ML19312A215
| ML19312A215 | |
| Person / Time | |
|---|---|
| Site: | 05000495 |
| Issue date: | 11/29/1978 |
| From: | NEW YORK STATE ELECTRIC & GAS CORP., STONE & WEBSTER, INC. |
| To: | |
| References | |
| NUDOCS 7909060008 | |
| Download: ML19312A215 (109) | |
Text
SWESSAR-P1 CHAPTER 12 RADIATION PROTECTION LIST OF EFFECTIVE PAGES Page, Table (T) , Amendment Page, Table (T) , Amendment or Figure (F) No. or Figure (F) No.
12 -a 28 12-i thru iii 17 12-iv Orig 12.1-1 thru 2A 4 12.1-3/4 2 12.t-S thru 6A 7 12.1-7 thru 14 17 12 .1.14 A 24 12.1.15 24 12.1-16 thru 19 7 12.1-20 16 T12.1-1 Orig T12.1-2 Orig T12.1.2-1 Orig T12.1-3 Orig T12.1.4-1 17 T12.1.6-1 16 T12.1.6-2 (3 sheets) 16 F12.1-1 (W)
(5 sheets) 1 F12.1-1 (BSW)
(5 sheets) 2 F12.1-1 (C-E)
(5 sheets) 3 12.2 -1 17 12.2-2,2A 19 12.2-3 thru 8 17 T12.2.3-1 (2 sheets) 19 T12.2.3-2 thru 5 19 T12.2.3-6 pi) 17 T12.2.3-6 (W-3S) 17 T12.2.3-6 (BSW) 28 T12.2.3-6 (C-E) 17 T12.2.4-1 9 T12.2.6-1 19 F12.2-1 (W) 17 F12.2-1 (W-3S) 17 F12.2-1 (BSW) 19 F12.2-1 (C-E) 17 F12.2-2 12 12.3 -1 13 12.4-1 2 12-a Amendment 28 8/6/76 609 25]
SWESSAR-P1 CHAPTER 12 RADIATION PROTECTION TABLE OF CONTENTS Section Page 12.1 SHIELDING 12.1-1 12.1.1 Design Objectives 12.1-1 12.1.2 Design Description 12.1-1 12.1.2.1 Primary Shielding 12.1-13 12.1.2.2 Secondary Shielding 12.1-14 12 . 1.3 Source Terms 12.1-16 12.1.4 Area Monitoring 12.1-17 12.1.5 Operating Procedures 12.1-18 12.1.6 Estimates of Exposure 12.1-18 12.1.7 Interface Requirements 12.1-20 Reference for Section 12.1 12.1-20 12.2 VENTILATION 12.2-1 12.2.1 Design Objectives 12.2-1 12.2.2 Design Description 12.2-2 12.2.3 Source Terms 12.2-2 12.2.4 Airborne Radiation Monitoring 12.2-3 12.2.4.1 Containment Atn.osphere Radiation Monitoring System 12.2-4 g 12.2.4.2 Annulus Building Monitors 12.2-4 12.2.4.3 Solid Waste and Decontamination Building Monitor 12.2-5 12.2.4.4 Fuel Building Monitors 12.2-5 12.2.4.5 Ventilation Vent Monitor 12.2-5 12-i BUY 's- Amendment 17 9/30/75
SWESSAR-P1 TABLE OF CONTENTS (COh"f)
Section Page 12.2.4.6 Ventilation Vent High Range Monitor 12.2-4 12.2.4.7 Ventilation Systems Multisampler 12.2-4 12.2.4.8 Containment Purge Air Monitors 12.2-5 12.2.4.9 Control Room Air Intake Monitors 12.2-5 12.2.5 Operating Procedures 12.2-5 12.2.6 Estimates of Inhalation Doses 12.2-5 17 References for Section 12.2.6 12.2-6 12 .3 HEALTH PHYSICS PROGRAM 12.3-1 12.4 RADIOACTIVE MATERIALS SAFETY 12.4-1 O
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12-ii Amendment 17 9/30/75
SWESSAR-P1 LIST OF TABLES Table 12.1-1 Radiation Zone Designations 12.1-2 Typical Locations for Designated Radiation Zones 12.1.2-1 Materials Used f or Source and Dose Rate Calcula-tions 12.1-3 Design Basis Accident, Accident Conditions 12.1.4-1 Area Radiation Monitor Locations and Ranges 12.1.6-1 Sumnury of Estimated Annual In-Plant Exposure, Nor~ mal Operations 11 12.1.6-2 Estimated Annual In-Plant Exposure, Normal Operations 12.2.3-1 Assumptions Used in the Calculation of Airborne Concentrations in Various Buildings 12.2.3-2 Annual Average Radioactivity Leakage into Con-tainment Structure Atmosphere 12.2.3-3 Annual Average Radioactivity Leakage into Turbine Building Atmosphere 12.2.3-4 Annual Average Radioactivity Leakage Rate into Annulus Building Atmosphere 12.2.3-5 Annual Average Radioactivity Leakage Rate into Fuel Building Atmosphere 12.2.3-6 Expected Radioactive Airborne Concentration Inside Major Plant Buildings 12.2.4-1 Airborne Radiation Monitors 12.2.6-1 Estimate of Inhalation and Whole Body Dose Rates in Ma jor Plant Buildings 17 12-iii Amendment 17 f6Q' g r .i CJ" 9/30/75
SWESSAR-P1 LIST OF FIGURES Fiqure 12.1-1 Design Basis Radiation Zones for Shielding, Sheet 1 12.1-1 Design Basis Radiation Zones for Shielding, Sheet 2 12.1-1 Design Basis Radiation Zones for Shielding, Sheet 3 12.1-1 Design Basis Radiation Zones for Shielding, Sheet 4 12.1-1 Design Basis Radiation Zones for Shielding, Sheet 5 12.2-1 I-131 Concentration in Containment vs Time with One or Two Recirculation Filters in Operation 12.2-2 Containment Atmosphere Airborne Radiation Monitoring System O
O 12-iv C 6 (3 )
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SWESSAh-P1 CHAPTER 12 RADIATION PROTECTION 12.1 SHIELDING 12.1.1 Design Obiectives Radiation protection, including radiation shielding, is designed to ensure that the criteria specified in 10CFR20 are met, that exposure of personnel is as low as practicable during normal operation, and that the guidelines specitied in 10CFR100 are met during accident conditions.
Maxilaum dose rates are bared on the expected frequency and duration of occupancy. Values of maximum dose rates are upper limits and are based on conservative calculations. Average operating dose rates are expected to be much lower than the maximum dose rates reported. Occupancy time and dose rates are such that no personnel receive more than a fraction of the radiation allowed by 10CFh20. Average dose rates in the visitor areas are less than 0.1 mrem per hour. Maximum dose rates in the various plant radiation zones are listed in Table 12.1-1.
Typical locations of these zones are listed in Table 12.1-2.
Design doses for accident conditions are listed in Table 12.1-3.
Plan views at various elevations throughout the plant are shown in Fig. 12.1-1. The radiation zones used as a design basis for shielding analysis are indicated by dif f erent colors.
12.1.2 Design Description The design approach is similar to that described in the topical report " Radiation Shielding Design and Analysis Approach For Light Water Nuclear Reactor Power Plants," dated April 1974, designated Stone G Webster Engineering Corp. Report RP-8.
The numerical values for dose rates for a typical PWR f ound in section 4 of RP-8 are indicative of the general radiation levels found in the SWESSAR plant. These values demonstrate that the shielding approach and analytical methods for the SWESSAR plant, which are described in detail in Sections 2 and 5 of RP-8, result in in-plant exposures from direct radiation which are as low as practicable. 4 Dose rates given in RP-B are us ed judiciously to determine relative radiation levels around similar equipment and to estimate preliminary shielding requirements. The radiation levels for the SWESSAR plant for a given piece of equipment at a given location are expected to be approximately 50 percent higher than the radiation
@ levels for the comparable equipment at comparable locations.
RP-8 reference pla nt for The basis for this 12.1-1 9E7 Amendment 4
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SWESSAR-P1 it the oltrerence in the power levels. The shielding requirements are not expected to ditter significantly f rom those determined in hP-8 since the calculated values for RP-8 are 4 generally well below the levels needed to achieve the design dose rates shown in 119 12.1-1.
hegulatory Guide 8.8 (Section 3A.2-8.8) requires s pecif ic features in the tacility and eq uipmen t design to limit radiation exposure to as low as practicable. The tollowing features are provided in the FWh Standard Plant arrangement. Regulatory Fositions 3.a through 3.r are restated for convenience.
- a. Equipment which may require servicing is aesigned and located to minimize service time.
It is assumed that anything with moving parts will require servic ino . In the annulus building all nonradioactive equipment, such as reactor plant component cooling system components and components processing the waste evaporator distillate, is located outsice cubicles in radiation zone IV or less.
hadiation zones are shown in Fig. 12.1-1. In most cases this equipment is located in radiation zone III. In the containment structure nonradioactive equipment requiring servicing is located in zone V areas, if possible.
Execptions are those items directly attached to the reactor coolant system components such as the RC pump motor cooling equipment and the equipment support snubbers.
Radioactive equipment requiring servicing is located in individually shielded cubicles. The cubicles are designed such that the contribution f rom the principal source in adjacent cubicles coes not exceed zone IV dose rate levels. (The actual dose rate in the cubicle is a function of the ettectiveness of draining and flushing the equipment to be serviced and the number and manner or cubicle penetrations.)
The service time for radioactive equipment is minimized by tirst providing zone III access to the cubicle entrance. .'he circumferential outer corridors and the radial corridors in the annulus building are zone III.
Almost all cubicles are accessible from this corridor system with the exception of the letdown heat exchanger complex. This complex has a zone IV corridor between the two zone III radial corridors. The zone III access corridor system is sized to allow forklift trucks access on the circumferential corridor and to allow hand carts and dollies access down the radial corridors with the r.inimum corridor width being 4 ft. Cubicle access opening size is dictated by equipment removal and by maintenance equipment requirements. Pump cubicle l
12.1-2 Amendment 4 hh9 2)b 11/1/74
SWESSAR-P1 openings (for horizontal pumps) are of sufficient size to skid the pump through the labyrinth. An exception is the charging pump cubicles. For other cubicles containing equipment requiring servicing, the openings are sized to allow the passage of the component, with the personnel access opening a minimum of 30 in. wide and 7'-6" high. -
Cubicle size is designed to allow a mininum of 2 ft clearance around equipment requiring service and a minimum of 3 ft adjacent to the equipment on one side near the cubicle access opening. The equipment service requirements for pull space and laydown space are provided within the cubicle without dismantling any piping other than that directly connected to the equipment. In certain cases where the equipment is always removed for dismantling such as the heat exchangers, there is no laydown space within the cubicle, other than that required for the channel cover.
@ Amendment 4 12.1-2A 11/1/74 fa orn bO/ LJ#
SWESSAR-P1 Layout in the cubicle also minimizes servicing time by placing the mechanical part of the equipment adjacent to the pipeway and the electrical portion towa rd the uninhibited access. Motor power boxes and other terminal boxes do not block access and are separated from the radioactive piping it possible. Platforms are provided where access is required for servicing. Stairs are used if possible to these platforum.
- b. Instruments requiring in situ calibration are located in the lowest practicable radiation fields.
The instrument racks are located in the annulus building above the cubicles or in the radial corridors. They are in the zone III areas. The instrument racks in the containment are located outside the cubicles in a zone V area.
- c. Equipment and components requiring servicing are located in or designed to be movable to the lowest practicable radiation fields.
As indicated above, radioactive equipment requiring servicing is located in shielded cubicles with access openings sized for equipment removal. In the case where a vertical heat exchanger or a demineralizer vessel requires repair due to failure of nonmoving parts, removal is accomplished through pluas in the cubicle ceiling. All equipment located on the top half of the annulus building can be transferred by means of the monorail system to the solid waste disposal building 2
'rane which transrers it to the decontamination area of de potentially radioactive machine shop. Equipment not directly under the monorail system can be moved on hand trucks to the freight elevators and to monorail system.
Equipment located on the bottom half of the annulus building is also served by a monorail system and freight elevator capable of handling equipment to the rolling steel door. At the rolling steel door, a truck loading space is provided. Much of the highly radioactive equipment is located on the upper half which has direct access to the solid waste and decontamination building.
Equipment in the containment structure is accessible to the polar crane for removal to the operating floor where it can be transferred through the equipment hatch to the annulus bux1 ding.
- d. Valves, valve packing, and gaskets will be selected to minimize leakage and spillage of radioactive materials.
Valve selections will be made based on abest product" available considering valve type, seat materials, and 12.1-3 Amendment 2 nr4. .g [- [; @ 8/30/74
SWESSAR-P1 service conditions. To minimize valve leakage, Stone &
Webster emphasizes a conprehensive owner maintenance program .
- e. Penetrations of shielding and containment walls by ducts and other openings are designed to minimize exposure and shield installation specifications limit void content.
Piping penetrations through shield walls in the annulus building are either into adjacent cubicles, into the radioactive pipe chases " adioactive piping, or into the radial corridors fo: .onradioactive piping. All piping penetrations of tne shield walls from zone III areas prevent any line of sight from zone III to the significant source. This is done by penetrating the radial walls at a location usually high and in the corner offset from the cource. Most piping penetrations between cubicles are designed to prevent any lL ' of sight from the significant source to the adjacent cubicle. Exceptions are made to this criterion by the recirculation line from the boron or waste evaporator to the circulating pump and to the reboiler. The interconnecting piping itself is a significant source and does not lend itself to being run at an angle through the separating wall. Another area where this criterion is not applied totally is in the piping from the filter and demineralizer cubicle to the valve cubicles below. In this case the valves are operated using extension stems to reach outside the cubicle.
During maintenance of these valves, the source is reduced as low as possible by first flushing out the 2 resin or removing the filter. The dose during maintenance is also limited to the hands of the personnel performing maintenance by providing suf ficient space in the cubicle to work out from the direct line of sight.
Electrical penetrations are made into shielded cubicles at angles which prevent direct line of sight to the significant source.
Instrument tubing penetrations are made into shield cubicles at angles which prevent direct line of sight to the significant source.
Ventilation ducts penetrations are made into shield cubicles at the highest possible elevation and at a location to minimize any direct line of sight. Any direct line of sight penetration is provided with shields located either inside or outside the cubicle penetration.
12.1-4 e ,c , Amendment 2 UO/ / '; 8/30/74
SWESSAR-P1 Shield walls are normally a part of the building structure and often serve as internal suppo rts in addition to shielding. Specifications for concrete or other types of shielding include requirements for minimizing void content.
- t. kadiation sources and occupied areas are separated if possible (in particular, pipes or ducts containing potentially highly radioactive fluids do not pass through occupied areas).
kadioactive piping (e .g . , process piping that carries radioactive materials) is located behind shielding to minimize radiation exposure to operating personnel.
Pipe tunnels, chases, or shafts as required properly segregate and shield radioactive piping trom nonradioactive piping, personnel passageways, and operating areas. The shielding thicknesses are adequate to shield operating personnel f ror single or multiple pipe lines, depending upon the piping arranaement. All radioactive piping 2-1/2 in. and larger is specifically located on piping drawings and is accurately placed in the positions indicated on the drawings when the piping is installed in the field. Piping 2 in. and smaller is shown field run on, drawings and for such piping field personnel are responsible for preparing isometric drawings to reflect both existing piping conditions and potentially radioa ctive systems. The drawings are I reviewed by the responsible design engineers prior to pipe installation. Regardless of size, radioactive piping is installed behind shielding as needed, as shown on piping drawings, to minimize radiation exposure to operating personnel.
The radioactive piping and duct work run circumferentially around the annulus buildino in radioactive pipeways and in the ventilation fan floor eleve'. ion respectively. The cubicles are radial off these circumferential runs which allows piping to be brought into them without going through any occupied areas. The piping to the waste disposal building does cross the access walkway. In this case sufficient shielding is provided to allow this piping to pass overhead on the -(14 '-6") floor elevation into shielded pipe tunnels in the waste building.
The radioactive piping in the containment is in most cases run directly out to the annulus building at elevations (-14 to 0) where it is dropped through the floor into the radioactive pipeway below. The penetration area is radioactive; however, access to every penetration isolation valve is possible without
@ exposure to o ther radioactive penetrations and piping 12.1- 5 ,ua g yd Amendment 7 bU' 2/28/75
SWESSAR-P1 due to the penetrations being spread around the annulus building and being shielded from the pipeway. Shield walls between radioactive penetrations are provided.
Physically locked barriers are provided for areas having radiation levels in excess of 100 mrem /hr.
g Pracautions are provincd (1) to minimize the spread nf contamination, and (2) to iacilitate decontamination in the event spillage occurs.
The tanks such as the boron recovery tanks, the high level waste drain tanks, the evaporator bottoms tank, the spent resin hold tank, and the primary drain tank are placed in shielded cubicles designed to contain the tank contents. These cubicles are pitched to a pit for the use of a portable sump pump. This arrangement contines any spillage. A level detector for these pits alerts the operators to the spillage situation. Other tanks with dikes minimize the spread of radioactivity including the low level waste drain tanks, the evaporator test tanks (the primary grade water tanks) ,
the resin recycle surge tank, the camponent coolina surge tanks, and the boric acid tanks. In addition, the refueling water storage tank and the auxiliary feedwater storage tank are also in cubicles to prevent the loss of the water for ESF systems.
The other tank cubicles such us the resin tanks and the filter tanks are in individual cubicles with many floor penetrations below. These penetrations are provided with raised floor sleeves to prevent leakage from the tank ir,to the valve area below. (Leakage of a pipe to a tank weld is carried along the pipe to the valve below but this serves to define the leak.) The tank cubicle has its floor drain piped to the nearest annulus building sump.
Other cubicles containing items such as pumps, heat exchangers, and valve stations are served by the vent and drain systems.
The evaporator bodies are not provided with this dike protection due to the necessity of locating the bottom cones through the floor and piping to the circulating pumps located below them. In this arrangement a leak in the evaporator passes to the pump cubicle where it is piped through the floor drains to the sumps.
The annulus' building floor is pitched from the clean outer circumference toward the sumps in the radioactive pipeway and from the penetrations toward the sumps. Any liquid contamination not handled by the floor drainage 12.1-6 Amendment 7
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SWESSAR-P1 within the cubicles flows radially inward to the subsequent cubicle through limbers.
12.1-6 A Amendment 7 m ( p' 2/28/75 t (. O t0 OU /
SWESSD -P1 The airborne c antamination is kept from spreading in the annulus build ag by the ventilation system. Clean air enters the co partment at the outer circumference and is withdrawn at the inner circumference for each radial equipment complex.
Provisions made to facilitate decontamination in the event of spillage are the concrete finish surface design and the availability of demineralized water.
Demineralized water hose stations are provided at each radial corridor and at other areas as required to allow flush water to be available to each cubicle in the annulus building.
- h. Interior surfaces as well as layout of ducts and pipes are designed to minimize buildup of contamination.
Interior surfaces of systems in radioactive liquid service are of stainless steel or other corrosion resistant alloy.
The piping systems are routed to avoid unnecessary sharp bends by careful selection of the elevation between points, and attempting to run at one or two elevations between these poir.tr. Pockets and low points are also avoided.
Resin piping is run with butt-welded connections and with long radius bends and laterals.
All radioactive waste systems utilize either ball or diaphragm valves.
The valve stations in shielded cubicles are designed to minimize the buildup of crud in these areas by minimizing the number of pockets and stagnant vertical legs.
- i. Systems which may become contaminated are designed to include provisions for flushing or remote chenical cleaning prior to servicing.
The liquid waste systems have permanent pipe flushing connections. All heat exchangers are provided with chemical cleaning connections which are hooked up prior to servicing.
- j. The ventilation system is designed to ensure control of airborne contaminants, especially during maintn .e operations when the normal air flow patterns - .y be disrupted (e.g. open access portals) . The vent.1ation i systems as described in Section 9.4 are designed 17 n/ C Amendment 17 12.1-7
/3U n t} LOJ 9/30/75
SWESSAR-P1 17 consistent with goal and principle that occupational radiation exposures will be as low as practicable.
The annulus building ventilation system is designed to bring air from uncontaminated areas into the potentially contaminate d cubicles and to exhaust frm the cubicles.
A description of the annulus building ventilation system is provided in Section 9.4.2.
The containment atmosphere filtration system is designed to remove airborne contaminants when necessary. A description of the containment structure ventilation system is provided in Section 9.4.5. Provisions for
- 7 portable ventilation equipment are described in Section 9.4.
- k. Wherever practicable, radiation and airborne contamination monitoring equipment with remote readout is included in areas to which personnel normally have access (where special conditions warrant, portable instrumentation may be substituted) .
Area and airborne radiation monitoring ensures that any substantial abnormal radioactivity release is detected within a ressonable time.
Area radiation monitors are located to serve the waste evaporator control board area and the boron evaporator -
degasifier control board area. Other areas also permanently monitored are the sample room, the equignent doors frm the annulus building, from the fuel building, and from the solid waste building. Other doors frm the annulus building are for emergency egress only and are one-way alarmed doors. The personne.1 access to the annulus building, fuel building, weste building, and con % inment is through the monitor station in the health physics building adjacent to the 29 ft floor of the annulus building.
Airborne radiation monitors are provided for the containment, the annulus building first level, the annulus building second level, the annulus building third and fourth level combined, r.he solid waste and decontamination building, and the fuel building.
The requirement for permanent monitors at other stations is not considered necessary due to the layout of the building allowing essentially shielded access to all radioactive equipment and piping areas. A description of the area and airborne radiation monitors is provided in Sections 12.1.4 and 12.2.4 respectively.
O i
12.1-8 r: ; c,V
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SWESSAR-P1
- 1. The ventilation system is designed for easy access and service to keep doses as low as practicable during alterations, maintenance, decontamination, and filter changes.
The ventilation systems are arranged to pre vide the same shielded access to the charcoal and hich efficiency particulate air filter (HEPA) trains as is provided for other radioactive equipment. Adequate space is provided on all sides and top to add, test, and remove charcoal and HEPA filters. The addition of fresh charcoal can be made by means of a cart with the barrels o f charcoal and air vacuum pump skid-mounted. This cart can be moved into the filter cubicles . Removal of charcoal can be done remotely by dischaming into the cart on the floor below the filter floor.
The ventilation fans and interconnecting ducts are located on the floor below the filter floor, thus shielding the mechanical equipment from the filters.
The ductwork in general is run circumferential1y at the fan floor elevation with vertical drops into the areas to be served generally placed in the radial corridors.
A more detailed description of ventilation system I maintenance is provided in Section 9.4. U
- m. Where practicable, shielding is provided between radiation sources and areas to which personnel may have normal or routine access, and shielding is designed for maintaining doses as low as practicable.
The annulus building radiation zone drawings show that the above has been implemented (Fig. 12.1-1) .
Radiation shielding is provided on the basis of maximum concentration of radioactive materials within each shield region rather than the annual average values.
For batch processes as an example, the highest radionuclide concentration in the batching process is assumed (e . g . , just prior to draining of tank). No shielding credit is taken for equipment and components located outside the shielded regions and only very large components within the shielded regions are considered as object shields. Some of these components are also treated as sources, depending on their contents.
The bases for the selection of maximum radionuclide concentrations are described in Chapter 11. An activities balance for the entire plant is first established with detailed computer modeling which uses mass flow data as input. This activities balance is used in assessing maximum expected discharges from the 12.1-9 fn o/7 Amendment 17 bU i 'Ul 9/30/75
SWESSAR-P1 plant. Once this balance is established, the sources in individual components are intentionally unbalanced to simulate worst case conditions. The levels of activity under these conditions may thus be 10 times the expected levels in some components. This approach is carried out on a component-by-component basis so that shielding in each local area is adequate for the worst conditions, e.g., activity in a tank just prior to drainage. Dose rates ccuputed under such conditions ensure that the maximum allowable levels in each zone will not be exceeded under normal design conditions. Obviously, these data cannot be used to obtain maximum expected levels throughout the plant because the entire plant does not operate in tha t fashion. In fact, for a component i= a system to be at worst case conditions, other compo1. Its in that system must be operating at minimum source conditions rather than average levels.
- n. "ovable shielding and convenient means for its utilization are available for use where permanent shielding is needed but impractical.
If areas appear to lend themselves to temporary shielding, steel or concrete supports for this shielding are provided.
Convenient means for transport and placement are provided by the access corridor system in the annulus building which allows forklift trucks acct.ss on the circumferential corridor and hand carts and dollies access down the radial corridors.
- o. Adequate shielding is provided for radioactive wastes -
High level liquid waste storage is completely shielded in the annulus building or in the boron recovery tanks by the full height walls.
Solid waste is shielded both by the storage area walls in the solid waste building and the individual transportation shields.
Process gas charcoal bed absorbers and associated equipment are located in shielded cubicles.
- p. Remote handling equipment is provided wherever it is needed and practicable.
Remote handling equipment is provided for removing filters from the filter vessel and for placing into shipping containers.
bb9 2bb 12.1-10 Amendment 17 9/30/75
SWESSAR-P1 The solid waste system is essentiaAly a remotely
[,"-
operated system but with certain nonremote operations necessary. The placement of the container lid is required to be done manually using the jib crane. The solid waste shipping container is placed in a shipping shield prior to filling. The manual placement of the lid results in a minimal exposure to the operator.
Valve stations for radioactive service in general are arranged either in shielded cubicles away from the equipment served or are provided with reach rods or both. 'Ihe demineralizer and filter valves are in cubicles beneath the vessels and are also provided with reach rods.
- q. All design features for radiation control are de signed to accommodate maximcm expected (technical specification limit) failures such as fuel element cladding and steam generator failures.
Design features such as shielding and radiation zones accommodate clad defects in fuel rods producing 1 percent of rated core thermal power.
- r. Sampling sites will be located so exposures will be as low as practicable during such routine operations as sampling of f -gas , primary coolant, and liquid waste.
A sampling room is provided for the remote taking of samples for routine radioactive sample points in the reactor and auxiliary systems with the exception of the 12 percent boric acid samples from the waste evaporators. Sample points are provided with a ventilation hood, sample sink, splash screen, and valves located outside the splash screen. The samples are provided with a recirculation path behind the shield wall at the sample sink with reach rods for the operator. The sample sink is located at the 0 8--6 "
elevation in the annulus building and is near to the hot lab in the health physics building located at elevation 12'-0".
The thickness and extent of coverage of radiation shielding are determined in a manner to ensure that the maximum design radiation dose rates pre 3ented in Section 12.1.1 are not exceeded when all sources of radiation in the vicinity, including penetrations in the shields, are considered. The basic radiation transport analysis is based on Discrete Ordinates and Point Kernel computer code calculations which are described in Section 5.2 of RP-8 with conservative estimates of the sources involved. The methods of analysis employing the QAD-F, GGG-F, 12.1-11 r 9/G Amendment 17
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SWESSAR-11 and GAMTRAN Point Kernel Computer Codes are described in Sections 5.2.1, 5.2. 2, and 5.2.3 respectively of RP-8 (pages 5-8 -
through 5-16) . The application of the Discrete Ordinates programs is described in Sections 5.2.0 and 5.2.5 of RP-8 (pages 5-17 through 5-1C . Results obtained with the ANISN 'is crete Ordinates Code are shown in Figures 4.1-1 and 4.1-2 of RP-8. The nu jority of the dose rate levels reported in RP-8 are examples of use or the QAD-F Point Kernel Code (e.g. , Figure 4.1-14 and Table 4.1-18) . The computed radiation levels are used to determine the shield thickness and coverage requirements, and consideration is given to streaming through penetrations if any exist in the area.
The density of materials used in source and dose rate calculations is f ound in Table 12.1.2-1.
Information gained in operating stationc of a sinilar type is reviewed to improve the mathematical and physical models.
This information may be categorized into four general areas:
(1) equipment performance (e .g . , fuel pool cleaning system radiation levels), (2) contact dose rate measurements en operating components (e .g . , steam generators) , (3) radiation survey measurements on the shielded side of subcompartments and cubicles, and (4) total exposures associated with specific operation (e .g . , reactor head removal and reinstallation) . This information is used as a check against the predicted radiation levels and in some instances the operating data help to establish the source terms used in a given component analysis. Initial shielding requirements, to a large extent, are estimated based on prior plant designs.
In computing the dose ratet on which the confirmation of shielding thicknesses is based, a number of explicit and implicit conservative measures are incluCed. The conservative features of the design approach are disctssed below. Dose points are generally calculated along verti al shield surf aces opposite the most intense source in the vicini ty. These calculations are based on the inherent assumption thct plant personnel will spend the required time in each zone in contact with the shield at this point. This is a demonstrably conservative approach, since the dose rate actually decreases dramatically as the dose points are moved along th e surface of the shield due to the slant penetration involved. The additional fallof f of intensity with distance is also ignored by this approach.
Dose rates are generally calculated at 3 and 6 ft levels above walking surfaces, particularly if significant sources are located on the next level above the zone within the building.
The object shielding value of only the most massive components is considered in the computer models, both in the source region and in the zones in which radiation levels are calculated.
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SWESSAR-P1 Dose rates for post-shutdown conditions are computed at the earliest reasonable time after shutdown; subsequent decay is ignored for conservatism, i.e., the dose rate at that point in time is quoted despite the fact that the radiation levels continue to decay to lower values.
At a number of locations an explicit desian margin of a factor of 2 on dose rate is us ed where Nitrogen-16 (N-16) or reactor sources are the dominant contributor. As an example, a calculated dose rate of 2.5 mrem /hr is used to determine the shield thickness required for a <5 mrem /hr zone.
This margin is included due to the fact that the source strengths in the reactor systems are nominal values, not intentionally conservative ones (with the exception of the treatment of fission product levels which are assumed to be at the end of an operational period just prior to refueling) .
Transit times, in coolant loops etc, are computed as precisely as practicable with no inten tional conservatism. "Few region" models which are used to describe large components are intentionally designed so as not to overestimate comporent self-shielding. This provides an anount of conservatism which varies from component to component. In some instances no component self-shielding is included.
The modeling reflects the designer's knowledge of specific components and limitations which exist when component specitications have not been processed and details on individual parts are lacking.
The models allow for this uncertainty in a conservative manner, thus ensuring that the actual radiation leakage from the supplied component will be equal to or less (but not greater) than predicted values. This is accomplished in either of two ways:
(1) the component self-shielding may be ignored completely, or (2) the geometric model of an individual component is intentionally modeled with minimal self-shielding assumptions (e.g. minimum possible wall thickness). The actual radiation leakage is thus assured to be less than the calculated value due to + J.e treatment of self-shielding. The self-shielding factor frequently ranges from a factor of 2 to a decade or more for the actual supplied components.
12.1.2.1 Primary Shielding Primary shielding is provided to limit the following radiation emanating from the reactor vessel: neutrons diffusing from the core, prompt fission gammas, fission product gammas, and gammas resulting from the slowing down and capture of neutrons.
12.1-13 Amendment 17 q'; 9/30/75 66 L i
SWESSAR-P1 The primary shield is designed to:
- 1. Attenuate neutron flux to minimize activation of plant components and structures.
- 2. Reduce residual radiation from the core to a level that allows access to the region between the primary and secondary shields at a reasonable time after shutdown.
- 3. Reduce the contribution of radiation from the reactor to optimize thickness of primary and secondary shields.
We primary shield consists of a water-filled reactor vessel support shield tank and a 4-1/2 ft thick reinforced concrete reactor cavity wall. The shield tank has an annular thickness of approximately 3 f t and is located between the reactor vessel and the concrete reactor cavity wall. To maintain the integrity of the primary shield, streaming shields fabricated from appropriate materials are provided where necessary.
We reac-tor plant component cooling water system, described in Section 9.2.2, cools the water in the reactor vessel support shield tank. The normal operating water temperature in the reactor vessel support shield tank is approximately 120 F.
Because this temperature is less than the 150 F maximum specified in proposed ASME III, Division 2, no additional cooling for the adjacent concrete is required.
12.1.2.2 Secondary Shielding Activation and fission products from the reactor coolant system (RCS) is are the radioactive sources for which secondary shielding required. This shielding is provided by a variety of structures and equipment. The principal secondary shieldings include the containment structure wall and auxiliary equipment shielding.
Reactor Coolant Loop Shielding Nitrogen-16 is the major source of radioactivity in the reactor coolant during normal operation and, therefore, establishes the basic shielding requirements for the containment structure wall.
Activated corrosion and fission products in the RCS establish the shutdown radiaticn It'vels in the reactor coolant loop areas.
Table 11.1.1-2 lists ti fission product activities ir the RCS with 1 percent failed fuel. Table 11.1.1-2 also lists the activated corrosion product activities. Nitrogen-16 activity at th e reactor vessel outlet nozzle during operation at 4,100 MWt core output is given in Table 11.1.1-2.
669 T/2 O 12.1-14 Amendment 17 9/30/75
SWESSAR-P1 Containment Structure Shielding The containment structure shielding consists of the steel-lined, reinforced concrete cylinder and hemispherical dome described in Sections 3.8.1 and 6.2. This shielding attenuates radiation during all anticipated modes of reactor operation including the assumed design basis accident. Radiation levels are reduced to design levels or below at the outside surface of the containment structure and at the site boundary.
Interior shield walls separate reactor coolant loop, pressurizer, in-core instrumentation, and containment access sectors. This shielding allows access to the in-core instrument sector during normal operation and facilitates maintenance in all sectors during shutdown. Shield walls a e provided around each steam 12.1- 14 A , - , Amendment 24 h ()') Li J 4/23/76
SWESSAR-P1 generator above the operating floor to a height required for personnel protection.
Fuel Handling Shielding The fuel handling shielding, including both water and concrete, attenuates radiation from spent fuel assemblies, control rods, and reactor vessel internals to design levels and permits the removal and transfer of spent fuel and control rods to the fuel pool in the fuel building.
The refueling cavity above the reactor is formed by a stainless steel-lined, reinforced concrete st ructure . This refueling cavity becomes a pool when filled with borated water to provide shielding during the refueling operation.
The depth of the shielding water in the cavity is such that the radiation dose rate at the surface of the water from a fuel assembly does not exceed 2.5 mrem per hr during the short time intervals when the fuel handling operation brings the fuel assembly to its closest approach to the pool surface.
The cavity is large enough to provide storage space for the upper and lower internals and miscellaneous refueling tools.
The fuel pool in the annulus building is perTnanently filled with water to provide the minimum depth of water shielding specified in Table 9.1.2-1 above a fuel assembly when being withdrawn f rom the fuel transfer system. This depth is sufficient to reduce the dose rate at the surface of the water frun a raised fuel assembly to 52.5 mrem per hr. Water height above stored fuel assemblies is a minimum of 26 ft. These fuel pool walls are concrete 6 ft thick to ensure a dose rate less than 2.0 mrem per hr outside the building .
Auxiliary Equipment Shielding The auxiliary components may exhibit varying degrees of radioactive contamination due to the handling of various fluids.
The and maintenanceof the auxiliary shielding is to protect operating function personnel working near the various auxiliary system components, such as those in the chemical and volume control system, the boron recovery system, and gaseous waste systems, and the sampling system.
the radioactive liquid Restricted access to these areas of the annulus building is maintained during reactor operation.
Major components ci systems are individually shielded so that compartments may be entered without having to shut down possibly decontaminate the entire system. and Ilmenite be used in certain areas. Potentially highly contaminated ion concrete may exchangers and filters are located in separate shielded cells in 12.1-15 bb9
] [!' Amendment 24 4/23/76
SWESSAR-P1 the annulus building. Each ion exchanger or filter is enclosed '
in a separate, shielded compartment. The concrete thicknesses provided around the shielded compartments are sufficient to reduce the surrounding area dose rate to less than 5.0 mrem per hr and the dose ra te to any adjacent cubicle to less than 100 mrem per hr. Tha shielding thicknesses Oround the mixed bed demineralizers are based upon a saturation activity that gives a contact radiation level of approximately 12,000 Rem per hr.
In some areas tornado missile protection in the form of thick concrete af fords more shielding than that required for radiation protection.
Waste Storage Shielding The waste storage and processing facilities in the annulus building and the solid waste and decontamination building are shielded to provide protection for operating personnel in accordance with the radiation protection design bases in Section 12.1.1.
Boron recovery tanks, which are used to store " letdown" prior to its recycling to the plant or processing as waste, are shield ed to reduce dose rates to 2.0 mrem per hr in the area outside the boron recovery tank shield walls but within the reactor plant t ank area, Fig. 1.2-1. In the yard area outside the reactor plant tank area, the boron recovery tank shielding is sufficient to reduce the dose rate from the boron recovery tank to 5 0.2 mrem per hr. These radiation levels are reflected by the radiation zones shown in Fig. 12.1-1 (Sheet 3) .
Accident Shielding Accident shielding is provided by the containment structure, which is of reinforced concrete lined with steel. For structural reasons, the thicknesses of the cylindrical walls and dome are at least 54 and 30 in., respectively. These thicknesses are more than adequate to meet the guideline limits of 10CFR100 at the exclusion area boundary.
Control Room Shielding The design basis for the walls bounding the control room is that the design basis accident radiation whole body dose to personnel inside the control room be less than 5 Rem. This dose includes:
(1) the ex ternal radiation contribution from the postulated radioactive plume leaking from the containment structure, and (2) the 30-day radiation dose f rom activity inside the containment structure, assuming no cleanup.
12.1.3 Source Terms Primary coolant and secondary side activities are given in Section 11.1. Liquid and gaseous activities are given in Sections 11.2 and 11.3 respectively.
12.1-16 / 7" b O.O u' Amendment 7 2/28/75
SWESSAR-P1 12.1.4 Area Monitoring The area radiation monitoring system monitors the radiation levels in selected areas throughout the plant.
Each monitor has a radiation detector with a remotely operated check source. Each detector has:
A readout device in the control room eouipped with alarms to indicate loss of voltage, hich radiation level, and high-hich radiation level.
Local audible and visual alarm capability and a local readout device.
An independent power supply located in the control room.
Associated electronic equipment and cabling.
A permanent record of radiological events is maintained for each d et ector .
The monitors use Geiger-Mueller tubes or ionization chambers for radiation d et ection. Each detector is mounted in a fixed position in areas where personnel are expected to remain for 7 exterded periods of time.
G The alarm setpoint for each monitor is at a level determined by the Utility-Applicant to provide sufficient warning of hiah radiation levels to operating personnel. Alarm setpoints are based on maximum design dose rates for the different zones as outlined in Section 12.1.1 and the criteria provided by the Utility-Applicant in Section 12.1.7.
The operability of each detector is checked as required with the check source.
Each detector is connected to automatic annunciators in the control room to warn of loss of voltace , high, or hich-high radiation levels.
Power is provided to the area monitors by the 120 V a-c vital bus system as described in Section 8.3.
A list of area radiation monitors and their sensitivities and ranges is presented in Table 12.1.4-1. The sensitivity of each monitor is the lower range of the monitor.
Two high range containment area monitors are provided to monitor the high gamma dose rate possible in the containment structure after a LOCA. One monitor is permanently mounted on the wall behind the double doors opposite the annulus building personn el
@ hatch, approximately on the axial centerline of the hatch, n
L 78 [3 12.1-17 /0V
< '] Amendment 7 2/28/75
SWESSAR-P 1 Fig. 1.2-4, sheet 6. The other monitor is similarly situated opposite the fuel building personnel hatch, Fig. 1,2-8, sheet 2.
1 The specific location of other monitors within the areas listed in Table 12.1.4-1 is the responsibility of the Utility-Applicant and is provided in Section 12.1.7.
12.1.5 Operating Procedures E xternal radiation exposures to plant personnel during operation and maintenance are kept as low as practicable by plant design and through implementation of the radiation protection program described in Section 12.3 of the Utility-Applicant's SAR.
12.1.6 Estimates of Exposure The expected annual doses to onsite personnel are governed by the controls impose 6 by the plant supervision and health physics personnel. However, dose estimates for in plant personnel for routine operation are expected to parallel those reported from operating plant experience as discussed below.
Extensive radiation shielding is provided on the basis of the maximum concentration of radioactive materials within each shielded region rather than on annual average values. For batch processes, as an example, the point of highest radionuclide concentration in the batching process (e .g . , just prior to draining a tank) is assumed. The shielding and occupancy zones for normal operation are intentionally very conservative such that the normally received dose rates will likely be a fraction of the limits specified in 10CFR20.
The highest level of personnel exposure is anticipated to occur during shutdown and maintenance periods on systems containino items such as coolant purification filters, condensate, cleanup and radwaste demineralizers, ion exchange resins, charcoal adsorber units, and solid radwaste handling components. Since this is the case, the plant shielding and machinery locations provide maximum laydown space, maximum working room, and minimum time required to perform operations consistent with reasonable operation of the plant. Experience gained in the operation of nuclear plants is factored into these designs with the objective of minimizing the total exposure to plant personnel.
Stone & Webster actively participates in the Atomic Industrial Forum, Inc. activities in compiling operational data from operating nuclear power stations. Additionally, SSW obtains engineering feedback from several operating stations relative to measured radiation levels and how they relate to predictive analyses. Examples of this are extensive measurements of radiation levels and performance data of components at the Connecticut Yankee and Surry stations which have been reviewed in relation to the design of systems and components for SWESSAR-P1.
12.1-18 Amendment 7 2/28/75 5(19 L s
?'l L /
SWESSAR-P1 Plant visits by the responsible SSW radiation protection engineers with discussions with plant operators and plant health physics personnel form an integral part of the process by which the design evolution of SWESSAR-P1 design has benefited by operating experience relative to meeting the objective of minimizing the total exposure to plant personnel.
A conservative estimate of accumulated annual exposure is 500 man-Rem, which is not expected to be exceeded for the SGW 7 Reference Plants over their operating lifetime. This estimate is based on the operating experience at pressurized water reactor nuclear plants.
The data campiled by the Atomic Industrial Forum, Inc. National Environmental Studies Project-Study No. 5(*) served as the basis for estimating anticipated annual man-Rem exposures lor SWESSAR-P1. This experience has shown that the average exposure level is approximately 1 Rem / year per worker. Approximately 0.7 Rem / year of this value results from maintenance and refueling activities and 0.25 Rem / year from normal operations. This latter value is a factor of 20 below the 5 Fem / year limit in 10CFR20 on which the 100 mrem / Week desi n basis is established. It can be s ee n , therefore, that the design approach used on similar plants in which a 100 mrem / Week value is divided by the expected occupancy in hours / week results in operational levels which are a facter of 20 below 10CFR20 limits.
Since most of the anticipated exposures are expected to occur under circumstances which do not lend themselves to analytical prediction, such as maintenance on radioactive components, shielding design is based on worst case assumptions and design features are provided which will minimize exposures. Thus, actual exposures should be consistent with the as " low as practicable" requirement .
The large number of design features, which are the basis for assuming that the occupational radiation exposures will be as low as practicable for the Stone G Webster Reference Plant, should produce substantially lower exposures than those experien ced to date at operating stations. However, sufficient data are not available to predict what the long + era exposures will be over the operating lif etime of present plants. The added design f ea tures should reduce exposures sufficiently to allow for unforeseen increases without exceeding the 500 man-Rem estimate.
For example, cobalt 60 equilbrium conditions, considering its 5.3 year half-life, may become a constraining isotope in the long I term.
It is not evident that exposures measured to date are representative of the annual exposures which may accrue in presently operating plants in the long term. It is a reasonably conservative assumption that the design features included in this plant should provide assurance that long term exposures in excess q]OU LQ L 12.1-19 Amendment 7 2/28/75
SWESSAR-P1 oi 500 man-Rem will not occur. Quantitative estimates woulc be very imprecise based on the available operating data and would be Q
W subject to wide interpretation. Qualitative estimates, such as discussed herein, are within the range of good engineering judgment.
A conservative estimate of personnel radiation expo 3ure for normal plant operations has been proposed. The results of this estimate are summarized by job classification and by location in Table 12.1.6-1. A more detailed breakdown by individual tasks is presented in Table 12.1.6-2.
These estimates are conservative in that the radiation zone designation levels were used (i .e . , in terms of raRem/hr) in conjunction with man-hours spent in each zone. It is 0 anticipated that the actual levels will be a fraction of these design radiation zone values.
Only selected nurmal operations activities were considered.
Routine and special maintenance, instrument calibration, fuel handling, and inservice inspection were not included. The number of design features included in this plant design to reduce exposures in these categories, as discussed in Section 12.1.2, provide reasonable assurance that those exposures will be as low as reasonably achievable.
12.1.7 Interface Requirements f The Utilitv-Applicant shall provide criteria for area radiation monitor al arm setpoints and the location of the area monitors within tne ureus listed in Table 12.1.4-1.
Reference f or Section 12.1 (1) Atomic Industrial Forum, Inc., National Environmental Studies Project-Study No. 5, " Compilation and Analysis ot Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants" (In Publication) .
669 279 (l) 12.1-20 Amendment 16 8/29/75
SWESSAR-P1 TABLE 12.1-1 RADIATION ZONE DESIGNATIONS Maximum Dose Rate Zon. No. Zone . De sc ription (mrem /hr)
I Unrestricted area- < 0.20 continuous access II Unrestricted area- < 2.0 occupational access III Restricted area- <5 periodic access IV Restricted area- < 20 controlled access V Radiation area- < 100 controlled infrequent access VI High radiation area 2100 not normally accessible 1 of 1 us m
SWESSAR-P1 TABLE 12.1-2 TYPICAL LOCATIONS FOR DESIGNATED RADIATION ZONES Zone No. Typical-Locations I control room and all administrative areas II Turbine building III Annulus, solid waste and decontamination, and fuel building passageways in general IV Secondary passageways in annulus, solid waste and decontamination, and fuel buildings V In-core instrumentation room VI Inside shielded equipmenc compartments 4
50v 20 1 of 1
SWESSAR-P1 TABLE 12.1.2-1 MATERIALS USED FOR SOURCE AND DOSE RATE CALCULATIONS Material Density (lb/ft3)
Ilmenite concrete 240 Ordinary concrete 145 Steel 490.5 Lead 707.6 Air, steam, or vapor 0.075 Water
- a. Reactor coolant 46
- b. All other 62.4 Core
- 270
,ea U
1 of 1 bbh
- A homogenized density of the fuel, coolant, and structural material used in the reactor core.
SWESSAR-P1 TABLE 12.1-3 DESIGN BASIS DOSES, ACCIDENT CONDITIONS Location Maximm Dose Site Boundary 25 Rem in ? hr, whole body dose Control Room 5 Rem in 30 days, whole body dose 669 x.
1 of 1
SWESSAR-P1 TABLE 12.1.4-1 AREA RADIATION MONITOR LOCATIONS AND RANGES Sensitivity and Detector Location Range (mrem /hr)
Reactor containment area, 1-10* fH low range Reactor containment area, 1-10e lg high range (2)
Manipulator crane 1-10*
In-core instrumentation transf er 1-104 device area Decontamination area 0.1 -103 New fuel storage area 0.1-103 Fuel pool bridge 0.1-103 Annulus building control areas (2) 0.1-103 bample room 0.1-103 Solid waste drum storage 0.1-103 .
and handling m a .
Annulus building equipiaent door 0.1-103 Fuel bui?. ding equipment door 0. h10 3 Solid vaste building equipment door 0.1-103 Annulus building equipment door 0.1-103 Fuel building equipment door 0.1-103 Solid waste and decontamination building door 0.1-103 Spares
- Control room 0.01-102 Laboratory 0.01-102
- Spares will be Utility-Applicant dependent.
@ (. o U
oL nU['3 1 of 1 Amendment 17 9/30/75
SWESSAR-P1 TABLE 12.1.6-1 9
SUMMARY
OF ESTIMATED ANNUAL IN-PLANT EXPOSURE NORMAL OPERATIONS By Job Classification Dose (man-rem)
Auxiliary Operators 50 Chemistry Technicians 5.1 Health Physicists 16 Supervisors 2.7 TOTAL Approx 74 16 By Location Annulus Building 63 Reactor Plant Tank Area 3.5 x 10-2 Containment Structure 3.9 Solid Waste and Decontamination 5.5 Building Fuel Building 2.1 TOTAL Appro; 74 bOI [05 1 of 1 Amendment 16 8/29/75
SWESSAR-P1 TABLE 12.1.6-2 ESTIMATED ANNUAL IN-PLA!TP EXPOSURM, !KmMAL OPERATIONS Incation Operation Personne1(8) Frequency Man-I!ours( 2 3 Ma n-Rm Annulus Building Shif t pump boron recx)very test tanks A 1/ day 498.8 2.5 +00 Shitt pu:np waste test tanks A 2/ week 211.5 1.1 +00 Shitt pump low level waste drain tanks A 1/ week 71.1 3.6 -01 Shift high level waste drain tanks A 2/ week 3.5 1.8 -02 Shitt boron recovery tanks A 1/nnnth 0.4 2.0 --03 Surveillance tour (3) A 1/ shift 1,095 5.5 +00 Equipnmt cubicle check (3 ) A 1/ week 104 6.0 +00 Man watch station A 1/ shift 6,369 3.2 +01 16 Sampiing of:
heactor coolant (2) D 5/ week 55.5 2.8 -01 Pressurizer liquid B 5/ week 52.0 2.6 -01 Steam generator (4) B 5/ week 21.7 1.1 -01 Coolant puritication demineralizer inlet B 1/ day 121.7 6.1 -01 (bo1 ant purification demineralizer outlet B 1/ day 121.7 6.1 -01 O
g Doron thermal regenerative
,wg demaneralizer outlet B 1/ day 152.1 7.6 -01 liigh level waste drain tank B 2/ week 8.7 4.4 -02 rs ) low level waste drain tank B 1/ week 4.3 2.2 -02 CD s (bmponent cooling loop (3) B 1/ week 12.9 6.5 -02 Saf ety injection accumula-tor (3) B 1/ month 7.8 3.9 -02 1 of 3 Amendment 16 8/29/75
SWESSAR-P1 TABLE 12.1.6-2 (CO!4T) lescation Operation Personnel (*) Fre<Tuency Ma n-hotars( 2 ) Ma n -P m Doron thennal regenerative chill water B 3/ week 31.2 1.6 -01 boric acid storage tank (2) B 1/ week 6.9 1.4 -01 Eciueling water storage tank B 1/ week 7.8 3.9 -02 Contai e nt spray chemical dddition tanh B 1/ month 1.2 6.0 -03 Walkaays and passageways C 1/ day 821 4.1 +00 surv ry Operating areas survey C 1/ day 304 1.5 +00 Et"ipment cubicles survey C 1/ week 79.7 4.0 +00 Inspection tour (3) D 1/ week 52. 2.6 -01 16 log review E 1/shif t 109 5.5 -01 Int:pection tour ( 3 ) E 1/ day 365 1.8 +00 SUB10TAL 6.3 +01 Reactor Survey C 1/ month 17.3 3.5 -02 Plant Tank Area Containment Surveillance A 1/ week 13.9 1.4 +00 Surveillance C 1/ week 13.9 1.4 +00 Boron injection tank sample B 1/ week 11.3 1.1 +00 SUB1VTAL 3.9 +00 Os Solid Waste and Fill, solidity solid waste 7 Decontamination containers A 1/ week 534 1.1 +00 q; Building Fill, cornpact solid waste A 1/ week 85 4.3 -01 drums DJ CD Sampling of:
..J Primary grade wat'r te k B 1/ day 97.3 4.9 -01 Doron recovery tes tank B 1/ day 36.5 1.8 -01 1
2 of 3 Amenctment 16 8/29/75
SWESSAR-P1 I
TABLE 12.1.6-2 (00!rr)
_Lt>ca t i on Ope *ra t ion Personnel (1) LTtNiuency Esn-hours (2) Kin -Pera Waste test tank B 2/ week 10.4 5.2 -02 Doron recovery tank B 1/ month 1.2 6.0 -03 Walkways and passageways survey C 1/ day 285.9 1.4 +00 Rooms and work areas survey C 1/ week 28.6 1.6 +00 Solid waste disposal C 1/ week 121 2.9 -01 SUBTOTAL 5.5 +00 Fuel Building Sampling of:
Fuel pool demineralizer inlet B 1/ week B.9 4.5 -02 Fuel pool demineralizer 16 outlet B 1/ week 5.2 2.6 -02 Walkways and work areas survey C 1/ week S7.2 2.0 +00 Shippirs] and receiving area survey C 1/ week 13. 6.5 -02 SUBTOTAL 2.1 +00 NOTES:
- 1. A - Auxiliary Operator; B - Chemistry Technician: C - llealth Physicist; D -- Operations Supervisor; E - Shitt Supervisor
- 2. Includes transit times where applicable. Total transit time for a group oi operations such as annulus building samp)ing is included in one of the operations.
- 3. Includes time spent in other buildings, but bulk of time is in annulus building.
& 4. Man-Rem resulting f rom operations in areu e <ith designated radiation zones 5 0.2 m Rem /hr are considered cn negligible.
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DESIGN BASIS RAClATION ZONES FOR SHIELOING PWR STANDARD PLANT SAFETY ANALYSIS REPORT SWESSAR-Fi-3.
AMENDMENT I 7/ 30/ 74
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-- K 2 ZONES FOR SHIELDING PWR STANDARD PLANT SAFETY ANALYSIS REPORT SWESSAR P-1 m
W AMENDMENT I 7/ 3 0 / 74 0 a n c,
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SWESSAF-P1 12.2 VENTILATION 12.2.1 Design Obiectives The function and design bases of the various ventilation systems are given in their respective subsections of Section 9.4.
Consistent with these, the following specific objectives pertain to radiation protection and the commitment to the goal and the principle that occupational radiation exposures will be as low as practicable.
- 1. The airborne radioactivity inside plant buildings, other than the containment structure, during normal operation and under anticipated operational occurrences, is less than the concentrations given in Column 1, Table I of Appendix B to 10CFR20.
- 2. Concentrations in areas accessible to station adminis-trative personnel and incite visitors ' areas are less than the concentrations given in Column 1, Table II of Appendix B to 10CFR20.
- 3. The airborne concentrations in all plant areas shall be as low as practicable.
H
- 4. The containment atmosphere filtration system shall be capable of reducing the airborne iodine concentration in the containment atmosphere to 1 MPC (I-131) after dpproximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of filter operation under the conditions of expected reactor coolant concentration and leakage.
- 5. The con tainment purge air system shall be capable of reducing airborne radiation levels in the containment to acceptable levels during extended personnel occupancy of the containment.
- 6. The fuel building ventilation system shall operate in the once-through mode in the event airborne radiation levels exceed a predetermined level.
- 7. Air flow within the annulus, solid waste and decon-tamination, fuel, and turbine buildings shall generally be from areas less likely to have airborne contamination to areas more likely to have airborne contamination.
The expected airborne radioactivity levels during normal opera-tion and under anticipated operational occurrences for the va rious plant buildings are given in Sect ion 12.2.3. The expected annual inhalation dose rates to plant personnel are given f or each building in Section 12.2.6. l hbh b 12.2-1 Amendment 17 9/30/75
SWESSAR-P1 12.2.2 Design Description Detailed descrip tions of the ventilation systems for the plant building are given in the following sections where applicable:
Number Title 6.2.3.1 Supplementary Leak Collection and Release System 9.4.1 Control Building Ventilation System 9.4.2 Annulus Btulding Ventilation System 9.4.3 Solid Waste and Decontamination Building Venti-lation System 9.4.4 Turbine Building Ventilation System 9.4.5 Containment Structure Ventilation Systems 9.4.6 Fuel Building Ventilation System 12.2.3 Source Tenns Significant radioisotopic leakage rates into the containment structure , turbine building, annulus building, and fuel building are listed in Tables 12.2.3-2 through 12.2.3-5 respectively based on assumptions listed in Table 12.2.3-1. The assumptions in Table 12.2.3-1 are, where applicablc, taken from Regulatory Guide 1.42. Airborne radioactivity at all other locations is negligible.
The expected airborne activity levels inside the containment structure , turbine building, annulus building, nnd fuel building are listed in Table 12.2.3-6. Reactor coolant and steam generator secondary side activities are given in Section 11.1.
The models used in calculating the airborne concentration in the various buildings are given below.
The results given in Tables 12.2.3-2 through 12.2.3-6, 12.2.6-1, and Fig. 12.2-1 for B&W are generally taken from previous unulyses which were based on values of parameters slightly 19 different than those given in Table 12.2.3-1. Since the results are consistent with those already presented f or the W and C-E NSSS, revisions in the calculations are unnecessary for this application f or Preliminary Design Approval.
A. Containment Structure The containment s tructure is not normally accessible to personnel except with health physics supervision. Two 15,000 cfm recirculating charcoal filters ensure th- airborne iodine in the 12.2-2 Amendment 19 7 ^. ; 12/12/75 J[b<a/ sci
SWESSAR-P1 containment is as low as practicable for work in that area of the plant (Section 9.4.5.3) .
Fig. 12.2-1 shows iodine-131 concentration in the containment building af ter one or two charcoal filters are operating.
After shutdown, the containment purge air system will further reduce the airborne activity within the containment structure.
12.2-2A Amendment 19 12/12/75 G0N 3LL
SWESSAR-P1 The containment atmosphere tritium concentratir n assumes the same relative tritium concentration in the containr.ent atmosphere as exists in the reactor coolant leakage.
B. Turbine Building The tritium concentration in the turbine bailding assumes that all the steam leakage into the turbine building evaporates.
C. Annulus Building The tritium concentration in the annulus building atmosphere assumes all the primary coolant leakage into the annulus building evaporates when in f act all the leakage assumed is collected in sumps and drains and is not available for evaporation.
D. Puel Building The tritium concentration in the fuel building atmosphere assumes that the spent fuel pool has the same relative tritium concentration as the reactor coolant.
12.2.4 Airbone Radiation Monitoring Fixed airborne radiation monitoring instruments are located throughout the plant. The location and number of these instruments ensure a continuous flow of information to the plant operating personnel concerning the airborne radioactivity levels throughout the plant and in ventilation effluent' streams. The instruments are designed for service, based on expected radioactivity levels during normal operation and anticipated occurrences.
Characteristics of the airborne radiation monitors are similar to those of the process and effluent nonitors described in Section 11.4.2.
All the monitors discussed in this section with the exception of the containment purge air exhaust and tl. > ventilation vent high 7
range monitors are a$.rborne radiation mGaitors of the standard type described in Section 11.4. '.3.1 and shown in Fig. 11.4-1 or 12.2-2.
Two detector (gas and particulate) airborne radiation monitoring is provided for the following locations:
Containment atmosphere Annulus building (first level)
Annulus building (second level)
Annulus building (third and fourth levels) 6 () O j23 12.2-3 Amendment 17 9/30/75
SWESSAR-P1 Solid waste and decontamination building h
Puel building Ventilation vent effluent Ventilation systems multisampler Control room air intakes Sensitivities and ranges for these monitors are given in Table 12.2.4-1.
Two single detectors are provided for the containment purge air exhaust line.
U A single detector is provided for the ventilation vent high range monitor.
The airborne radiation monitoring system is supplemented with portable fixed filter paper sampling systems to check the in stalled system and for major maintenance, This equipment provides airborne radiation monitoring where there is a potential for high airborne radioactivity.
12.2.4.1 Containment Atmosphere Radiation Monitoring System The containment atmosphere monitor is shown in Fig. 12.2-2. The containment atmosphere monitor draws a sample from the con-tainment atmosphere recirculation system or the containment filtration system p.'ction 9.4.5) and monitors the radioactivity concentration leveb. in the containment structure.
The containment atmosphere radiation monitoring system is in operation continuously during normal plant operation . In the event of a los s-of-coolant accident, a containment isolation phase A (CIA) signal closes the containment isolatior, valves in the containment atmosphere monitoring lines.
These motor operated and require testing as specified in Sections valves are 11.4.2 and 16.4.4.
A sample of the containment atmosphere for laboratory analyser may also be taken using the containment atmosphere radic cion monitoring system.
12.2.4.2 Annulus Buildino Monitors For airborne radiation monitoring, the annulus building venti-lation is divided into three separate zones: the first level, the second level, and the upper two levels. For each monitor, the sample point is upstream of the point where the three h
669 3 c:v.
12.2-4 Amendment 17 9/30/75
SWESSAR-P1 Various areas in the annulus building (5)
Supplementary leak collection and release system i
n 12.2.4.8 Containment Purge Air Monitors A description of the containment purge air monitors is given in Section 11.4.2.3.1.
12.2.4.9 Control Room Air Intake Monitors U Each control room air intake is continuously monitored for airborne radioactivity by means of airborne radiation monitors similar to those described in Section 11.4.2.3.1. The normal control room air intake has a single monitor (two detectors) .
Each remote con trol room air intake has two monitors (each monitor with two detectors) . The location of these monitors is shown in Fig. 9.4.1.1. During normal operation the normal control room air intake monitor warns of increasing airborne radiation levels. During a LOCA , the remote control room air intake airborne radiation monitors indicate airborne radiation levels at the intakes and enable selection of the uncontaminated intake.
12.2.5 Operating Procedures This section is within the Utility-Applicant's scope and SAR.
12.2.6 Estimates of Inhalation Doses The expected do se rates in the containment, annulus, fuel, and turbint buildings are given in Table 12.2.6-1 based on parameters given in Section 12.2. 3, Table 12.2.3-1 and the concentrations in Tables 12.2.3-6, exclusive of tritium. Airborne dose rates in all other locations are negligible.
Whole body dose rates in Table 12.2.6-1 are calculated according to H D wg = Dy + 1/6 D Thyroid dose rates in Table 12.2.6-1 are calculated according to D.
I
=[ i (B.R.) (A- ) (C I) (CTHY).
3 where D = whole body dose rate, mrem /hr WB Dy = }[ 0.25 E y A i Ci for the gamma dose rate in air in i a semi-infinite cloud ( 2 ) . This form is used with isotopes having 7 mean f ree paths in a rr/ q0
@(
32 12.2-6 W it2W Amendment 17 9/30/75
SWESSAR-P1 ventilation exhaust paths mix prior to discharge via the ventilation vent.
12.2.4.3 Solid Waste and Decontamination Building Monitor The sample point of this monitor is downstream of the last point of discharge to the ventilation vent. Upon receiving a high activity alarm in the control room, the operator can manually divert the discharge flow through the solid waste and decon-tamination building ventilation system filter bank.
12.2.4.4 Puel Building Monitors The sample poin t of these monitors is downstream of the last point of discharge to the ventilation vent. Upon a high activity alarm, the discharge flow is automatically directed through the supplementary leak collection and release system (SLCRS) (Section
- 6. 2. 3.1) . These monitors are designated Safety Class 3 and Seismic Category I.
12.2.4.5 Ventilation Vent Monitor A description of the ventilation vent monitor is given in Section 11.4.2.3.1.
12.2.4.6 Ventilation Vent High Range Mcaitor A description of the ventilation vent hiar. range monitor is given in Section 11.4.2.3.1. U 12.2.4.7 Ventilation Systems Multisampler This system consists of a single airborne radiation monitor used to monitor each of twelve remote locations (listed below) on a rotating basis. A manifold and selector valve arrangement driven by an automatic sequencer gives each of the twelve areas access to the airborne radiation monitor on a periodic basis rangina from 30 min to 24 hr as chosen by the operator. A manual override of the automatic sequence permits the operator to select and lock in on any one sample point.
The sources sampled are:
Control room ventilation discharge duct Decontamination area in solid waste and decontamination building Turbine building ventilation exhaust Solid waste and drumming area 3 24 g
Sampling sinks exhaust hoods in annulus building (2) 12.2-5 Amendment 17 9/30/75
SWESSAR-P1 D
={ (R) (A- ) (C2) for the gamma dose rate in air in a i semi-hinite cloud. This form is used with isotopes having 7 mean free paths in air >70 feet. It is derived by considering the dose due to a dif-ferential shell element and integrating over the entire volume.
Dg ={ 0.23 h A i C i for the beta surface body dose rate i in an infinite cloud (13 The factor of 1/6 in the calculation of the beta dose rate contribution is based on the E to 1 ratio for the allowable skin doae to the allowable whole body dose for occupa-tional exposure, given in 10CFR20.101(2 ).
Ey = average ganma energy per disintegration (Mev/ dis)
E = cverage beta energy per disintegritrion (Mev/ dis)
A.
' = building airborne concentration, Table 11.2.3-6, of isotope A; , (uci/cc)
C; = 3,600 sec/hr x 103 mrem / rem 17 R = radius of hemisphere with volume equal to building volume D
T = thyroid dose rate (mrem /hr)
C ,= thyroid dose conversion factor from TID 14844(33 B.R. = breathing rate, m3/sec C = 1/2 (3600 see) (10 mrem)(p4 cc ergs
)
7'l j:1 7
M Ua I = absorption coef ficient at gamma energy E (cm )
y
- = density of air at F's th th E = average gamma energy for the i isotope in the j T '. .l energy group, Mev/ dis. A total of seven energy groups are used for the isotopes of interest Occupancy times for the various buildings are given in Table 12.1.6-2.
g 327 669 8 12.2-7 Amendment 17 9/30/75
SWESSAR-P1 i
References for Section 12.2.6
- 1. U.S.A.E.C., " Meteorology and Atomic Energy 1968," TID-24190, July 1968.
11
- 2. 10CFR20 - Standards for Protection Against Radiation.
- 3. J .J . DiNunno, et al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, March 1962.
O bbh b 12.2-8 Amendment 17 9/30/75
o M AR-P1 TABLE 12.2.3-1 ASSUMPTIONS USED IN THE CALCULATION OF AIRBORNE CONCENTRATIONS IN VARIOUS BUILDINGS Containrent Turbine Annulus Fuel Building Building Building Building
- 1. Reactor Coolant Equilibrium Concentratials (Expected) Table 11.1.2-2 -
Table 11.1.2-2 -
- 2. Secondary Side Equilibrium Concentrations (Expected) -
Table 11.1.3-2 - -
- 3. Core Thermal Power (Mwt) - - -
Table 11.1.1-1 weg fu q B&W 3,876 Il v-/ C-E 3,876
([.. ) W41 W-3S 3,876 v.
p . rg"' 3,636
- 4. Irak Rate into Building
[ ) Equivalent Hot Reactor (bolant (Ib/ day) 240 160 r . , eg - -
g;] Equivalent Main Steam Leakage p :-3 (1b/hr) -
1,700 - -
L .~ ; n s v 5. Normal Moisture in Atmosphere (lb) 2,450 - - -
, ..v
', ? 6. Partition Factor (PF) Far e* Noble Gases 1.0 1.0 1.0 -
Halogens 0.1 1.0 .005 ( (W-t 1,%-3S) -
.001 (B&W,C-E)
- 7. Mixing In Building At2nosphere, 1 70 100 100 100
- 8. Building Ventilation Rate (cfm) -
7.0 x 105 6.0 x 10* 1.5 x 10*
sc 9. Building Free Volume (ft3) 3.4 x 10 * (B5W,C-F,W-41) 8.3 x 106 3.3 x 106 9.3 x 10*
3.1 x 106 (W -3S)
L~;
rs s s
1 of 2 Amendment 19 12/12/75
O SWESSAR-P1 TABLE 12.2.3-1 (CONT)
Containment Turbine Annulus Fuel Building Building Isuild ing Building Recirculation Filters Yes No No No Filter Efficiency 90% - - -
Amount of Core in Spent Fuel Pool - -
1 Decay Time of Spent Fuel (hr)
BLW -
C-E -
92 l13 72 W-41 - -
W-3S 20 100 Escape Rate Coefficients of Spent Fuel Noble Gases (sec-1) - - -
6.5 x 10-3a Haloger's (s ec-8 ) - - -
1.3 x 10-aa Fuel Pool Decontamination Factors Noble Cases - -
1 Halogens - - -
100 Fuel Pool Evaporation kate (lb/hr) - - -
12
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2 of 2 Amendment 19 12/12/75
SWESSAR-P1 TABLE 12.2.3-2 ANIRJAL AVERAGE RADIOACTIVITY LEAKAGE INTO CONTAINMENT STRUCTURE ATMDSPHERE Expected Leakage Rate, uCi/sec Isotope B&W {-E W-41 W-3S H3 1.3 E 00 1.3 E 00 1.3 E 00 1.3 E 00 I-131 5.5 E-02 3.5 E-02 2.9 E-02 3.6 E-02 I-132 1.5 E-02 1.4 E-02 1.3 E-02 1. 4 E -0 2 1-133 6.7 E-02 5.0 E-02 4.4 E-02 5.2 E-02 I-134 6.6 E-03 6.5 E-03 6.4 E-03 6.5 E-03 I-135 3.0 E-02 2.6 E-02 2.4 E-02 2.6 E-02 Kr-83m 2.6 E-02 2.4 E-02 2.4 E-02 2.5 E-02 Kr-85m 1.2 E-01 1.0 E-01 9.9 E-02 1.1 E-01 19 Kr-8 5 4.2 E-03 2.6 E-03 2.2 E-03 2.7 E-03 Kr-87 7.6 E-02 7.3 E-02 7.2 E-02 7.4 E-02 Kr-88 2.3 E-01 2.1 E-01 2.1 E-01 2.2 E-01 Kr-89 6.9 E-03 6.9 E-03 6. 9 E -0 3 6.9 E-03 Xr-131m 1.1 E-02 6.7 E-03 5.6 E-03 7.0 E-03 Xe-133m 7.1 E-02 4.8 E-02 4.2 E-02 5.0 E-02 Xe-133 3.1 E 00 2.0 E 00 1.7 E 00 2.1 E 00 Xe-13Em 1.8 E-02 1.8 E-02 1.8 E-02 1. 8 E -0 2 Xe-135 3.1 E-01 2.5 E-01 2.3 E-01 2.6 E-01 Xe-137 1.2 E-02 1.2 E-02 1.2 E-02 1.2 E-02 Xe-138 5.9 E-02 5.9 E-02 6.0 E-0 2 6.0 E-02 Amendment 19 1 of 1 12/12/75 l
SWESSAR-P1 TABLE 12.2.3-3 ANNUAL AVERAGE RADIOACTIVE LEAKAGE Ilff0 TURBINE BUILDING A'INOSPHERE Expected Leakage Rate, UCi/sec Isotope BSW C-E W-41 W-3S H-3 2.1 E-01 2.1 E-01 2.1 E-01 2.1 E-01 I-131 5.9 E-05 1.7 E-05 1.9 E-05 2.6 E-05 I-132 1.6 E-05 1.2 E-05 1.6 E-05 1.9 E-05 I-133 7.4 E-05 2.2 E-05 2.6 E-05 3. 3 E-0 5 1-134 3.7 E-06 1.3 E-07 1.1 E-01 1.4 E-07 I-135 3.3 E-05 1.0 E-05 1. 2 E -0 5 1.4 L-05 Kr-83m 1.2 E-0 6 1.1 E-06 1.1 E-06 1.3 E-06 Kr-85m 5.7 E-06 4.8 E-06 4.6 E-06 5.5 E-0 6 Kr-85 2.0 E-07 1.2 E-07 1.0 E-07 1.4 E-07 19 Kr-87 3.5 E-06 3.3 E-06 3.3 E-0 6 3. 8 E-0 6 Kr-88 1.1 E-05 9.7 E-06 9.5 E-06 1.1 E-05 Kr-89 3.2 E-07 3.1 E-07 3.1 E-07 3.5 E-07 Xe-131m 5.0 E-07 3.0 E-07 2.6 E-07 3. 6 E-07 Xe-133m 3.4 E-06 2.2 E-06 1.9 E-0 6 2.6 E-06 Xe-133 1.5 E-04 9.1 E-05 7.9 E-05 1.1 E-04 Xe-135m 9.4 E-05 7.0 E-0 5 7.1 E-0 5 8.0 E-05 Xe-135 1.4 E-05 1.4 E-05 1.3 E-05 1. 6 E-0 5 Xe-137 5.9 E-07 5.7 E-07 5.8 E-07 6.5 E-07 Xe-138 2.7 E-06 2.6 E-06 2.6 E-06 3.0 E-06 hh3 )
1 of 1 Amendment 19 12/12/75
SWESSAR-P1 TABLE 12.2.3-4 ANNUAL AVERAGE RADIOACTIVITY LEAKAGE RATE Iff1'O ANNULUS BUILDING ATMOSPHERE Expected Leakaqe Rate, uCL/nec Isotope B&W C-E W-41 W-3S H-3 8.4 E 01 8.4 E-01 8.4 E-01 8.4 E-01 I-131 3.7 E-04 2.3 E-04 9.8 E-04 1.2 E-03 I-132 9.8 E-05 9.1 E-05 4.4 E-04 4.6 E-04 I-133 4.5 E-04 3.3 E-04 1.5 E-03 1.7 E-03 I-134 4.4 E-05 4.3 E-05 2.1 E-0 4 2.2 E-04 I-135 2.0 E-04 1.7 E-04 7.9 E-04 8.7 E-04 Kr-83m 1.7 E-02 1.6 E-02 1. 6 E-0 2 1.6 E-02 Kr-85m 7.9 E-02 7.0 E-02 6.6 E-02 7.1 E-02 jg Kr-85 2.8 E-03 1.7 E-03 1.5 E-03 1.8 E-03 Kr-87 5.1 E-02 4.9 E-02 4.8 E-02 4.9 E-02 Kr-88 1.6 E -01 1.4 E-01 1.4 E-01 1.5 E-01 Kr-89 4.6 E-03 4.6 E-03 4.6 E-03 4.6 E-03 Xe-131m 7.0 E-03 4.4 E-03 3.8 E-03 4.7 E-03 Xe-133m 4.8 E-02 3.2 E-02 2.8 E-02 3.4 E-02 Xe-133 2.1 E 00 1.3 E 00 1.1 E 00 1.4 E 00 Xc-135m 1.2 E-02 1.2 E-02 1.2 E-02 1.2 E-02 Xe-135 2.0 E-01 1.7 E-01 1.5 E-01 1.7 E-01 Xe-137 8.2 E-03 8 . 3 E-0 3 8.3 E-03 8.3 E-03 Xe-138 4.0 E-02 4.0 E-02 4.0 E-02 4.0 E-02 b;O NJb 1 of 1 Amendment 19 12/12/75
SWESSAR-P1 TABLE 12.2.3-5 ANNUAL AVERAGE RADIOACTIVITY LEAKAGE RATE INTO FUEL BUILDING ATMOSPHERE Expected Leakage Rate, uCi/sec Isotope IEW C-E W-41 W-3S H-3 1.5 E 00 1.5 E 00 1.5 E 00 1.5 E 00 I-131 1.1 E-04 1.2 E-04 1.4 E-04 1.0 E-04 I-132 9.5 E-05 1. 2 E-0 4 1.9 E-04 8.6 E-05 I-133 1.6 E-05 3.2 E-05 1.8 E-04 1.2 E-05 I-134 - -
2.4 E-10 -
I-135 2.2 E-08 1.8 E-07 3.9 E-05 9.2 E-09 Kr-83m -
5.6 E-11 1.6 E-04 -
Kr-85m 1.6 E-08 3.9 E-07 1.4 E-03 -
g Kr-85 8.0 E-04 8.3 E-04 8.3 E-04 7.8 E-04 Kr-87 - -
1.1 E-06 -
Kr-88 -
1.6 E-09 6.3 E-04 -
Xe-131m 1.3 E-04 1.2 E-04 8.4 E-0 5 1.3 E-04 Xe-133m 2.0 E-03 2.6 E-03 3.9 E-03 1.8 E-03 Xe-133 1.2 E-01 1.3 E-01 1.6 E-01 1.1 E-01 Xe-135m 1.7 E-06 1.4 E-05 3.0 E-03 7.1 E-07 Xe-135 3. 8 E-0 4 1.7 E-03 4.5 E-02 2.1 E-0B Note: " " indicates concentration <1.0 E-11 b
1 of 1 Amendment 19 12/12/75
SWESSAR-P1 1
TABLE 12.2.3-6 EXPECTED RADIOACTIVE AIRBORNE CONCENTRATION INSIDE MAJOR PLANT BUILDINGS Radioactive Airborne Concentration, uCi/cc Containment Turbine Annulus Fuel Isotope Structuren Building gilding Building H-3 1.2 E-05 6.5 E-10 3.0 E-08 2.1 E-07 I-131 4.6 E-09 5.9 E-14 3.5 E-11 2.0 E-11 1-132 7,8 E-10 4.4 E-14 1.2 E-11 2.0 E-11 I-133 4.7 E-09 7.7 E-14 5.1 E-11 2.5 E-11 I-134 2.1 E-10 3.0 E-16 4.4 E-12 1.9 E-17 1-135 2.0 E-09 3.4 E-14 2.6 E-11 5.0 E-s2 Kr-83m 2.5 E-09 3.3 E-15 4.2 E-10 1.6 E-11 Kr-85m 2.3 E-08 1.4 E-14 2.0 E-09 1.7 E-10 Kr-85 1.8 E-07 3.0 E-16 5.1 E-11 1.2 E-10 ;y Kr-87 4.9 E-09 8.9 E-15 1.1 E-09 9.5 E-14 Kr-88 3.1 E-08 2.7 E-14 4.2 E-09 7.1 E-11 Kr-89 2.0 E-11 2.6 E-16 1.3 E-11 -
Xe-131m 8.9 E-08 7.9 E-16 1.3 E-10 1.? E-11 Xe-133m 1.2 E-07 5.8 E-15 9.7 E-10 5.8 E-10 Xe-133 1.2 E-05 2.4 E-13 4. 0 E -0 8 2.4 E-08 Xe-135m 1.5 E-09 1.4 E-13 1.2 E-10 1.1 E-09 Xe-135 1.2 E-07 4.1 E-14 5.0 E-09 6.5 E-09 Xe-137 4.3 E-11 5.6 E-16 2.7 E-11 -
Xe-138 7.6 E-10 5.1 E-15 3.8 E-10 -
Note: " " indicates concentration <1.0 E-18 1 Concentrations after 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> recirculation through charcoal filters at 30,000 cfm.
W 1 of 1 Amendment 17 9/30/75 6 d 3 3 ')
SbT !'S AP -P 1 TABLE 12.2.3-6 EXPECTED FADIDACTIVE AIEPOhNE CONCENTiIsTIO?-
INSIDE MAJOR PLANT BUILDII:GS bud i oa cti vo Airborno Concentr a t io:: , uCi/cc Cont a inm en t Turbine Annulus Furl Isotope Structure! Euildin:7 Puildino builcina H-3 9.1 E-06 6.5 E-10 3.0 E-08 2.1 E-0 7 I-131 S.1 E-09 7.8 E-14 4.3 E-11 1.4 E-11 1-132 8.6 E-10 S.4 E-14 1.3 E-11 9.2 E-12 I-133 S.4 E-09 9.9 E-14 S.9 E-11 1.6 E-12 I-134 2. 3 E-10 3.7 E-16 4.4 E-12 -
I-135 2.3 E-09 4.2 E-14 2.8 E-11 1.2 E-15 Kr-83m 2.9 E-09 3.6 E-15 4.3 E-10 -
Kr-85m 2.8 E-06 1.6 E-14 2.2 E-09 6.2 E-16 17 Kr-85 2.5 E-07 4.3 E-16 6.4 E-11 1.1 E-10 Kr-87 5. 6 E -09 1.0 E-14 1.2 E-09 -
Kr-88 3.6 E-08 3.2 E-14 4.2 E-09 -
FJ-89 2.2 E-11 3.0 E-16 1.3 E-11 -
Xe-131m 1. 2 E-07 1.1 E-15 1.6 E-10 2.1 E-11 Xe-133m 1.7 E-07 7.9 E-15 1.2 E-09 2.5 E-10 Xe-133 1.6 E-05 3.3 E-13 4.9 E-08 1.6 E-08 Xe-135m 1.8 E-09 1.6 E-13 1.2 E-10 2.9 E-13 Xe-135 1.5 E-07 S.0 E-14 S.6 E-09 2.8 E-11 Xe-137 4.8 E-11 6.3 E-16 2.7 E-11 -
Xe-138 8.5 E-10 S.7 E-15 3.8 E-10 -
Note: "" indicates concentration <1.0 E-18 1Concentratione ali.er 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> recirculation through charcoal filters at 30,000 cfm.
W-3S 1 of 1 Amendinent 17 9/30/75
/, t. y ' i7, b
SKESSAR-P1 TABLE 12.2.3-6 MAXIMUM EXPECTED "ADIOACTIVE AIREORNE CONCENTRATIO:.
INSIDE MAJOR PLANT BUILDINGS Radioactive Airborne &>ncentration, uCi/cc Containment Turbine Annulua Fuel Isotope Structure 1 Buildino Buildina buildirg H-3 1.2 E-05 6.5 E-10 3.0 E-08 2.1 E-07 I-131 8,6 E-09 6.3 E-13 1.3 E-11 1.4 E-11 I-132 8.6 E-10 1.5 E-13 2.7 E-1 9.5 E-12 1-133 7.0 E-09 7.8 E-13 1.5 E-11 1.9 E-12 I-134 2.2 E-10 2.9 E-14 8.8 E-13 -
I-135 2.6 E-09 3.4 E-13 6.4 E-12 2.5 E-15 Kr-83m 2.7 E-09 1.1 E-14 4.4 E- 10 -
Kr-85m 2.8 E-08 5.6 E-14 2.4 E-09 1.8 E-15 Kr-85 3.4 E-07 2.1 E-15 9.8 E-11 9.9 E-11 Kr-87 5.2 E-09 3.0 E-14 1.2 E-09 -
Kr-88 3.5 E-08 1.1 E-13 4.4 E-09 -
Kr-89 2.0 E-11 5.1 E-16 1.2 E-11 -
Xe-131m 1.7 E-07 5.3 E-15 2.5 E-10 1.9 E-11 Xe-133m 2.1 E-07 3. 6 E--14 1.7 E-09 2.5 E-10 Xe-133 2.2 E-05 1.6 E-12 7.3 E-08 1.5 E-08 Xe-135m 1.8 E-09 4.9 E-13 1.2 E-10 7.2 m.-13 Xe-135 1.6 E-07 2.2 E-13 6.7 E-09 4.5 E-11 Xe-137 4.3 E-11 1.1 E-15 2.5 E-11 -
Xe-138 7.6 E-10 1.3 E-14 3.7 E-10 -
- Concentrations after 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> recirculation through charcoal _ , _ ,
filters at 30,000 cim. rn O s) O1 j '; /
ESh 1 of 1 Amendment 28 8/6/76
SWESSAR-P1 TABLE 12.2.3-6 EXPECTED PADIOACTIVE AIRLORNE CONCEN'IkR1 ION INSIDE MAJOR PIAITI' BUILDINGS Radioactive Airborne Concentration, UCi/cc Annulus Fuel Containment Turbine Buildina Building Buildino Isotope Structure 1 1.2 E-05 6.5 E-10 3.0 E-08 2.1 E-07 H-3 5.1 E-14 8.1 E-12 1.7 E-11 I-131 5.5 E-09 2.5 E-12 1.3 E-11 I-132 8.0 E-10 4.4 E-14 6.7 E-14 1.1 E-11 4.3 E-12 I-133 5.2 E-09 2.1 E-10 3.3 E-16 8.8 E-13 - - -
I-134 2.2 E-0 9 3.0 F-14 5.5 E-12 2.3 E-14 I-135 3.1 E-15 4.3 E-10 6.7 E-18 Kr-83m 2.5 E-0 9
- 2. 5 E-0 8 1.4 E-14 2.2 E-09 4.7 E Kr-85m 3.6 E-16 6.1 E-11 1.2 E-10 g Kr-85 2.1 E-0 7 5.0 E-09 8.9 E-14 1.2 E-09 ---
Kr-87 2.8 E-14 4 .1 E -0 9 1.9 E-16 Kr-88 3.2 E-08
-.0 E-11 2.6 E-16 1.3 E-11 ---
Kr-89 1.1 E-07 9.2 E-16 1.6 E-10 2.0 E-11 Xe-131m 6.7 E-15 1.1 E-09 3.7 E-10 Xe-133m 1.4 E-07 2.8 E-13 4.7 E-08 1.9 E-08 Xe-133 1.4 E-0 5 1.4 E-13 1.2 E-10 5.6 E12 Xe-135m 1.6 E-0 9 4.3 E-14 5.5 E-09 2.2 E-10 Xe-135 1.3 E-07 4.3 E-11 5.5 E-16 2.7 E-11 ---
Xe-137 7.6 E-10 5.0 E-15 3.8 E-10 ---
Xe-138 Note: " " indicates concentration <1.0 E-18 2 Concentrations after 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> recirculation through charcoal filters at 30,000 cfm.
b ('. 'd Amendment 17 Ub 1 of 1 C-E 9/30/75
SWESSAR-P1 TABLE 12.2.4-1 AIRDORNE RADIATION MONITORS Minimum Sensitivity Fangc Expected Monit or Number Medium (uCi/cc) (Deca des) Concentrations Locat ion Containment Atmosphere 1 Air Ta ble 12.2.3-6, Fig. 9.4.5.1-1 (2 detectors) Fi'. 12.2-1 Par ticula t e 1x 10-*(I-131) 4 Gas 1x 10-* [Xe-133) 4 Annnlus Building 3 Air Ta ble 12.2.3-6 Fig. 9.4.2-1 (2 cetectors)
Pa r ti cula t e 1x 10-t o (I- 131) 4 Gas 1x 10-* (Xe-133) 4 Solid Waste and DecouT- 1 Air Ba ckground Fig. 9.4.3-1 3 tamination Building (2 detectors) 4 Particulate 1x 10-89(I-131)
Gas 1x 10-* (Xe-13 3) 4 5
Fuel Building 2 Air Background Fig. 9.4.6-1
- .
- ' (2 detectors) 4
- ' *I Particulate 1x 10-80(I-131) 1x 10-6 (Xe-133) 4 Qy Gas es" y Air Ba dtground -
v7 Ventilation Systems 1 Multisampler (2 detectors)
{4#-[)
- Particulate 1x 10-s o (I-131) 4
,.]. Gas 1 x 10-* (Xe-133) 4 7-.
w Control I<oom Air 5 Air Ba ckground Fig. 9.4.1-1 1,
[r$ . }T-g Intakes 7' * (2 detectors) hhto Part.iculate 1x 10-s o (I- 131) 4 Gas 1x 10-* (Xe -133) 4 1
O CB d
t_ 4 1 of 1 Amendment 9 O 4/30/75
O SWESSAR-P1 TABLE 12.2.6-1 ESTIMA1T OF INHALATION AND WHOLE DODY DOSE RATES IN MAJOR PIANT BUILDINGS ext >ected Dose Rate (mrem /PJ')
B& W C-E W-41 Plant Building Nhole body Thyroid W-3S Whole Body Thyroid Whole Body Thyroid Whole Body "hyroid Containment Structures 5.6 E-01 2.0 E 01 3.7 E-01 1.3 E 01 3.2 E-01 1.1 E 01 4.3 E-01 1.3 E 01 Turbine building 4.8 E-07 1.6 E-03 8.4 E-08 1.4 E-04 8.7 E-08 1.6 E-04 1.1 E-07 2.0 E-04 Annulus building 3.9 E-03 3.3 E-02 3.1 E-03 2.2 E-02 2.8 E-03 "
9.4 E-02 3.2 E-03 1.1 E-01 Fuel building 3.3 E-04 2.7 E-02 4.5 E-04 3.4 E-01 1.0 E-03 5.2 E-02 3.5 E-04 2.8 E-02 After 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />' operation of containment atmosphere recirculation filters.
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S AFETY AN ALYSIS REPORT SWESSAR-Pl , , ,
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AMENDMENT 17 9/30/ / t.
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S AFETY AN ALYSIS REPORT /s (-, d _' .y 4 SWESSAR-PI C-E A M E N D M E N T 17 9/30/75
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AMEN MENT 12 6/16/15 f.) > !'t f.)
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SWESSAR-P1 12.3 HEALTH PHYSICS PROGRAM This section is within the Utility-Applicant's scope and SAR.
Space, however, has been provided for a comprehensive Health Physics Area shown on Fig. 1.2-1 and 1.2-10 which includes: 13 Health Physics Work Area Health Physics Storage Calibration Room Respirator Fitting Room Contaminated Clothing Storage Area Decontamination Showers Annulus Building Access Control Point Laundry Facilities Change Area Locker Space and Uncontaminated Shcuers Counting Room Chemistry Storage Area chemistry Lab The space allocated for these functions can be expanded to meet specific needs and requirements of a Utility-Applicant.
(Q) 12.3-1 Amendment 13 6/30/75
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SWESSAR-P1 12.4 RADIOACTIVE MATERIALS SAFETY 2
This material will be supplied in the Utility-Applicant's SAR.
4
() N l ' ;0 12.4-1 Amendment 2 8/30/74
VOLUnfE 9
. PRESSUR17FD WATER REAClUn i
REFnRENG NUCLEAR POWER PLANF SAFETY ANALYSIS REPORT SWESSAR-P1 STONE & WEBSTER ENGINEERING CORPORA' HON P. O. BOX 2325 BOSTON, MASSACHUSETIS 02107 Copyright 1974 by Stona & Webster Engineering Corporation All material herein is
- operty of said corporaton under which all copy and other rights have been rewrved and no such rights have been granted 2.ners. S'cne & Webster Engineering Corporation willin allinstances take such steps as are necessary for the preservation ofits rights and the enforcement of appleable la=.
6c9 351
SWESSAR-P1 TABLE OF CONTD;TS Section Voltrw CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT
1.1 INTRODUCTION
1 1.2 GENERAL PLANT DESCRIPTION 1 1.3 COVPARISON TABLES 2 1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS 2 1.5 REQUIREME!TTS FOR FURTHER TECHNICAL INFORMATION 2 1.6 MATERIAL INCORPORATED BY REFERENCE 2 1.7 TERMINOLOGY AND FLOW DIAGRAM SYMi>OLS 2 1.8 INTERFACE WITH NSSS VENDOR AND UTILITY-APPLICANT SAR 2 CHAPTER 2 13 SITE CHARACTERISTICS 2.1 GEOGRAPHY AND DEMOGPAPHY 2 2.2 NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES 2 2.3 METEOROIDGY 2 2.4 HYDROLOGIC E?UINEERI!G 2 2.5 GEOLOGY AND SEISMOLOGY 2 CHAPTER 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEPS 3.2 CONIVRMANCE WITH NRC GENERAL DESIGN CRITERIA 2 3.2 CLASSIFICATION OF STRUCTURES, SYSTEMS, AhT COMPONE?TTS 2 i Amendment 13 bb9 [,- [ c 6/30/75
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Seetion Volum3 CHAPTER 3 (CONT) 3.3 WIND AND TORNADO LOADINGS 2 3.4 WATER LEVEL (FLOGIq DESIGN 2 3.5 MISSILE PROTECTION 2 3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED 2 WITH THE POSTULATED RUPTURE OF PIPING 3.7 SEISMIC DESIGN 3 3.8 DESIGN OF CATEGORY I STRUCTURES 3 3.9 MECHANICAL SYSTEMS AND COMPONENTS 3 3.10 SEISMIC DESIGN OF CATEGORY I INSTRUMENTATION 3 AND ELECTRICAL EQUIPMENT 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND 3 ELECTRICAL EQUIPMENT APPENDIX 3A C CONFORMANCE WITH NRC REGULA'IVRY GUIDES 3A.1 DIVISION I REGULATORY GUIDES, POWER REACTORS 3 3A.2 OTHER DIVISION REGULA'IORY GUDES 3 APPENDIX 3B 20l COMPUTER PROGRAMS FOR ANALYSIS OF 3 THE CONTAINMENT STRUCTURE CHAPTER 4 20 REACTOR 3 b t; 'y , r-JJ>
ii Amendment 20 1/23/76
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Seetion Vo1ume CIULPTER S REACTOR COOLANT SYSTEM AND CONNECTED SYSTDIS
5.1 INTRODUCTION
3 5.2 IhTEGRITY OF RE ACTOR COOLANT PRESSURE BOUNDARY 3 5.3 TIIERMI4 HYDRAULIC SYSTEM DESIGN 3 S.4 REACTOR VESSEL AND APPURTENANCES 3 S.5 COMPO?ENT AND SUBSYSTEM DESIGN 3 CIIAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERI1 4 6.2 CONTAINMENT SYSTEMS 4 6.3 EMERGENCY CORE COOLING SYSTEM 4 6.4 IIABITABILITY SYSTDIS 4 APPENDIX 6A DATA FOR DETERMINING THE IvDI?TE REMOVAL EFFECTIVENESS FOR THE CONTAINMENT ATEOSPHERE 6A.1 THE SPRAY DROP DISTRIBUTION AND CHARACTERISTIC SPRAY DROP DIAMETERS FOR THE SPRAY HEADERS 4 6A.2 TIIE SPRAY COVERAGE OF THE CONTAINMENT 4 ATMOSPHERE 6A.3 AN71YSIS OF SLCRS PERFORMANCE 4 APPENDIX 6B LOCTIC INTERFACE WITH NSSS SUPPLIED DATA 20 6B.1 POST - REFIDOD PERIOD 4 6B.2 LONG TEPR MASS - ENERGY RELEASES 4 bb9 iii Amendment 20 1/23/76
SWESSAR-P1 TABLE OF COITTENTS (CONT)
Section Volume CHAPTER 7 INSTRUMENTATION AND CONTROLS
7.1 INTRODUCTION
4 7.2 REACTOR TRIP SYSTEM 4 7.3 ENGINEERED SAFETY FEATURES SYSTEM 4 7.4 SYSTEMS REQUIRED FOR SAFE Snufv0FN 4 7.5 SAFETY RELATED DISPLAY INSTRUMENTATION 4 7.6 ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR 4 SAFETY 7.7 CONTROL SYSTEMS NOT REQUIRED FOR SAFETY 4 7.8 INTERFACE REQUIREMENTS 5 CHAPTER 8 ELECTRIC POWER
8.1 INTRODUCTION
5 8.2 OFFSITE POWER SYSTEM S 8.3 ONSITE POWER SYSTEM S 8.4 INTERFACE DESIGN INFORMATION 5 20 8.5 ELECTRIC HEAT TRACING S CHAPTER 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE A?O HANDLING 6
.g b
6b9 555 iv Amendment 20 1/23/76
SWESSAR-P1 TABLE OF CONTENTS (cot?T)
Volurne Section CHAPTER 9 (CONT) 6 9.2 WATER SYSTEMS 6
9.3 PROCESS AUXILIARIES AIR CONDITIONING, HEATING, COOLING, AND 6 9.4 VENTILATION SYSTEMS OTHER AUXILIARY SYSTEMS 6 9.5 CHAPTER 10 STEAM AND POWER COINERSION SYSTD4 10.1
SUMMARY
DESCRIPTION 7 10.2 TURBINE-GENERATOR AND TURBINE STEAM SYSTEM 7 10.3 MAIN STEAM SYSTEM 7 10.4 OTHEP FEATURES OF STEAM AND POWEP COINERSION 7 SYSTEM CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE ITEMS 7 11.2 PADIOACTIVE LIQUID WASTE SYSTD4 7 RADIOACTIVE GASEOUS WASTE SYSTEM 8 11.3 11.4 PROCESS AND EFFLUENT RADIATION MONITORING 8 SYSTEM 11.5 RADIOACTIVE SOLID WASTE SYSTEM 6 11.6 OFFSITE RADIOLOGICAL MONITORING PROGRAF 8
/,. 1 '~
001 J J U.
v Amendment 20 1/23/76
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Section Volume CHAPTER 12 RADIATION PROTECTION 12.1 SHIELDING 8 12.2 VENTILATION 8 12.3 HEALTH PHYSICS PROGRAM 8 12.4 RADIOACTIVE MATERIALS SAFETY (FSAP) 8 CHAPTER 13 CONDUCT OF OPERATIONS 13.1 ORGANIZATION STRUCTURE 9 13.2 TRAINING PROGRAM 9 13.3 EMERGENCY PLANNING 9 13.4 REVIEW AND AUDIT 9 PIANT PROCEDURES 13.5 9 13.6 PLANT RECORDS 9 13.7 INDUSTRIAL SECURITY 9 CHAPTER 14 INITIAL TESTS AND OPEPATIONS 9 CFAPTER 15 ACC11n.aT ANALYSIS 15.1 GLNERAL 9 CHAPTER 16 TECHNICAL SPECIFICATIONS 16.1 DEFINITIONS 9 16.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 9 vi Amendment 20 i ,
1/23/76 i t- -
b01 sJ/
SWESSAR-P1 TABLE OF CONTENTS (CONT)
Section Vollme CPAPTER 16 (CONT) 16.3 LIMITING CONDITIONS FOR OPERATION 9 16.4 SURVEILLANCE REQUIREMENTS 9 16.5 DESIGN FEATURES 9 16.6 AD?'INISTRATIVE CONTROLS 9 CHAPTER 17 QUALITY ASSUPANCE 17.1 QUALITY ASSURANCE DURING DESIGN AhT 9 CONSTRUCTION 17.2 QUALITY ASSURANCE FOR STATION OPERATION 9 APPENDIX A ENCLOSURE BUILDING WIT:iOUT MIXING 9 APPENDIX B ENCLOSURE EUILDING WITH MIXING 9 f){)l 'J0 vii Amendment 20 1/23/76
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