 18. WSW 2787 1.3
- 19. W 2825 1.8
- 20. WNW 2636 2.8
- 21. NW 2571 3.4
- 22. NNW 2818 4.3
(1) Locations of receptor points and origin are shown on Figure 12.4 -1. For receptor points 7 through 22, the distances given are the closest location to the origin within each of the 16 sectors.
(2) For Unit 1 only, operation at a capacity of 100%
(3) Dose rates include 0.01 mRem/hr for direct radiation from the radwaste enclosure.
(4) For both units operating at full power at a capacity of 80%
(5) Based on continuous occupancy of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per year at 100% plant capacity.
CHAPTER 12 12.4-30 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-15
ESTIMATED EXPOSURE TO UNIT 2 CONSTRUCTION WORKERS ESTIMATED AVERAGE ESTIMATED ASSUMED OCCUPANCY (1) DOSE RATE EXPOSURE(2)
PERSONNEL LOCATION (MAN-HOURS) (mRem/hr)(4) (MAN-REM)
Manual Yard Area (3) 89,000 0.0335(5) 2.39
Manual Unit 2 Reactor 91,000 0.01 0.73 Enclosure Below Refueling Floor and Containment
Manual On and Above 25,000 0.311 6.22 Refueling Floor
Manual Unit 2 Turbine 150,000 0.01 1.20 Enclosure Below Turbine Deck
Manual On and Above 62,000 0.546 27.08 Turbine Deck
Nonmanual Field Office 62,000 0.0215 1.07
Nonmanual Unit 2 Reactor 19,000 0.01 0.15 Enclosure Below Refueling Floor and Containment
Nonmanual On and Above 5,300 0.311 1.32 Refueling Floor
Nonmanual Unit 2 Turbine 32,000 0.01 0.26 Enclosure Below Turbine Deck
Nonmanual On and Above 13,000 0.0711 0.74 Turbine Deck
TOTAL 548,300 41.16 (1) For remainder of Unit 2 construction, after Unit 1 operation commences.
(2) Based on an assumed availability of 80% for Unit 1.
(3) Average of points 1 through 5 shown on Figure 12.4-1.
(4) Based on operation at 100% power.
(5) From Table 12.4 -14.
CHAPTER 12 12.4-31 REV. 13, SEPTEMBER 2006 LGS UFSAR
Table 12.4-16
AIRBORNE RADIOACTIVITY DOSES TO CONSTRUCTION WORKERS IN YARD AREA
DOSE RATES (mRem/yr)
TOTAL BODY SKIN THYROID RECEPTOR POINT (1) (GAMMA) (BETA) (GAMMA)
- 1. Field Office 2.47 4.77 0.57
- 2. Lay-down Area 1.14 2.21 0.27
- 3. Circulating Water 1.14 2.21 0.27 Pump Structure
- 4. Sewage Treatment 1.14 2.21 0.27 Plant
- 5. Access Road 2.47 4.77 0.57
(1) Locations of receptor points are shown on Figure 12.4-1.
CHAPTER 12 12.4-32 REV. 13, SEPTEMBER 2006 LGS UFSAR
12.5 HEALTH PHYSICS PROGRAM
12.5.1 ORGANIZATION
The organizational structure, including the responsibilities of those charged with performance of health physics duties, is described in Section 13.1. Health physics operations are made effective and ALARA goals are achieved through the cooperation and active participation of the various group represented in this structure.
The objectives of health physics operations are to:
- a. Detect, identify and define radiation hazards.
- b. Provide protection for personnel against radiation hazards.
- c. Monitor and measure radioactive effluents from the plant.
- d. Control plant-related radiation exposures (occupational and general public) to levels ALARA.
- e. Control plant radioactive effluents to levels ALARA.
- f. Conduct plant activities in conformance with authorized procedures and applicable regulations.
These objectives are accomplished or enhanced by the various training programs, by the use of prepared procedures, by performance of area surveys and monitoring of personnel, by record keeping, by applying good health physics practices, by periodic review and revision of procedures, by evaluation of activities for ALARA purposes, and by the control of plant operations to minimize occupational exposures and releases to the environment. Health physics operations include the posting, notification, and reporting provisions of 10CFR19 and General Employee Trainin g (Section 13.2) provides the requisite instruction to workers.
Health physics operations conform with the guidelines of Regulatory Guide 8.2 (Rev 0) and of ANSI N13.2 (1969), Guide for Administrative Practices in Radiation Monitoring, with the clarification that controlled areas which are locked or otherwise prevent personnel access are not required to be surveyed at a specific periodicity. In addition to this section, Sections 13.1, 13.2, 13.5, and 12.1 describe the management commitment, organization, r esponsibilities, authority, training, procedures, and review techniques which implement Regulatory Guide 8.8 (Rev 3), as it applies to the operating phase and Regulatory Guide 1.8 (Rev 1 -R). As described in Section 12.1, a formal ALARA review program, whi ch is consistent with the guidelines given in Regulatory Guide 8.8, was implemented during the design and construction phase. The qualification requirements for the Radiation Protection Manager are described in Section 13.1.3. Chapter 17 addresses implementation of other applicable regulatory guides. Implementation of Regulatory Guide 1.16 is addressed in the Technical Specifications.
12.5.2 FACILITIES, EQUIPMENT, AND INSTRUMENTS
This section describes the basic facilities, equipment, and instruments directly related to the health physics operations.
CHAPTER 12 12.5-1 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.2.1 Health Physics and Chemistry Facilities
12.5.2.1.1 Supervisory Offices and Field Offices
Health Physics and Chemistry supervisory offices are located outside the RCA. Field offices and work areas for technicians are located adjacent to the RCA.
12.5.2.1.2 Chemical Laboratories
Chemical laboratories are located in the Chemistry Laboratory Expansion including the Hot Lab, the Instrumentation Lab (within the bounds of the RCA) and the Cold Lab (beyond the bounds of the RCA). A conventional chemistry section and radiochemistry section are provided. These sections are segregated so that noncontaminated samples (such as river water) can be prepared and analyzed without having to work under radiologically controlled conditions. The laboratories are accessible to the counting room for convenience in transporting prepared samples for counting.
Sample stations are located near their respective systems for sampling condensate, feedwater, reactor water, and radwaste. The laboratories are designed to support routine chemical analysis, as well as, developmental work for emergent issues and appropriate emergency response functions as described in the Emergency Plan.
12.5.2.1.3 Counting Room
The counting room is located in the Chemistry Laboratory Expansion in a low background area.
The room is designed with thick shield walls to further reduce background radiation, thus improving instrumentation sensitivity.
12.5.2.1.4 Control Points
Control points and change areas are located throughout the plant near the various job sites. These facilities are for the most part temporary. They are set up during outages and special jobs. They are staffed and stocked to provide health physics monitoring and protective equipment on an as-needed basis.
12.5.2.1.5 Testing and Laboratory Facilities
Calibration and instrument repair facilities are provided. Licensed sources are stored in this facility under approval of the Manager Radiation Protection. Sources are used to calibrate portable radiation survey meters, direct reading dosimeters, certain process and effluent radiation instruments, and area monitors.
12.5.2.1.6 Medical Support
Arrangements with local physicians and hospitals pr ovide offsite support for continued medical treatment of traumas involving radioactive contamination.
A contract is maintained with a primary health physics-medical consultant. The consultant retains a qualified staff that provides expertise in the long-term treatment of patients exposed to excessive external radiation or internal contamination.
CHAPTER 12 12.5-2 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.2.1.7 Personnel Decontamination Facility
A personnel decontamination facility located in the radwaste enclosure at el 217' is provided with a sink, shower, change area, and contamination monitoring devices. Personnel decontamination is performed in accordance with Health Physics procedures. Control of personnel decontamination is provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program.
No male/female discrimination has been made in the design for the decontamination facilities.
Adequate showers and sinks are available for personnel decontamination. Procedures will be implemented to provide the appropriate privacy. Additional male and female facilities are provided in the administration building but are not intended for contamination removal.
12.5.2.1.8 Laundry and Respirator Facility
A laundry and respirator facility is located adjacent to the Unit 1 turbine enclosure at el 217'. The facility includes space for sorting anticontamination clothing, and for receiving clothing from an offsite laundry. The facility includes space for inspecting, surveying, bagging, and repairing respirators. Procedures are established for inspection, cleaning and maintenance of such protective equipment.
12.5.2.2 Instruments and Equipment
Instrumentation design and capabilities improve as the state of the art changes. Normally, the instrumentation listed in Table 12.5-1 or instruments of similar capabilities will be provided. Any instrumentation or equipment described in these sections may be replaced by items providing similar or improved capabilities.
Instrumentation for detecting and measuring radiation consists of counting room equipment, portable instrumentation, and air samplers. Capabilities for detecting alpha, beta, gamma, and neutron radiation are provided. Sufficient inventory is provided to accommodate use, repairs and calibration. The equipment is described later in this section and is listed in Table 12.5-1. Sufficient chemical equipment and analytical instruments are provided to perform the required sample preparations and analyses.
LGS will equip the counting room and provide the portable instruments, personnel monitoring instruments, and protective equipment with the capabilities as described in Regulatory Guide 8.8, section 4. Support facilities at LGS will also follow the guidance of Regulatory Guide 8.8.
LGS is in compliance with Regulatory Guide 1.97 (Rev 2) in regard to Health Physics and Chemistry Laboratory and survey equipment.
Control of instrument storage, calibration and maintenance is provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program.
Operability of personnel monitors, radiation survey instruments, and laboratory equipment is verified periodically in accordance with plant procedures and work processes.
CHAPTER 12 12.5-3 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.2.2.1 Chemical Laboratories
The chemical laboratories are segregated so that one section handles only low level or background samples and the remaining section handles contaminated samples.
Grab samples are transported to the appropriate section of the laboratories. The laboratories, are equipped with constant air flow fume hoods. The fume hoods permit safe preparation and processing of samples under controlled conditions. Effluent of fume hoods within the RCA are filtered through HEPA filters. The laboratories are stocked with chemical reagents and equipment to provide for sample preparation and analysis.
The laboratories contain conventional analytical instruments, equipment and chemicals. Samples intended for activity analysis and isotopic identification are prepared for transport to the counting room. A frisking station is located in the Chemistry Laboratory Expansion.
12.5.2.2.2 Counting Room
Samples processed in the chemical laboratories for activity analysis and isotopic identification, and samples direct from the plant (such as air samples and smears) ar e transported to the counting room located in the Chemistry Lab Expansion. Equipment is available for gross alpha, gross beta, and gross gamma activity measurements and for determination of the activity levels of specific isotopes. Proportional counters and gamma spectrometers are provided. The room itself has thick shield walls to reduce instrument background radiation. The room is temperature controlled and the voltage supply is regulated for instrument stability.
Major instrumentation includes a Gamma Spectrometer for specific isotopic identification; alpha and beta counters; a scintillation detector; and a liquid scintillation analyzer for low energy emitters such as tritium.
Background and efficiency checks are performed routinely. Alpha and beta plateaus are established to determine operating voltages for proportional counters. Calibrations are based on NIST related standards for the isotopes of interest. Conventional reference standards such as Sr, C, and Ra in equilibrium with daughter activity, Cs, and Co are used for calibration when gross counting is the objective.
12.5.2.2.3 Portable Survey Instruments and Equipment
Portable survey instruments and equipment are used primarily for conducting area surveys and for monitoring personnel throughout the plant. Some portable instruments and equipment are designated for emergency use and will be stored at the Technical Support Center, readily accessible to personnel responding to an emergency.
Portable instruments and equipment for routine plant use are provided to permit alpha, beta, gamma, and neutron radiation measurements, and for obtaining samples of surface and airborne contamination. The inventory includes ion chambers, GM probes, neutron rem counters or equivalent, alpha probes, alpha and beta counters, scintillation detectors, high and low volume air samplers, filters, charcoal cartridges, and smears. Portable instruments and equipment available for emergency use include air sampling equipment with particulate filters and silver zeolite cartridges, portable ion chambers, alpha scintillation probes, energy compensated beta/gamma GM probes (for low energy photons), and portable beta/gamma geiger counters.
CHAPTER 12 12.5-4 REV. 20, SEPTEMBER 2020 LGS UFSAR
At least some of the ion chambers and GM probes have movable beta shields to enable distinguishing between beta and gamma radiation. The ion chambers are the prime devices for dose rate determinations and have beta factors specified as appropriate for the type of meter and its use.
GM probes may be used for dose rate determinations. Their application for this purpose is of value when the ability of an ion chamber to respond reliably is impaired by high humidity, high or low temperature, or very low dose rates. Certain instrument designs with GM probes have extendable arms. This feature allows the user to remain in a relatively lower radiation area while measuring high dose rates at the point of interest. This application is a good example of an effective ALARA policy.
The ranges and numbers of instruments are adequate for their intended use whether routine or emergency.
12.5.2.2.4 Personnel Dosimeters
Personnel monitoring is provided by the use of such devices as optically stimulated luminescent, direct reading dosimeters (electronic dosimeters or pocket ion chambers), or calculations from area survey data and exposure times. Personnel monitoring is provided per 10CFR20. The form of personnel monitoring depends on the type of radiation and the expected radiation level. Types of dosimetry devices change as the state of the art improves.
Devices are available and used appropriately for determining whole body or equivalent exposure, extremities exposure and skin exposure. Health Physics practices include the use of multiple badges in addition to the whole body badge as required by procedures.
A direct reading dosimeter will be used by all personnel entering the RCA who are required to be monitored for exposure per 10CFR20. The DRD is used to provide real time indications of dose received by an individual and provides estimated exposure data. As required, these dosimeters are calibrated and/or drift checked on a routine basis per procedures.
Dosimetry devices used for determining the official exposure doses are subject to extensive quality control programs. This is true whether the processing is by a contractor or onsite. Currently, DLRs are used for penetrating (gamma, neutron) and nonpenetrating (beta) radiation. Whenever neutron exposure is of concern, the current technique is to use neutron survey instruments to determine a rem dose rate, then to multiply by the stay time or a monitoring device/ DLR which has been accredited for neutron monitoring through NAVLAP N13.11 criteria.
The personnel dosimetry program is conducted by qualified personnel under the direction of the Radiation Protection Manager.
Control of personnel monitoring (e.g., whole body counter) and external (e.g., DLR system) is provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program.
CHAPTER 12 12.5-5 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.2.2.5 Miscellaneous Instruments and Equipment
Other monitoring devices are available and may be assigned to personnel or located in work areas for the purpose of alerting personnel to changes in radiation condition.
Radiation monitors that "chirp" at a rate proportional to gamma dose rate or contain an audible alarming feature are assigned to personnel as deemed necessary. These devices are of particular value when worn by individuals whose duties routinely involve entering various areas of the plant.
Changing operating conditions or the progression of the individual into the area could cause sudden changes in radiation levels. These devices alarm at preset values to alert personnel of the changed conditions as the individual walks into the area.
Alarming rate meters may be positioned in a given work area. They serve the purpose of alerting individuals working in that area of increases in radiation level above a preset value.
Friskers, personnel contamination monitors, or portal monitors are positioned at or near various change areas and plant exit points. The purpose of these devices is to control the spread of radioactive contamination. Friskers are generally used within the plant for personnel to monitor themselves at any time, especially when leaving a controlled area. Personnel contamination monitors or friskers are generally used at all primary exits from the radiologically controlled area and at the access points to approved eating, drinking, and smoking areas within the radiological controlled area. The placement of these devices may vary in support of the station's contamination control program, and other devices with the required sensitivity may be considered for use if the situation warrants. Instrumentation, such as small article monitors or friskers, is used to monitor tools, equipment and personal items. The instruments are used to ensure that items exiting the RCA are within procedurally defined limits.
Continuous air monitors are used to monitor airborne concentration at s pecific work locations.
These CAMs provide the means to observe trends or sudden changes in the airborne concentrations. They are not intended for quantitative analysis. The fixed filter-type can be used as a low volume grab air sample in that the filter medium can be removed and analyzed in more detail in the counting room.
Fixed area radiation monitors are mounted in selected locations. Each contains a gamma -sensitive detector, local indicator and local alarm, as well as indicators, alarms, and recorders in the control room. These monitors will alert personnel to unexpected or abnormally high radiation levels in these areas.
12.5.2.2.6 Bioassay
Internally deposited radioactive material is evaluated by use of whole body counting or in-vitro analysis. The whole body counter sensitivity for gamma emitting isotopes of interest will be equivalent to small fractions of isotopic organ burdens.
The bioassay program will incorporate features of Regulatory Guide 8.9 and Regulatory Guide 8.26. Conservative investigation levels are established. When investigation levels are exceeded, an investigation will be performed and, as necessary, will include consultation with independent experts for further evaluation of internal exposure consistent with Regulatory Guide 8.9 and International Commission on Radiological Protection criteria.
CHAPTER 12 12.5-6 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.2.2.7 Personnel Protective Equipment
Special protective equipment such as coveralls, plastic suits, shoe covers, gloves, head covers, and respirators are available, and are stored in various plant locations and clothing change areas.
This equipment is used to prevent deposition of radioactive material internally or on body surfaces.
Most of the plant will be kept free of contamination so that no special protective equipment will be needed. Contaminated areas are identified with posted signs. Radiological postings, work orders or RWPs (or equivalent) are the primary mechanisms for defining the equipment required to enter these contaminated areas.
A variety of combinations of protective equipment may be prescribed depending on the nature and level of the contamination. The use of engineering controls will be considered to minimize the need for protective equipment and clothing as practical. Cotton clothes may be adequate normally; but in wet areas plastic rain suits or bubble suits may be prescribed. Respirators may be required if airborne hazards exist or if surface contamination could cause an airborne hazard as defined in the implementing procedures. Sufficient quantities of NIOSH approved respiratory equipment will be provided to adequately protect workers in general and emergency conditions. This will include SCBAs, respirators (general use or welding), and all the necessary supplementary equipment required (hoses, regulators, etc.).
The guidance of Regulatory Guide 8.15 will be followed to ensure the proper selection, care, fitting, and use of respiratory protective equipment. Regular sampling and survey programs will be provided to determine the need for the appropriate respiratory protective devices. Bioassays, record keeping, and medical evaluations will be incorporated into the LGS respiratory protection program. Written procedures will be developed with the help of the operating experience at PBAPS to provide an effective and acceptable respiratory protection program. Section 12.5.3.5.3 also addresses compliance with Regulatory Guide 8.15.
Control of respiratory protection, including testing, is provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program.
12.5.3 PROCEDURES AND PRACTICES
Health physics procedures are classified according to specific concerns. Thus, there are operating procedures, analytical procedures, emergency procedures, and surveillance test procedures.
Health physics is a consideration in other plant procedures as well. This section describes the fundamental procedures applicable to health physics operations. The procedures and methods of operation which, in combination with licensee policy and training, ensure that radiation exposures will be ALARA are those which implement the controls and prescribe the use of equipment described in this section and in Section 12.5.1.
12.5.3.1 Radiation Surveys (Area Surveys)
Area survey procedures describe the purpose and techniques of detecting the presence of and measuring the level of radiation and contamination. Contamination may be on surfaces or airborne. Area surveys are conducted throughout the plant. Such surveys may be routine or may be related to specific jobs upon request. An area survey may be performed before, during, and after various work activities. Area surveys are performed by health physics technicians or other trained and qualified personnel (i.e. Advanced Radiation Workers). Control of radioactivity sampling (air, surface, liquids) is provided by procedures that are responsive to the appropriate portions in the Quality Assurance Program.
CHAPTER 12 12.5-7 REV. 20, SEPTEMBER 2020 LGS UFSAR
12.5.3.1.1 Radiation Detection
The primary instrument for beta/gamma dose rate measurements is an ion chamber.
Circumstances may require the use of other instruments to determine dose rates. GM probes may be used for low radiation levels or where environmental conditions (temperatures, humidity) cause erratic responses from ion chambers.
Surveys for neutrons are performed by instruments designed for that purpose. A rem counter or equivalent that has the ability to measure neutron dose rate in rem per hour is the preferred neutron measurement instrument. However, other instruments or devices for determining neutron energies may also be used. For example, moderated and unmoderated boron triflouride (BF 3) probes may be used to detect the presence of low and high energy neutrons. Some instruments are designe d with extension arms. These types of instruments may be used in high dose rate areas to provide distance from the source to the technician, thus reducing personnel dose.
12.5.3.1.2 Surface Contamination Detection
A variety of techniques are necessary to detect and measure radioactive contamination.
Procedures describe the use of smears (e.g., small paper discs) and swipes (e.g., large area survey) to wipe a surface to pick up removable contamination. Smears and swipes are analyzed using portable survey meters and/or counting room equipment. Fixed contamination is determined by scanning a surface with portable survey meters. Equipment is available for alpha and beta/gamma activity measurements.
12.5.3.1.3 Airborne Contamination
Airborne contaminat ion is determined by using air samplers to draw a known volume of air through a filter paper or charcoal cartridge. A charcoal cartridge is used with the filter paper where iodine is of concern. The filter paper is analyzed by gross beta/gamma count and/o r gamma spectrometry.
Gamma spectroscopy is also performed on the charcoal cartridges. The gamma spectrometry identifies the particulate and iodine isotopic activity. DAC-hours shall be tracked in accordance with Heath Physics procedures. Gross beta/ga mma count data is used to judge the need for filter paper gamma spectrometry. High volume air samplers and low volume air samplers having nominal sample rates of 25 scfm and 1 scfm, respectively, are available. The high volume air sampler is used primarily to obtain grab samples rapidly before, during, and after work activities.
The low volume air sampler is used primarily to obtain the average air concentration for the work period.
12.5.3.1.4 Survey Frequency and Techniques
The frequency and extent (scope) of area surveys is a function of dose rate in the area, accessibility to the area, and nature of the area in plant operations. For example, contamination checks in control rooms and eating areas are performed nominally on a daily basis; seldom entered high radiation areas are surveyed as infrequently as monthly, or only as needed for entries. As part of the ALARA program, the performance of area surveys are coordinated so that, wherever feasible, routine area survey data is used to perform RWP work. This practice avoids duplication and reduces exposure to technicians.
Area survey data are usually obtained or known before work starts in an area. Resurveys may be performed in the area during the job if the work activity is prolonged. Other conditions for
CHAPTER 12 12.5-8 REV. 20, SEPTEMBER 2020 LGS UFSAR
resurveys would be if the work activity or other plant operation caused changed conditions.
Surveys may also be performed at the completion of the work activity.
Depending on the survey results, the area surveyed is roped off and posted ap propriately to alert personnel to the radiation conditions and requirements for entry.
Procedures for area surveys describe the use of instruments, effective survey techniques, and documentation of data. The specifications for clean, radiation and high radiation areas are defined.
Various levels of surface and airborne contamination are established as guidelines for prescribing protective equipment such as clothing and respirators. Procedures include consideration for potential as well as actual radiation hazards.
12.5.3.2 Radiation Work Permits (or Equivalent)
Where radiation dose rates, airborne concentrations, or surface contamination levels exceed procedural limits, an RWP (or equivalent) is required for work. Health Physics will evaluate the radiological conditions associated with the work to be performed. Health Physics will specify the appropriate protective clothing, equipment, and monitoring required including dosimetry by use of radiological postings, work orders, or RWPs. Area survey fre quency is established by Health Physics. All personnel performing work under a particular RWP (or equivalent) must be familiar with the permit conditions and must sign a RWP (or equivalent) compliance sheet. Information documented from RWP (or equivalent ) entries include name, time in and out, and estimated exposure for entry. The RWP (or equivalent) computer may be used in lieu of a compliance sheet to document this information. Health Physics may terminate an RWP (or equivalent) if radiation conditions change.
Health Physics supervision selectively reviews RWPs and survey documentation which are then filed. RWPs (or equivalent) serve as a source of data for dose comparison on repeat jobs. They can be used to determine the effectiveness of ALARA ef forts.
12.5.3.3 Handling and Storage of Radioactive Material
Health Physics personnel are notified of the receipt of radioactive material, and of intended shipment of radioactive material. This is done so that required surveys can be performed and to verify that correct labeling and placarding has been accomplished.
Calibration sources for radiation instrumentation and sources used to prepare secondary standards are stored in a source vault. This vault is capable of being locked. The lock is under the control of the Radiation Protection Manager.
Small quantities of sealed or unsealed sources may be stored for convenience in shielded cabinets, caves, or safes. Such sources are used locally in the chemical laboratories, counting room, or when response -checking instruments throughout the plant.
Spare instruments containing built-in sources and slightly contaminated equipment intended for reuse may be stored at times in the warehouse. Such items are clearly identified as containing radioactive material.
12.5.3.4 Controlling Access and Stay Time
The plant is surrounded by security fencing. Entrance must be via the guardhouse. Security procedures are applied by the guards at this point to identify each individual and to determine
CHAPTER 12 12.5-9 REV. 20, SEPTEMBER 2020 LGS UFSAR
their purpose for entry. Security and dosimetry badges are assigned. Escorts are provided to satisfy procedural requirements. Entrance to the RCA is via the Health Physics Access Point.
Health Physics procedures are applied at this point to set exposure limits and ensure Health Physics Training of each individual prior to entry.
12.5.3.4.1 Access to Radiation and High Radiation Areas
Radiation areas are identified by posted radiation signs. Signs are used to define requirements for entry. Where appropriate, yellow and magenta or black rope or tape is used as a barrier to prevent access or to divert personnel to a specific control point for access. RWPs are used to define the exposure limits, document entry and exit, and record estimated exposure for each individual.
Procedures describe the purpose and application of the RWPs. Administrative guides for personnel exposure are established by procedure. These guides are set at values less than the exposure dose limits in 10CFR20. Variations from these guides must be requested. Procedures describe the steps for approval of dose extensions. Additionally, positive control is exercised over each high radiation area by use of barricades, conspicuously posting the area, and requiring an RWP for entrance. Areas greater than 1000 mR/hr are controlled by using lock doors or gates. Exceptions to this practice are permitted per Technical Specification 6.12. Issuance of the keys are controlled by Health Physics. Each key used and the individual using the key are recorded. Certai n work activities having high dose rates and short stay time may be monitored directly by Health Physics Technicians to prevent personnel from inadvertently exceeding the prescribed dose for the work.
Personnel are advised to observe the reading on their direct reading dosimeter frequently.
12.5.3.4.2 Contamination Control
Contaminated areas are conspicuously identified to prevent inadvertent access. Floor coverings, called step off pads, typically are used to access these areas. Appropriate procedures give guidance for the selection and application of protective clothing and respirators under various specific conditions. Personnel requiring access to contaminated areas must determine the radiological controls and requirements via Health Physics briefings, radiological postings, work orders or RWPs.
As personnel leave the work area, they remove the protective clothing and respirators before stepping across the step-off-pad. Personnel must monitor themselves to assure that no contamination has been transferred to their bodies or clothing.
Tools or equipment to be removed from a contaminated area are surveyed or contained prior to removal from the area. Items suspected of or having loose surface contamination must be placed in a container (e.g., a plastic bag), surveyed, and tagged at the step-off-pad. The tools may be placed in storage, may be taken to another contaminated area for reuse, or they may be designated for appropriate storage, disposal, or decontamination. Tools, equipment and personal items are surveyed prior to removal from the RCA. Health Physics procedures describe the survey techniques and limits.
The presence of radioactive contamination, whether surface or airborne, inhibits mobility of personnel around the plant. Protective equipment that must be worn creates inconveniences and introduces other factors that affect performance. For these reasons, plus the obvious external and internal radiation hazards, decontamination is initiated judiciously to confine the contamination to as small an area as practicable and/or to reduce the contamination levels to minimize protective requirements. Special coatings that aid in decontamination are applied to walls and floors. The ventilation flow pattern is from clean areas to contaminated areas. Process equipment is isolated in various cavities or cells. These cells are vented in a controlled manner, usually through filters to
CHAPTER 12 12.5-10 REV. 20, SEPTEMBER 2020 LGS UFSAR
effluent stacks that monitor flow rate and radioactivity. Highly contaminated equipment drains are piped to sumps to avoid the use of floor drains and attendant spillage of fluids on the floor.
Control of radioactive contamination measurement and analysis and radioactive contamination control are provided by procedures that are responsive to the appropriate portions of the Quality Assurance Program.
12.5.3.5 Training
See Section 13.2.
12.5.3.5.1 Respiratory Fit Test
A functioning Respiratory Protection Program, which meets the requirements of 10CFR20. and Regulatory Guide 8.15 (October 1976), permits the application of protection factors to select the appropriate type of respirators. This can only be done for individuals who have received training in the use of the respirators, have been tested for the validity of fit for the type of respirators intended for their use, and who have been medically certified. Procedures are established that describe the technique, fit test frequency, and define acceptance criteria.
12.5.3.5.2 Compliance with Regulatory Guides
Regulatory Guide 8.2 - Section 12.1 describes the general ALARA program and policies. As specific procedures and the LGS ALARA plan are developed, Regulatory Guide 8.2 will be used for guidance.
Regulatory Guide 8.7 - Exposure records will be kept in accordance with 10CFR20 and maintained until the NRC authorizes their disposition. As the procedures are developed, the guidance of ANSI N13.6 (1969) as endorsed by Regulatory Guide 8.7 will be followed.
Regulatory Guide 8.8 - Sections 13.1, 13.2, 13.5, 12.1, and 12.5 address the management commitment, organization responsibilities, authority, training, procedures, and review techniques which implement Regulatory Guide 8.8 for LGS operation.
Regulatory Guide 8.9 - As stated in Section 12.5.2.2.6, the bioassay program will incorporate features of Regulatory Guide 8.9. When conservative investigative levels are exceeded, a consultant will assist in a more detailed evaluation of internal exposure. These procedures and practices will follow the guidance of International Committee of Radiation Protection publications and Regulatory Guide 8.9.
Regulatory Guide 8.10 - The development and implementation of a formal ALARA program will follow the guidance of Regulatory Guide 8.8 which is a nuclear power plant specific reference of Regulatory Guide 8.10.
Regulatory Guide 8.13 - Instruction to workers concerning prenatal radiation exposure will be given as part of the General Employee Training Program. This program will provide all employees with the information that is identified in Regulatory Guide 8.13.
Regulatory Guide 8.15 - The development and implementation of a respiratory protective equipment program will follow the guidance of Regulatory Guide 8.15.
Regulatory Guide 8.26 - The bioassay program will incorporate features of Regulatory Guide 8.26.
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Regulatory Guide 1.8 - LGS is in compliance with this guide to the extent discussed in Sections 13.1 and 13.2. The licensee has implemented ANSI/ANS 3.1-2014, section 5, which is endorsed by Regulatory Guide 1.8.
Regulatory Guide 1.16 - LGS Technical Specifications were based on NUREG-0123, Revision 4, "Standard Technical Specifications for General Electric BWRs," as described in Chapter 16.
Reporting of operating information will be in accordance with the Technical Specifications,
Regulatory Guide 1.33 - LGS will follow the guidance of Regulatory Guide 1.33 which endorses/modifies ANSI N18.7 (1976) with certain alternate approaches which are addressed in detail in Section 17.2.
Regulatory Guide 1.39 - LGS is in general conformance to the guidance of Regulatory Guide 1.39, which endorses/modifies ANSI N45.2.3 (1973). LGS will comply with Regulatory Guide 1.39 with certain alternate approaches which are outlined in detail in Section 17.2.
Regulatory Guide 1.97 - LGS is in compliance with Regulatory Guide 1.97 in regard to Health Physics and Chemistry Laboratory and Survey Equipment.
CHAPTER 12 12.5-12 REV. 20, SEPTEMBER 2020 LGS UFSAR
Table 12.5-1
HEALTH PHYSICS INSTRUMENTATION
NORMALLY REQUIRED QUANTITIES
INSTRUMENT SENSITIVITY/RANGE USE NOW
Gamma Spectrometer 10 minute count (Ci): Gamma scans 4 Co-60, Cs-134, Cs-137, I-131, - 1x10-4 Zn-65 Noble Gases - 1x10-3
Alpha/Beta Counter Beta Efficiency 10%-50% Alpha/beta 2 (E dependent) discrimination
Whole Body Counter System <1/20 of the International Internal 1 Committee of Radiation dosimetry Protection Publication 30 Annual Limit of Intake
NORMALLY REQUIRED QUANTITIES
RADIOSURVEILLANCE INSTRUMENTS SENSITIVITY/RANGE USE NOW
RO-2 0-5000 mR/hr Field survey 10
RO-2A 0-50 R/hr Field survey 15
E-520 0.2-2000 mR/hr Field survey 5
CHAPTER 12 12.5-13 REV. 19, SEPTEMBER 2018 LGS UFSAR
Table 12.5-1 (Cont'd)
NORMALLY REQUIRED QUANTITIES
RADIOSURVEILLANCE INSTRUMENTS SENSITIVITY/RANGE USE NOW
E-140N 0-50,000 cpm Field survey 15
HP270 probe 0.1 mR/hr-10 R/hr Field survey 15
HP210AL probe - Field survey 15
HP210T probe - Field survey 15
Teletector 0-1000 R/hr Field survey 3
HP220A probe - Field Survey 1
HP260A probe - Field Survey 10
EC4-X (& cables) 0.01 mR/hr-10,000 R/hr Portable area monitors 10
DA1-6 probe (for EC4-X) Portable area monitors 10
PRM-6 with AC-3 0-500,000 cpm Alpha surveys 0
PNR 4 0-5000 mR/hr Neutron surveys 4
HP280 - 3 in. sphere Neutron surveys 0
RO7 + BH midrange detector 0-200 R/hr; High range ion chamber 2
+ BH high range detector, 0-20,000 R/hr
+ RX5 - 5 ft extender,
+ 2 (15 ft) cables
Underwater probe 0-5000 R/hr High range under water 2 5-5000 mR/hr
Rm 20 0-500,000 cpm Frisking stations 25
Alpha + Beta CAMs Air monitoring 3
Low Volume air samplers - Air monitoring 25
SAC-4 Alpha Survey 3
CHAPTER 12 12.5-14 REV. 19, SEPTEMBER 2018 LGS UFSAR
Table 12.5-1 (Cont'd)
NORMAL REQUIRED QUANTITIES
RADIOSURVEILLANCE INSTRUMENTS SENSITIVITY/RANGE USE NOW
High Volume air samplers - Air monitoring 10
Portal Monitors Approx. MDA 1 mR Contamination control 5
Pocket Ion Chamber:
Charging Units Exposure control as needed 0-200 mR 0-200 mR 0-500 mR 0-500 mR 0-1500 mR 0-1500 mR
Electronic Dosimeter Exposure Control as needed
Portable MCA + HP Ge 1 detector + 24 hr dewar
E-530N 0.02-20 R/hr Field surveys 2
Shepherd 142-10 0-1,600 mR/hr Calibration 1 panoramic calibrator
Shepherd 89 calibrator 0.01 mR/hr-1200 R/hr Calibration 1
Condenser R meter - Calibration 1
CHAPTER 12 12.5-15 REV. 19, SEPTEMBER 2018