ML22111A120

Amendment 27 to Updated Final Safety Analysis Report, Chapter 12, Radiation Protection

ML22111A120
Person / Time
Site:	Saint Lucie
Issue date:	03/04/2022
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22111A137	List: L-2022-018, Amendment 27 to Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant ML22063B120 ML22111A100 ML22111A101 ML22111A102 ML22111A103 ML22111A104 ML22111A105 ML22111A107 ML22111A108 ML22111A109 ML22111A112 ML22111A113 ML22111A114 ML22111A115 ML22111A116 ML22111A117 ML22111A118 ML22111A119 ML22111A120 ML22111A121 ML22111A122
References
L-2022-018
Download: ML22111A120 (178)

Text

UFSAR/St. Lucie - 2 RADIATION PROTECTION CHAPTER 12 TABLE OF CONTENTS Section Title Page 12.0 RADIATION PROTECTION .......................................................................... 12.1-1 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) ........................................ 12.1-1 12.1.1 POLICY CONSIDERATIONS ........................................................................ 12.1-1 12.1.2 DESIGN CONSIDERATIONS ....................................................................... 12.1-3 12.1.3 OPERATIONAL CONSIDERATIONS............................................................ 12.1-9 12.2 RADIATION SOURCES ................................................................................ 12.2-1 12.2.1 CONTAINED SOURCES............................................................................... 12.2-1 12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES..................................... 12.2-6 REFERENCES ............................................................................................ 12.2-10 12.3 RADIATION PROTECTION DESIGN FEATURES ....................................... 12.3-1 12.3.1 FACILITY DESIGN FEATURES .................................................................... 12.3-1 12.3.2 SHIELDING ................................................................................................... 12.3-6 12.3.3 VENTILATION ............................................................................................. 12.3-12 12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION .................................................................................. 12.3-14 REFERENCES ............................................................................................ 12.3-23 12.3A TMI SHIELDING STUDY ............................................................................ 12.3A-1 12.3A.1 INTRODUCTION ........................................................................................ 12.3A-2 12.3A.2 SOURCE TERMS ...................................................................................... 12.3A-2 12.3A.3 RADIOACTIVE SYSTEMS ......................................................................... 12.3A-3 12.3A.4 VITAL AREAS REQUIRING OCCUPANCY/ACCESS ............................... 12.3A-4 12.3A.5 DOSE RATE AND DOSE CALCULATIONS .............................................. 12.3A-4 REFERENCES ........................................................................................... 12.3A-5 12.4 DOSE ASSESSMENT (HISTORICAL) .......................................................... 12.4-1 12.4.1 ANTICIPATED DOSE RATES....................................................................... 12.4-1 12-i Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 Section Title Page 12.4.2 ESTIMATE OF EXPOSURE OF PLANT PERSONNEL ................................ 12.4-1 REFERENCES .............................................................................................. 12.4-5 12.5 HEALTH PHYSICS PROGRAM .................................................................... 12.5-1 12.5.1 ORGANIZATION ........................................................................................... 12.5-1 12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES ............................... 12.5-2 12.5.3 PROCEDURES ............................................................................................. 12.5-6 12-ii Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 CHAPTER 12 RADIATION PROTECTION LIST OF TABLES Table Title Page 12.2-1 MAXIMUM NEUTRON SPECTRA OUTSIDE REACTOR VESSEL ............T12.2-1 12.2-2 MAXIMUM GAMMA SPECTRA OUTSIDE REACTOR VESSEL ................T12.2-2 12.2-3 N-16 ACTIVITY ...........................................................................................T12.2-3 12.2-4 SHUTDOWN GAMMA SPECTRA OUTSIDE REACTOR VESSEL ............T12.2-4 12.2-5 SHUTDOWN MATERIAL ACTIVATION SPECTRA ....................................T12.2-5 12.2-6 PRESSURIZER STEAM SECTION ACTIVITY ...........................................T12.2-6 12.2-7 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW)............................................................T12.2-7 12.2-8 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW)............................................................T12.2-8 12.2-9 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW)..........................................................T12.2-10 12.2-10 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW)..........................................................T12.2-11 12.2-11 LIQUID WASTE MANAGEMENT SYSTEM MISCELLANEOUS COMPONENTS ESTIMATED MAXIMUM INVENTORIES (Curies)..........T12.2-12 12.2-12 LIQUID WASTE MANAGEMENT SYSTEM MISCELLANEOUS COMPONENTS ESTIMATED AVERAGE INVENTORIES (Curies)..........T12.2-13 12.2-13 SOLID WASTE MANAGEMENT SYSTEM COMPONENT MAXIMUM INVENTORIES (Curies) (2700 MW)..........................................................T12.2-14 12.2-14 SOLID WASTE MANAGEMENT SYSTEM COMPONENT AVERAGE INVENTORIES (Curies) (2560 MW)..........................................................T12.2-15 12.2-15 GASEOUS WASTE MANAGEMENT SYSTEM COMPONENT MAXIMUM INVENTORIES (Curies) (2700 MW) .......................................T12.2-16 12.2-16 GASEOUS WASTE MANAGEMENT SYSTEM COMPONENT AVERAGE INVENTORIES (Curies) (2560 MW) .......................................T12.2-17 12.2-17 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENT MAXIMUM INVENTORIES (Curies) (2700 MW) .......................................T12.2-19 12-iii Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 Table Title Page 12.2-18 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) .......................................T12.2-20 12.2-19 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW) .......................................T12.2-21 12.2-20 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED AVERAGE INVENTORIES (Curies) (2560 MW)..................T12.2-22 12.2-21 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED MAXIMUM INVENTORIES (Curies) (2700 MW)..................T12.2-24 12.2-22 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) .......................................T12.2-25 12.2-23 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW) .......................................T12.2-27 12.2-24 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) .......................................T12.2-28 12.2-25 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED MAXIMUM INVENTORIES (Curies) (2700 MW)..................T12.2-30 12.2-26 CHEMICAL & VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED AVERAGE INVENTORIES (Curies) (2560 MW)..................T12.2-31 12.2-27 SAFETY INJECTION SYSTEM COMPONENT MAXIMUM INVENTORIES (Curies) (2700 MW)..........................................................T12.2-33 12.2-28 SAFETY INJECTION SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW)..........................................................T12.2-34 12.2-29 SAMPLING SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW) ...................................................................................T12.2-35 12.2-30 SAMPLING SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) ...................................................................................T12.2-36 12.2-31 SAMPLING SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW) ...................................................................................T12.2-38 12.2-32 SAMPLING SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) ...................................................................................T12.2-39 12.2-33 FUEL POOL SYSTEM COMPONENTS MAXIMUM INVENTORIES (Curies) (2700 MW) ...................................................................................T12.2-40 12.2-34 FUEL POOL SYSTEM COMPONENTS AVERAGE INVENTORIES (Curies) (2560 MW) ...................................................................................T12.2-41 12.2-35 SPENT FUEL GAMMA SOURCE (MEV/WATT-S) ...................................T12.2-43 12-iv Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 Table Title Page 12.2-36 FISSION PRODUCT GAMMA SOURCE IN CONTAINMENT BUILDING (MEV/SEC) ................................................................................................T12.2-44 12.2-37 ASSUMPTIONS & PARAMETERS USED TO CALCULATE AIRBORNE CONCENTRATIONS .................................................................................T12.2-45 12.2-38 AVERAGE AIRBORNE C/MPC IN REACTOR AUXILIARY BUILDING CONTAINMENT AND FUEL HANDLING BUILDING ................................T12.2-47 12.2-39 REACTOR AUXILIARY BUILDING & FUEL HANDLING BUILDING ROOM BY ROOM C/MPC & WHOLE BODY DOSE COMMITMENT VALUES ....................................................................................................T12.2-48 12.2-40 POST ACCIDENT SAMPLING SYSTEM COMPONENT MAXIMUM INVENTORIES (CURIES) .........................................................................T12.2-51 12.3-1 ALLOWABLE DOSE RATES.......................................................................T12.3-1 12.3-2 AREA RADIATION MONITORS ..................................................................T12.3-2 12.3-3 AIRBORNE RADIATION MONITORS .........................................................T12.3-4 12.3A-1 CORE INVENTORY ................................................................................. T12.3A-1 12.3A-2 UNDILUTED REACTOR COOLANT SOURCE TERMS .......................... T12.3A-2 12.3A-3 RECIRCULATED (CONTAINMENT SUMP) WATER SOURCE TERMS T12.3A-3 12.3A-4 CONTAINMENT ATMOSPHERE SOURCE TERMS ............................... T12.3A-4 12.3A-5 CONTAINMENT PLATEOUT SOURCE TERMS ..................................... T12.3A-5 12.3A-6 SYSTEMS POTENTIALLY CONTAINING HIGH LEVELS OF RADIOACTIVE MATERIALS .................................................................... T12.3A-6 12.3A-7 AREAS IDENTIFIED IN SHIELDING REVIEW AS REQUIRING ACCESSIBILITY FOLLOWING AN ACCIDENT ....................................... T12.3A-7 12.4-1 DATA FROM OPERATING PWR PLANTS .................................................T12.4-1 12.4-2 YEARLY AVERAGES AND GRAND AVERAGE FOR NUMBER OF PERSONNEL AND MAN-REM DOSES FOR OPERATING PWR PLANTS ......................................................................................................T12.4-3 12.4-3 DISTRIBUTION OF MAN-REM DOSES FOR VARIOUS FUNCTIONS FOR OPERATING LIGHT WATER REACTORS ........................................T12.4-4 12.4-4 OCCUPATIONAL RADIATION EXPOSURES EXPERIENCED AT OPERATING NUCLEAR POWER PLANTS BY TYPE OF WORK .............T12.4-5 12.4-5 ESTIMATED ANNUAL EXPOSURE DOSES TO PLANT PERSONNEL ....T12.4-7 12.4-6 PREDICTED DOSES TO PLANT PERSONNEL DURING ROUTINE PATROL ......................................................................................................T12.4-8 12-v Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 Table Title Page 12.4-7 ESTIMATED DOSES TO PLANT PERSONNEL DURING PERIODIC MAINTENANCE ........................................................................................T12.4-11 12.5-1 COUNTING ROOM EQUIPMENT ...............................................................T12.5-1 12.5-2 PORTABLE INSTRUMENTS FOR RADIATION MONITORING .................T12.5-2 12.5-3 SELF-READING DOSIMETERS .................................................................T12.5-3 EC 291 105 12-vi Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 RADIATION PROTECTION CHAPTER 12 LIST OF FIGURES Figure Title 12.3-1 Deleted 12.3-2 Deleted 12.3-3 Reactor Bldg Miscellaneous Structure Steel SH. 20 12.3-4 Radiation Zones Outside of Reactor, Reactor Auxiliary and Fuel Handling Bldgs 12.3-5 Radiation Zones Reactor Containment Building Floor Elev. 18, 23 and 62 12.3-6 Radiation Zones Reactor Containment Building Floor Elevation 45 12.3-7 Radiation Zones Reactor Auxiliary Building Floor Elevation - 0.5 12.3-8 Radiation Zones Reactor Auxiliary Building Floor Elevation - 19.5 12.3-9 Radiation Zones Reactor Auxiliary Building Floor Elevation - 43 12.3-10 Radiation Zones Reactor Auxiliary Building Floor Elevation - 62 12.3-11 Radiation Zones Fuel Handling Building Floor Elevations 19.5, 48 and 62 12.3-12 Radiation Zones Fuel Handling Building Vertical Sections 12.3-13 HVAC - Fuel Handling Building Plans, Sections & Details 12.3-13a Location of CIS and Post Accident Radiation Monitors 12.3-14 Flow Diagram Miscellaneous Sampling Systems 12.3A-1 TMI Radiation Zone Map 12.3A-2 TMI Radiation Zone Map 12.3A-3 TMI Radiation Zone Map 12.3A-4 TMI Radiation Zone Map 12.3A-5 Access Routes to Vital Areas 12.3A-6 Access Routes to Vital Areas 12.3A-7 Access Routes to Vital Areas 12-vii Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 12.0 RADIATION PROTECTION 12.1 ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES ARE AS LOW AS REASONABLY ACHIEVABLE (ALARA) 12.1.1 POLICY CONSIDERATIONS 12.1.1.1 Management Policy It is the policy of Florida Power & Light Company (FP&L) Management to keep personnel doses within prescribed State and Federal limits and further to maintain personnel exposure and contamination levels as low as is reasonably achievable (ALARA), consistent with job performance.

In striving to achieve the proper ALARA balance between the benefits of power production and the possible detrimental effects of radiation exposure, the operation and maintenance of a nuclear power plant is such that some personnel might be exposed to a radiation environment.

The dose limits recommended by the National Council on Radiation Protection and those adopted by the NRC are thought to involve a risk that is small as compared to everyday hazards of life. Nevertheless, it is felt that every effort should be made to reduce the exposure from ionizing radiation to the lowest level that is reasonably achievable.

Administrative and Operational procedures have been developed to implement the ALARA concept. These procedures are based on applicable regulations and many man-years of practical experience in all areas of constructing, operating, and maintaining a nuclear power plant.

12.1.1.2 Organization In conformance with applicable regulations the Florida Power & Light Company radiation protection policy and program is established by the Corporate Health Physics staff. The detailed procedures for implementing the program at each plant are established by the Radiation Protection Manager. The Radiation Protection Manager at each plant is responsible for ensuring the exposures are ALARA. The Corporate Health Physics staff is responsible for ensuring that the ALARA policy is implemented.

Radiological engineering support and ALARA coordination is provided at the plant level by an ALARA Supervisor and plant engineering personnel. The plant efforts are assisted by the Juno Health Physics Personnel and the Company's Engineering Department.

12.1.1.3 Responsibilities 12.1.1.3.1 Corporate Health Physics Manager The Corporate Health Physics Manager is responsible for:

a. establishing the radiological protection policies and standards to be used at the nuclear stations and assist the plant, as needed, in their implementation;

b. providing technical assistance for conducting the program; 12.1-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2

c. assuring management that personnel exposures are maintained ALARA;

d. monitoring radiological indicators for trends; and

e. recommending modifications to the program as required by experience, regulatory and technical changes.

12.1.1.3.2 Corporate and Plant Management Corporate and Plant Management are responsible for:

a. ensuring implementation of the company's radiation protection program;

b. providing a favorable attitude toward radiation safety;

c. committing to radiation protection policies;

d. providing sufficient staff, equipment and funding to the Health Physics programs for implementing effective radiological control programs; and

e. holding supervisors and workers accountable for their performance.

12.1.1.3.3 Radiation Protection Manager The Radiation Protection Manager is responsible for:

a. establishing and implementing a complete and comprehensive radiation protection program; and

b. establishing the necessary facilities to control personnel and public exposure to radiation and radioactive materials.

12.1.1.4 Training A training program covering the fundamentals of radiation protection and exposure control procedures has been implemented by the Radiation Protection Manager. All individuals whose job responsibilities require (1) working with radioactive materials, (2) entering radiation areas, (3) directing the activities of others who work with radioactive materials or enter radiation areas receive this training. The detailed training program is referenced in Section 13.2.

12.1.1.5 Review of New or Modified Designs and Equipment Selection

a. All work groups (e.g., maintenance, operations, radiation protection, technical support, engineering, and safety groups) are involved with the review of the design of the facility and the selection of equipment. This coordinated effort ensures that the objective of the ALARA program is achieved.

b. Design concepts and station features reflect consideration of the station personnel activities in order to reduce exposures to station personnel.

c. Selection of equipment reflects the objectives of the ALARA program. Equipment has been selected based on reliability and serviceability in radiation areas.

12.1-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.1.2 DESIGN CONSIDERATIONS The guidelines and criteria followed in order to achieve ALARA personnel exposure doses comply with the provisions of Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable (Nuclear Power Reactors),"June 1978, (R3), both in philosophy and in detailed features. In addition, experience gained from prior plant design (including St. Lucie Unit 1) as well as operating plants is factored into the design.

Special attention is given to equipment layout, pipe routing, shielding design, and operating procedures to reduce personnel exposures received during maintenance.

The following design guidelines have been used for plant layout:

a. Radiation Zone Segregation The division of the plant into radiation zone areas is one of the fundamental features of ALARA design. Subsection 12.3.2.2 describes the criteria for the establishment of each zone; the dose limits, access restrictions and warning sign requirements are presented.

Controlled and uncontrolled areas are separated by access control points. The controlled zones encompass areas which either house radioactive equipment, or which can become contaminated during movement of personnel or components.

Special attention is devoted to the segregation of potentially radioactive systems from those systems and areas which do not see any radioactivity. Even though non-radioactive areas (personnel facilities) and systems (electrical systems, control room systems) are located in the Reactor Auxiliary Building, they are kept separated from radioactive systems by suitable shielding walls so that the maximum dose rate in these areas does not exceed 0.25 mrem/hr. The means of ordinarily gaining access from nonradioactive to radioactive areas is through the access control points. However temporary check points and controlled areas may be established to facilitate maintenance operations.

The plant ventilation systems are used to control airborne activity. They are designed so that flow is from lower to higher potential activity areas. These systems are discussed in Subsection 12.3.3 and also in Section 9.4.

b. Shielding Design Considerations Original design of the plant shielding was performed assuming a core power level of 2700 MWt, a 12-month fuel cycle length, and system activity levels stemming from one percent fuel cladding defects. The plant shielding was re-evaluated for the extended power uprate assuming a core thermal power of 3030 MWt and an 18-month fuel cycle. Taking into consideration the conservative analytical techniques used to establish the original shielding design and the plant Technical Specifications, which restrict the reactor coolant activity to levels significantly less than 1% failed fuel (due to fuel defects), it is concluded that the increase in the core power level and current operation with an 18-month fuel cycle will have not significant impact on plant shielding adequacy and safe plant operation.

12.1-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 Shielding is designed so that the zone levels chosen for ALARA considerations and plant operation are not violated. The sources are considered to be within the contained boundaries, e.g., tanks, valves, and ion exchangers. The thicknesses of the shielding is selected such that the highest dose rate chosen for the desired zone level is not violated. Since the shield walls, floors and ceilings are of constant thicknesses, other areas within the zone are less than this dose rate.

Permanent shielding is provided whenever feasible to minimize the need for portable shielding during maintenance and inspection operations.

Shield plugs and concrete hatches are designed with stepped sides to minimize radiation streaming.

Labyrinths and/or shielding doors are used to eliminate radiation streaming through access openings out of the shielded cubicles. Labyrinth walls as well as the shielded cubicle walls are generally designed extending between floors so that the floors above and below also act as shielding barriers. Partial height shielding walls are used in cubicles housing horizontal tanks and pumps, the activity of which is either expected to be low or concentrated at the tank bottom.

The height of the shielding barriers in such cases is determined by geometric (line-of-sight) considerations, i.e., a person standing at the farthest point from the cubicle in question, (in an accessible area) cannot see any of the radioactive components within the cubicle including piping.

As an example, the equipment drain tanks and chemical drain tanks are separated from their respective pumps by a wall which is over seven feet high; thus, a person in the pump cubicle is not directly exposed to the piping in the tank cubicle. Similarly, the wall separating the pumps from the corridor is sufficiently high so that a person standing next to the pump wall on the north side of the accessible corridor cannot see any radioactive piping housed in either the tank or pump cubicle (see Figure 1.2-12).

Penetrations for piping and ducts in shielding walls are designed so as not to be on a direct line with a major radioactive source. Openings for the penetrations are kept as small as possible. Where required, boots are provided or openings are packed to meet exposure criteria.

c. Equipment and Piping Layout To the extent possible (taking into account separate criteria), systems and components handling high activity fluids are located in the same general area of the plant.

Systems and components containing high activity fluids are separated from systems and components containing low activity fluids. In addition, the latter are located away from "clean" systems and components. With this arrangement, heavy shielding walls can be shared by various components, thus minimizing space requirements and costs.

To the extent possible components and piping which do not normally contain radioactivity, nor can be expected to become radioactive, are separated from the 12.1-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 radioactive portions of the plant. This simplified division of the plant into controlled and uncontrolled areas aids in traffic flow, reduces the possibility that radioactive piping is run in clean areas, minimizes the need for shielded pipe chases, and helps in controlling contamination spread into clean areas.

Shielded chases are provided where required in order to reduce exposure from radioactive piping. Components belonging to a system handling radioactive fluids, are generally arranged on that side of a wall which faces away from a corridor. This is done to minimize piping runs in corridors, thereby minimizing the need for the use of shielded pipe chases. The height of piping above floors is considered in design. In addition, when performing maintenance on a piece of equipment, personnel are not exposed to radiation from equipment in another process line.

Major components in a given system are separated to the extent practical from other major components by use of shielding walls. Tanks, filters, heat exchangers and demineralizers are the major sources of activity in any given system; these components are isolated in shielded cubicles. Valves and pumps are not expected to contain more volumetric activity than the piping associated with them; thus the extent to which they are housed in individual cubicles is dictated by the calculated piping activity. The extent to which each tank, filter and demineralizer is housed in individual cubicles is dictated by the projected activity content and the capability for taking components housed in the same cubicle out of service at the same time. Tanks which are known to contain potentially high radioactivity are housed in individual cubicles: examples are holdup tanks, boric acid holdup tank, and spent resin tank. Tanks which are of fairly low activity (e.g., floor drain) tanks, laundry tanks and condensate tanks) or tanks which have the potential for occasional high activity (e.g., boric acid makeup tanks) are housed in common cubicles with space provided for portable shields.

Equipment and components which require manual operation, visual inspection or are expected to need servicing are arranged in the lowest possible radiation field.

As an example, a radioactive fluid handling system, which consists of a tank, pump, associated valving, sampling lines and instrumentation, is laid out as follows:

The tank, which is a component usually requiring the least servicing, is normally placed in a separate shielded cubicle. Serviceable valving and piping are excluded from this cubicle to the maximum practical extent. The pump and valves, which require maintenance, are placed in a separate cubicle, and radioactive piping within the cubicle is kept to a minimum. When practicable or where required, further compartmentalization is achieved by placing the pumps andvalves in their own individual cubicles. Sampling lines and instrumentation requiring personnel attendance are brought outside the shield walls to low level radiation zones (II or III). Within the containment, such instrumentation is brought to the lowest possible radiation area. A description of radiation zones is provided in Subsection 12.3.2.2.

12.1-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 Clearance is provided within shielding cubicles housing components potentially requiring maintenance and repair (such as valves, pumps, or heat exchangers),

so that work on the particular component is allowed. Overly restrictive compartments, while saving space, require lengthier stays by maintenance or repair personnel, since their work will be hampered and inefficient. Furthermore, large compartments ease installation of temporary shielding barriers should they be required. Provisions for slings, chains, and hoists, etc., are provided for equipment with consideration given to their size, expected frequency of maintenance, and handling problems.

Instrumentation (flow meters, level gauges) is located in the lowest practical radiation area and is readily accessible to operators. Operators are not required to enter cubicles housing radioactive components to either read or activate instruments.

An exception is instrumentation located by necessity within the containment, and even there efforts are made to locate it in the lowest possible radiation field.

Radioactive piping is either ran in shielded pipe chases or within shielded cubicles housing low maintenance equipment, where practical. Radioactive piping is not run unshielded in accessways, and the amount of radioactive piping near frequent maintenance equipment has been minimized. Clean and radioactive lines do not share the same pipe chases. Provisions are made to support the weight of shielding wrapped around radioactive pipes, where required.

Piping two inches in diameter or smaller which is field run is classified in two categories: special and miscellaneous. Special piping includes seismic Category I piping, radioactive waste process piping, and other piping which by nature of its fluid content is expected to carry radioactivity in sufficient quantity to require special consideration. Miscellaneous piping includes drains, vents, sampling lines and instrumentation lines.

Routing of special class piping is accomplished in compliance with the same radiation protection criteria used for routing of larger pipes. Drawings indicating the routing were originally made either by A/E home office or field personnel. In either case the drawings are reviewed both by a shielding engineer and by main office supervisory personnel who sign such drawings.

A shielded sampling room is provided for the major radioactive samples (reactor coolant). Sampling takes place under a hood with sufficient shielding provided between the sample piping and the actual sample location (see Subsection 9.3.2). Care is taken to locate the sampling points for local samples in the lowest practical radiation area.

Shielded valve stations for systems handling radioactive fluids are employed, wherever feasible, in order to perform valve maintenance without drainage of associated equipment. To further minimize personnel exposure, remotely operated valves are utilized where practical and necessary. If manual valves are employed, extension rods through a shield wall to an accessible low radiation 12.1-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 area are utilized as necessary. In order to greatly reduce the problems of radiation streaming, the reach rod penetrations are generally offset from the major source of radiation (usually a tank), or are provided with an internal offset.

Waste Management System processes are controlled locally either from control panels located in low radiation areas or from positions shielded from the process equipment and lines. This reduces exposures accumulated during operation of the Waste Management System.

d. Crud and Contamination Control To the largest extent possible, but especially for components and piping handling primary coolant, radioactive resins and concentrates, connections are provided for flushing portions of a system. The portion of a system to be flushed is dictated by the expected frequency of maintenance of the components housed in a shielded cubicle, the size of the cubicle, the numbers of components housed therein, the geometry of piping and the valving arrangement.

In general, components and piping within a single cubicle are flushable. Flexible hoses from the low drain point to the floor drain are considered acceptable for flushing procedures except for resin carrying lines. These are flushed back to their respective tanks (spent resin and dewatering tanks). If a hose is used to flush, the flushing connection is brought out to a relatively accessible area.

Equipment and piping handling radioactive fluids are designed and laid out in a manner to minimize radioactive crud retention and to facilitate decontamination.

Some specific criteria applied are:

1) Preferential use is made of round bottomed tanks to minimize crud buildup at the bottom. Effluent process lines are placed as close as possible to the bottom, preferably at the lowest point in the tank. Drain valves are positioned away from the tank bottom, and preferably located in an accessible area. Operators do not need to crouch below tanks to operate any valve, thereby reducing their exposure from deposited crud.

2) Piping runs are sloped wherever possible to minimize crud buildup, and to assist in the removal of crud deposits.

3) The number of elbows, tees, vees, deadlegs, standpipes, etc., are minimized since these can act as crud traps.

4) Large radius elbows are preferred.

5) Orifices are installed in vertical rather than horizontal runs where there is an option.

6) Two and one half inch and larger pipe consumable inserts are employed at welds where possible to minimize use of backing rings.

7) Low leakage valves are specified for radioactive fluid service.

12.1-7 Amendment No. 24 (09/17)

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8) Decontamination of equipment is done in place wherever possible.

Flushing and cleaning capabilities for decontamination prior to maintenance are provided where required.

9) Suitable coatings are used on floor surfaces of areas in a controlled zone.

Personnel decontamination facilities and shielded cubicles have surfaces suitably treated for easy decontamination.

10) Cubicles, with the exception of the spent resin tank cubicle and the boric acid holdup tank cubicle, are provided with drains to the respective equipment and floor drain collection headers. Floor drains in the spent resin and boric acid holdup tank cubicles are plugged and the equipment drains raised so that in the unlikely event of a rupture of a tank there is no transport of resins or concentrates through the floor or equipment drainage system. Curbing is provided in those cubicles to contain any accidental spills. Backgassing through the drain system from the collection tanks is prevented by providing a loop seal between the tank cubicle drain and the header. Additional loop seals are provided in the drain systems, some of which are upstream of the header.

e. Additional Containment Shielding Features Accessways are provided to normally inaccessible areas between the primary and secondary shield to allow for inservice inspection. The shielding thus provides limited entry into the secondary shield area for inspection and short maintenance. Longer stays may require decontamination.

f. Design Review During plant design and construction, radiation protection engineers perform shielding and exposure calculations, and work with the designers and other engineers to ensure that the guidelines noted above are followed. They advise on the most desirable design option for radiation protection when alternate designs are possible in satisfying the process requirements. Typically, this involves a study of the exposures which are likely to result from alternate designs and the selection of the option resulting in the lowest exposure.

The radiation protection design review is an ongoing review throughout the phases of the design. Calculations, which form the bases for such reviews, are made in the following categories (calculational methods are presented in Subsection 12.3.2.3) for the following:

1) source calculations and/or verification for each system involved in the review

2) shield thickness calculations

3) activation calculations if the system involves neutron radiation

4) mapping of exposure dose rates for the particular layout 12.1-8 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 In addition, a review of shielding design in the operating St. Lucie Unit 1 plant has resulted in several changes in the St. Lucie Unit 2 plant design.

12.1.3 OPERATIONAL CONSIDERATIONS To the greatest extent possible, all efforts have been made to factor operational considerations of radiation exposure in the plant layout and system design by utilizing the guidelines of Subsection 12.1.2 throughout the design effort. These guidelines incorporate known operational considerations derived from experience. Information from operating plants (especially St. Lucie Unit 1) have continuously been factored in the design as it progressed to reflect new operational considerations, not well known or appreciated at the time the guidelines were initially developed.

In this regard, the guidelines of Regulatory Guides 8.8 (R3) and 8.10, "Operating Philosophy For Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable,"September 1975 (R1) will continue to be factored in the design.

Administrative procedures are established in the plant, which along with design shielding, ensure that the exposure to personnel is kept as low as reasonably achievable during plant operation and maintenance. These procedures are incorporated within the Formal Plant Procedures prior to initial plant startup, and are constantly updated to reflect PWR operating exposure records. The procedures applicable to ALARA are: Health Physics Manual, Respiratory Protection Program, Radiation Work Permits, Protective Clothing Requirements, Personnel Monitoring, Radiation Surveys, and Decontamination Procedures. These procedures are presented in Section 12.5.

These written procedures are intended to conform to 10 CFR 20, to minimize exposures. They include procedures that provide guidelines for the use of special temporary shielding to augment the permanent shielding in areas where extensive maintenance is required and the radiation levels are such that the estimated time of occupancy would cause the individual's exposure to approach or exceed administrative guidelines. Such procedures detail the type and amount of shielding to be used. Typically, materials are available for temporary shielding against gamma and neutron radiation, respectively.

Radiation Work Permits specify the use of protective clothing and health physics respiratory equipment when required during the performance of maintenance and repair involving potential contamination.

Time-radiation dose schedules are developed indicating the number of hours per day and week personnel can work in different radiation fields before exceeding administrative exposure guidelines. Such schedules are implemented by a combination of radiation work permits and administrative procedures on work in radiation controlled areas.

Implementation of all of these procedures is the specific responsibility of the Radiation Protection Manager.

The radiation protection program together with a more detailed description of planning and developing the above procedures for radiation exposurerelated operation is given in Section 12.5. Summarized below, however, are the salient points.

Stationary detectors are provided and portable ones are provided for monitoring direct, surface and airborne activity in operating and maintenance locations, particularly in those locations susceptible to radioactive contamination.

12.1-9 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 Strict administrative controls, based on a radiation zone classification system, govern personnel activities in radiation areas. Control points will separate non-radioactive from radioactive areas.

Increasing levels of control and of warning signs are applied to access the areas of increasing radiation zone levels. Access to the highest radiation zone areas (Zone V) is controlled through approved procedures and the issuance of a radiation work permit. These areas are normally inaccessible, and entrance requires obtaining a key. Key control is maintained through approved procedures.

Health physics personnel periodically observe jobs in progress in the radiation controlled areas and make radiation surveys in accordance with plant procedures to ensure that exposure to radiation and contamination levels are kept as low as practicable. Administrative exposure guidelines are designed to evenly distribute each individual's annual exposure. When personnel are assigned to a job or a location where there exists the possibility that administrative guidelines may be exceeded, health physics personnel investigate the exposure records of the personnel involved and authorize work by such personnel only after the current exposure history, the amount the guidelines might be exceeded, and the alternatives that are available to complete the job under consideration have all been considered. This will call for consultation and pre-job planning between the supervisor in charge of the job and health physics personnel. The same applies to the issuance of a radiation work permit.

12.1-10 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.2 RADIATION SOURCES 12.2.1 CONTAINED SOURCES The radiation sources used for the original design and analysis of the shielding requirements are based on a core power level of 2700 Mwt. These sources are given for all phases of plant operation including full power operation, shutdown conditions, refueling operations, and for various postulated accidents. They include the neutron and gamma fluxes outside the reactor vessel, the reactor coolant activation, fission and corrosion product activities, deposited corrosion product sources on reactor coolant equipment surfaces, spent fuel handling sources and accident sources. In addition, radiation source for various auxiliary systems are also tabulated. Source term information presented in Tables 12.2-1 through 12.2-40 is based on original plant design and is considered historical. The tables are retained to maintain a record of original licensing basis.

12.2.1.1 Reactor Coolant Fission and Corrosion Product Activity The activities utilized for shielding calculating are shown in Tables 11.1-2 and 11.1-3.

The basis for the above reactor coolant radioisotope concentrations is specified in Table 11.1-1, and the manner in which the concentrations have been derived is given in Subsections 11.1.1.1 and 11.1.2.1. Those values form the basis for the component inventories discussed in Section 12.2.1.6.

12.2.1.2 Neutron and Gamma Fluxes Outside the Reactor Pressure Vessel at Full Power Operation The primary radiations within the containment during normal operation are neutrons and gamma rays emanating from the reactor core. Tables 12.2-1 and 12.2-2 list neutron and gamma multigroup fluxes in the reactor cavity at the side of the reactor vessel; these tables are based on nuclear parameters discussed in Chapter 4. The maximum neutron spectra during full power operation, divided into 22 energy groups, for a point at the core midplane, one-half foot from the reactor vessel surface, is shown in Table 12.2-1. The maximum gamma spectra during full power operation for the same location, divided into 18 energy groups is shown in Table 12.2-2.

12.2.1.3 Reactor Coolant N-16 Activity The neutron activation product, nitrogen-16, is the predominant activity in the reactor coolant pumps, steam generators, and reactor coolant piping during power operation; and it determines the shielding requirements for the secondary shield wall and portions of the Chemical and Volume Control System (CVCS). It is discussed in Subsection 11.1.3.1. The N-16 activity at various points is determined from the activity at the reactor vessel outlet nozzle and taking into account the decay, due to transit time, between the nozzle and the point of interest.

The original licensing basis N-16 activities for various points in the Reactor Coolant System are presented in Table 12.2-3.

12.2-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.2.1.4 Reactor Coolant System Sources at Shutdown Following shutdown, residual radiation from the Reactor Coolant System is due to fission product decay gamma radiation emanating from the core; material activation sources, radioactive corrosion products which have deposited on surfaces, and the fission and corrosion products in the reactor water. The reactor vessel maximum decay spectra, divided into 13 energy groups for two days after shutdown, is shown in Table 12.2-4 for a point at core midplane, one-half foot from reactor vessel surface.

The material activation spectra at the same location and decay time is shown in Table 12.2-5.

The material activation sources include contributions from the reactor vessel wall, core barrel, core shroud, and reactor coolant water. Activation of the reactor vessel insulation and supports is considered and evaluated on the basis of the fluxes computed to exist in the annular cavity space during full power operation.

The fission and corrosion product reactor coolant inventory assumed to be present for shutdown can be calculated from the values in Tables 11.1-2 and 11.1-3, by correcting for decay up to the point in time of interest.

The activity of radioactive crud and its thickness on Reactor Coolant System surfaces have been evaluated using data from eight pressurized water reactors. For the circulating crud, the observed activities were compared to activity values recommended for design by the Draft ANSI N237 Standard, Source Term Specification, 1976.

While operating data crud activity concentrations in the reactor coolant are generally lower than those recommended by the ANSI Standard N237, the N237 values have been used as design source terms, since they yield a more conservative design. The design source terms are shown in Section 11.1, for circulating crud.

The zinc injection system will inject zinc acetate solution into the RCS. The zinc solution will displace cobalt from the resident RCS oxide films. This will result in a temporary increase in the RCS coolant activities from 58Co and 60Co. The elevation in the radiocobalt activity will be temporary, lasting only until the plant oxide layers are fully conditioned. The period of increase in activity will be approximately two operating cycles depending upon the rate of zinc injection.

The temporary increased activity in the RCS coolant as a result of zinc injection will cause an increase in resin and filter activity for approximately two cycles. Increased resin and filter usage during the first two cycles will occur, but resin depletion during cycle operation is not expected.

12.2.1.5 Pressurizer Activity The liquid section of the pressurizer has source terms due to plateout of radioactive crud plus dispersed fission and corrosive product activity in the water. Typical crud film thickness are shown in Table 11.1-10.

As an upper limit, the liquid section will have dispersed fission and corrosion product activity equal to the reactor coolant activity with one percent failed fuel specified in Table 11.1-2.

The activity in the steam section is due to the buildup of the gaseous fission products Xe and Kr from one percent failed fuel. This activity is shown in Table 12.2-6.

12.2-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.2.1.6 Contained Sources in Other Plant Systems The source intensity in equipment and pipelines handling radioactive fluids is determined from activity in the reactor coolant by considering the processes that the reactor coolant has undergone prior to entering equipment and piping (dilution, filtering, demineralization, relay, change of phase, etc.)

In all cases the process or combination of processes leading to the highest activity is considered for conservatism. The individual component inventories are presented in Tables 12.2-7 through 12.2-34, and are based on the Reactor Coolant System specific activities, as noted in Subsection 12.2.1.1, and the assumptions listed.

12.2.1.6.1 Liquid Waste Management System (LWMS) Sources Assumptions for this system include:

a. Filter and demineralizer on-line times are one core cycle.

b. For the average case, isotopic concentrations in system piping are the result of applying expected component decontamination factors (as discussed in Subsection 11.2.1) and average Reactor Coolant System specific activities of Table 11.1-3.

For the maximum case, isotopic concentrations in system piping are calculated using the decontamination factors of Subsection 11.2.1 and the maximum reactor coolant specific activities of Table 11.1-2.

c. Isotopic inventories on filters are the result of assuming 90 percent removal of input corrosion product activity in the average and maximum cases, respectively.

d. Isotopic inventories on ion exchangers for the average and maximum cases are the result of assuming 90 percent removal of input fission and corrosion product activity, respectively.

e. A concentration factor of 20 is assumed for the waste concentrator for purposes of both maximizing the concentrator activity and maximizing inputs to the concentrate tanks.

Section 11.2 contains a description of the LWMS. Tables 12.2-7 through 12.2-12 contain the LWMS component inventories.

12.2.1.6.2 Solid Waste Management System (SWMS) Sources The Solid Waste Management System is discussed in Section 11.4. Input activities are based upon sources calculated for various interfacing systems' components. The bases for the spent resin tank inventory are:

a. One bed change per year for all ion exchangers, except the purification ion exchangers. In their case, only one bed changeout between them is accounted for.

12.2-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2

b. Allowance is made for sluice water used, by assuming the spent resin tank is full with the water/resin slurry. Hence total spent resin tank volume is used to calculate the radionuclide concentrations within it.

The spent resin tank inventories are listed on Tables 12.2-13 and 12.2-14. The waste concentrator bottom inventories are listed in Table 12.2-12.

12.2.1.6.3 Gaseous Waste Management System (GWMS) Sources Sources, volumes, and flow rates to the gas surge header are shown in Table 11.3-4. Input activities are based on one percent failed fuel and application of applicable DF and gas-to-liquid partition coefficients. The equipment decontamination factors are shown in Table 11.2-2(a) and the gas to liquid partition coefficients are presented on Table 11.2-2(b). (See Subsection 11.2.1). Section 11.3 discusses the GWMS. Tables 12.2-15 and 12.2-16 contain the GWMS component inventories.

12.2.1.6.4 Chemical and Volume Control System (CVCS) Sources Reactor coolant system maximum activities serve as the input to the CVCS source calculations.

Assumptions for this system include:

a. Filter and ion exchanger on-line times of one core cycle.

b. Isotopic concentrations in system piping (including pumps and coolers) are the result of applying expected component decontamination factors as given in Reference 1.

c. Isotopic inventories on filters are the result of assuming 90 percent removal of input corrosion product activity.

d. Isotopic inventories on ion exchanger are the result of assuming removal efficiencies as defined in Reference 1.

e. Reactor coolant pump seal bleedoff is assumed to be identical to reactor coolant activity.

f. Reactor drain tank activity is assumed to be identical to reactor coolant activity.

g. Equipment drain tank activity is assumed to be equal to 0.35 percent of reactor coolant activity.

h. An average concentration factor of 10 is assumed for the boric acid concentrator for purposes of both maximizing the concentrator activity and maximizing inputs to the boric acid holding tank. Credit is taken for decay in filling this tank.

i. A refueling shutdown and startup is assumed for determination of holdup tank activity. Credit is taken for decay in filling the tank, but no credit is taken for decay between steps in the shutdown operation.

The CVCS is described in Subsection 9.3.4. Tables 12.2-17 through 12.2-26 contain the CVCS component inventories.

12.2-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.2.1.6.5 Safety Injection System (SIS) Sources The values shown on Tables 12.2-27 and 28 are based upon the following:

a. The Reactor Coolant System specific activities are shown in Table 11.1-2. They serve as the input basis for the shutdown data. The shutdown values serve as the basis for shielding design in this area.

b. Post-LOCA containment sump activities (per Regulatory Guide 1.7, "Control of Combustible Gas Concentrations in Containment Following a LOCA, September 1976 (R1). These are the input bases for the core LOCA data. The core LOCA data is used in the analysis of component design life integrated dose.

c. System design and operation parameters as discussed in Section 6.3.

d. All N-16 is assumed to have decayed.

12.2.1.6.6 Primary Sampling System (PSS) Sources The bases for the PSS sources are the Reactor Coolant System specific activities (Table 11.1-2) and the PSS parameters and system operation as discussed in Subsection 9.3.2.

The PSS source terms are listed for the major components in Tables 12.2-29 through 32. The only components of significant volume within the system are the heat exchangers and sample vessels.

12.2.1.7 Fuel Pool Cooling and Purification System (FPS) Sources The bases for the maximum activities in the spent fuel pool are discussed in Subsection 11.1.1.4. The maximum fission and corrosion product activities in the spent fuel pool are given in Table 11.1-12. The expected fuel pool activities are discussed in Subsection 11.1.2.2 and presented in Table 11.1-13.

Assumptions for FPS component sources are identical to items a) through d) of Subsection 12.2.1.6.4. The FPS is described in Subsection 9.1.3. Fuel pool system component inventories are given in Tables 12.2-33 and 34. A purification rate of 150 gpm is assumed.

12.2.1.7.1 Spent Fuel Handling and Transfer The spent fuel assemblies are the predominant long term source of radiation in the containment after plant shutdown for refueling. A reactor operating time necessary to establish near-equilibrium fission product buildup for the reactor at rated power is used in determining the source strength. The initial fuel composition that produces the maximum decay source is used.

The spent fuel decay gamma source is given in Table 12.2-35.

12.2.1.7.2 Spent Fuel Storage and Transfer The predominant radioactivity sources in the spent fuel storage and transfer areas in the fuel building are the spent fuel assemblies. The spent fuel decay gamma source to be used in shielding design is given in Table 12.2-35.

12.2-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.2.1.8 Accident Sources The accident source terms which are employed to determine shielding requirements for emergency accessways, control room and containment shielding, as well as potential doses to equipment inside containment following a loss-of-coolant accident (LOCA) are shown in Table 12.2-36. The table assumes a release to containment of the activity stated in TID 14844(2), namely 100 percent noble gases, 50 percent halogens, and one percent of the solid fission product inventory.

12.2.1.8.1 Safety Injection System Sources The accident sources for the Safety Injection System are shown in Tables 12.2-27 and 12.2-28.

See Subsection 12.2.1.6.5 for further information.

12.2.1.8.2 Post-Accident Sampling System (PASS) Sources A brief description of the PASS may be found in Subsection 9.3.6. The post accident component inventories are shown in Table 12.2-40.

Assumptions for this system include:

a. No decay occurs from time of release from the core, to appearance in the component.

b. RCS activity is based on a release of 100 percent noble gases, 50 percent halogens and one percent solid fission product core inventory.

c. Containment atmosphere activity is based on a release of 100 percent noble gases, and 25 percent halogen fission product core inventory.

12.2.1.9 Other Sources A remote radiation controlled area for the long term storage of certain byproduct material that is used primarily during plant outages has been developed and is located between the site intake and discharge canals. Reference 5 discusses the requirements for this remote facility.

12.2.2 AIRBORNE RADIOACTIVE MATERIAL SOURCES Equipment cubicles, corridors, and areas normally occupied by personnel could contain small amounts of airborne radioactivity as a result of equipment leakage. For the purpose of evaluating the potential exposure to personnel from this activity, this subsection presents a description of the sources of activity, and the models and assumptions and parameters used to evaluate airborne radionuclide levels in the Reactor Auxiliary Building, the Reactor Building and the Fuel Handling Building. Table 12.2-37 indicates assumptions and parameters used in the analysis.

Reactor coolant source terms assume 0.12 percent failed fuel. The sources are determined for each area assuming that the leaking fluid contains a fraction of the reactor coolant activity and that leakage occurs in that area. This fraction is determined from process consideration of leaking fluid (i.e., amount of filtering, degassing and demineralization prior to leak). The leak rate is based on typical data from operating plants. The equilibrium airborne concentration is then determined by use of the standard equation of build-up and removal, where build-up is 12.2-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 caused by leakage and removal is caused by both radioactive decay and ventilation. A neglible amount of radioactivity is expected to be released due to removal of the reactor vessel head, movement of spent fuel or relief valves venting. Therefore, contributions from these sources to airborne activity are not considered.

Note: The following discussion and related tables use units of MPC, Maximum Permissible Concentration and not units of DAC, Derived Air Concentration, as is currently used by 10 CFR 20. At the time of plant license MPCs were the appropriate units of measure per 10 CFR 20. Since this information represents the plant's licensing basis and it is not used with respect to plant operations, including radiological controls, it is considered historical and will not be revised.

The isotopic airborne concentrations, as a fraction of maximum permissible concentration (MPC) in air (per 10 CFR 20), are calculated for those normally occupied buildings where a potential for exposure exists. These values are presented in Table 12.2-38 and indicate that the dose to a critical organ of a worker, when adjusted on the basis of weekly occupancy, would be well below the maximum allowable limit. Furthermore, the dose calculated, based on these values, would be highly conservative since the MPCs given in 10 CFR 20 are based on infinite cloud assumptions while the volumes of applicable areas are limited. To account for anticipated operational occurrences and reactor coolant concentration at 1.0 percent, it is estimated that values presented would be increased by a factor of eight. At that level of airborne concentration, the occupancy in a few areas may be limited to a few minutes duration.

For airborne radionuclides dose calculations in the Reactor Building, it is assumed that there is no exhaust ventilation, thereby making the Reactor Building isolated.

The airborne concentration of radioisotope in an area having a constant leak rate, source strength and exhaust rate, can be calculated by the equation given below.

where, Ci(t) = airborne concentration of the ith radioisotope at t, ci/cm3 MPCi = maximum permissible concentration for the ith isotope W = leak rate of fluid, cm3 /hr ai = concentration of ith isotope in the reactor coolant, ci/cm3 SS = source strength defined as fraction of reactor coolant present in the leakage liquid PFi = partition factor of ith isotope i = total removal rate constant for ith isotope, di + e hr-1 di = decay constant for ith isotope, hr-1 12.2-7 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 e = removal rate constant due to exhaust, hr-1 e =

where:

1.76 x 106 = conversion factor f = exhaust rate (cfm) t = time interval, hours v = free volume of the area where leak occurs, cm3 For small rooms and other operating areas where e is found to be much larger than di for most of the radionuclides, the peak or equilibrium activity(Ceqi):, is given by the following equations:

Using the above type of equation, a detailed analysis is also performed on both the Reactor Auxiliary Building and Fuel Handling Building for rooms containing equipment from which potential airborne exposures exist as a result of possible leakage. For these rooms in the Reactor Auxiliary Building and the Fuel Handling Building, airborne radionuclide concentrations in the form of fraction of maximum permissible concentration ( C/MPC) are calculated as well as the whole body dose per hour occupancy, resulting from inhalation and external exposure.

Assumptions used in the analysis are the same as those listed in Table 12.2-37. In addition, the inhalation and external whole body dose conversion factors are taken from Tables E-7 of Regulatory Guide 1.109," Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

October 1977 (R1) and Table D-3 of WASH-1258(3), respectively. The WASH-1258 values for a semi-infinite cloud model result in highly conservative dose values. Individual rooms are identified by elevation, installed equipment and equipment leakage rates. Radionuclide concentrations in areas such as corridors which adjoin equipment housing areas are calculated assuming the air in the corridor can be contaminated by exhaust from those areas.

Table 12.2-39 lists the results of the analysis In order to calculate tritium concentration in the Fuel Handling Building the following equations(4) were used to calculate evaporation rate from the fuel pool:

12.2-8 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 And ventilation rate:

where, wp = Evaporation rate of water, lbm/hr.

A = Surface area of pool, ft2

= Air velocity over surface, ft/min.

w = Saturation pressure of vapor at the surface water temperature, inches of mercury.

a = Saturation pressure of vapor at room air dew point temperature, inches of mercury.

y = Latent heat at pool surface water temperature, Btu/lb.

f = Ventilation rate ft3/min.

Wi = Moisture content of indoor air, lbm/lbm of dry air Wo = Moisture content of outdoor air lbm/lbm of dry air The results of the tritium concentration in the fuel pool area are shown in Table 12.2-38.

12.2-9 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 SECTION 12.2: REFERENCES

1. NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from a P.W.R., May 1976.

2. DiNunno, Anderson, Bakes and Anderson, "Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Atomic Energy Commission, March 23, 1962.

3. "Numerical Guides for Design Objectives and Limiting Conditions for Operation to meet the Criterion "As Low As Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents," WASH-1258, AEC, July 1973.

4. ASHRAE Guide and Data Book Applications 1968 Heating Refrigeration Ventilating and Air Conditioning - American Society of Heating Refrigerating and Air Conditioning Engineers.

5. Safety Evaluation PSL-ENG-SENS-98-011, Revision 1, On-Site Storage of Radioactive Materials in a Remote RCA, March, 1998.

12.2-10 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-1 MAXIMUM NEUTRON SPECTRA OUTSIDE REACTOR VESSEL*

Neutron Spectra Neutron Energy (Mev) (Neutrons/cm2-s) 13.60 6.78(+06)**

11.10 2.13(+07) 9.10 4.32(+07) 7.27 7.82(+07) 5.66 1.22(+08) 4.51 9.90(+07) 3.53 1.72(+08) 2.73 2.16(+08) 2.40 7.15(+07) 2.09 4.43(+08) 1.47 1.61(+09) 8.30(-01) 5.46(+09) 3.30(-01) 1.62(+10) 5.70(-02) 9.43(+09) 1.96(-03) 2.32(+09) 3.42(-04) 2.01(+09) 6.50(-05) 1.31(+09) 1.98(-05) 8.30(+08) 6.90(-06) 8.56(+08) 2.09(-06) 5.72(+08) 7.60(-07) 4.46(+08) 2.50(-08) 2.46(+09)

• At core midplane, one half foot from vessel surface.

• • Numbers in parentheses denote powers of ten.

T12.2-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-2 MAXIMUM GAMMA SPECTRA OUTSIDE REACTOR VESSEL*

Gamma Spectra Gamma Energy (Mev) (Gamma/cm2-s) 9.00 2.87(+08) 7.25 1.23(+09) 5.75 8.51(+08) 4.50 7.36(+08) 3.50 1.04(+09) 2.75 7.11(+08) 2.25 1.08(+09) 1.83 8.64(+08) 1.50 9.42(+08) 1.16 1.18(+09) 0.90 9.13(+08) 0.70 1.14(+09) 0.50 2.65(+09) 0.35 1.52(+09) 0.25 2.34(+09) 0.15 2.49(+09) 0.075 3.06(+09) 0.025 2.98(+06)

• At core midplane, one half foot from vessel surface.

T12.2-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-3 N-16 ACTIVITY (Historical)

Activity Location (Disintegrations/cm3 - s)

Vessel Outlet Nozzle 4.82(+06)*

Vessel Outlet Line Midpoint 4.76(+06)

Steam Generator Midpoint 3.90(+06)

Pump Midpoint 3.16(+06)

Vessel Inlet Line Midpoint 3.06(+06)

• Numbers in parentheses denote powers of ten.

T12.2-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-4 SHUTDOWN GAMMA SPECTRA OUTSIDE REACTOR VESSEL*

Decay Gamma Gamma Energy (Mev) (Gamma/cm2-s) 2.75 1.84(+04)**

2.25 6.86(+04) 1.83 1.60(+05) 1.50 2.70(+05) 1.16 3.80(+05) 0.90 2.96(+05) 0.70 3.76(+05) 0.50 5.08(+05) 0.35 3.40(+05) 0.25 5.46(+05) 0.15 7.54(+05) 0.075 1.28(+05) 0.025 3.96(+02)

• At core midplane, one half foot from vessel surface, and two days after shutdown.

• • Numbers in parentheses denote powers of ten.

T12.2-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-5 SHUTDOWN MATERIAL ACTIVATION SPECTRA*

Material Activation Gamma Energy (Mev) (Gamma/cm2-s) 2.75 2.25 1.83 1.50 1.16 3.24(+06)**

0.90 1.02(+07) 0.70 1.88(+06) 0.50 2.78(+06) 0.35 2.16(+06) 0.25 4.38(+06) 0.15 5.14(+06) 0.075 9.36(+05) 0.025 3.14(+03)

• At core midplane, one half foot from vessel surface, and two days after shutdown.

• • Numbers in parentheses denote powers of ten.

T12.2-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-6 PRESSURIZER STEAM SECTION ACTIVITY Isotope Activity* (Ci/cc)

KR-85 M 1.2(-01)**

KR-85 2.7(+02)

KR-87 2.8(-02)

KR-88 1.9(-01)

XE-131M 3.1(+01)

XE-133 9.1(+02)

XE-135 1.4(+00)

XE-138 3.7(-03)

• Assumes that all XE and KR isotopes build up for 292 days and are supplied to the pressurizer at continuous primary water rate of 1.5 gpm, with complete stripping of the gas from water.

• • Denotes power of ten T12.2-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-7 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Equipment Chemical Laundry Waste Nuclide Drain Tank Drain Tank Drain Tank Condensate Tank N-16 0. 0. 0. 0.

KR-85M 0. 0. 0. 0.

KR-85 0. 0. 0. 0.

KR-87 0. 0. 0. 0.

KR-88 0. 0. 0. 0.

XE-131M 0. 0. 0. 0.

XE-133 0. 0. 0. 0.

XE-135 0. 0. 0. 0.

XE-138 0. 0. 0. 0.

BR-84 3.8E-04 7.5E-03 2.2E-04 2.3E-07 RB-88 3.6E-02 7.0E-01 2.0E-02 2.1E-05 RB-89 8.6E-04 1.7E-02 4.9E-04 5.1E-07 SR-89 7.6E-05 1.5E-03 4.3E-05 4.5E-08 SR-90 2.5E-06 4.9E-05 1.4E-06 1.5E-09 Y-90 2.5E-06 4.9E-05 1.4E-06 1.5E-09 SR-91 4.9E-05 9.5E-04 2.8E-05 2.9E-08 Y-91 8.1E-05 1.6E-03 4.6E-05 4.8E-08 ZR-95 1.0E-04 2.0E-03 5.8E-05 6.1E-08 MO-99 6.1E-03 1.2E-01 3.5E-03 3.6E-06 RU-103 8.5E-05 1.6E-03 4.8E-05 5.1E-08 RU-106 2.3E-05 4.4E-04 1.3E-05 1.3E-08 TE-129 1.7E-04 3.3E-03 9.8E-05 1.0E-07 I-129 4.5E-10 8.8E-09 2.6E-10 2.7E-13 I-131 4.9E-02 9.5E-01 2.8E-02 2.9E-05 TE-132 4.6E-03 9.0E-02 2.6E-03 2.8E-06 I-132 1.0E-02 2.0E-01 5.8E-03 6.1E-06 I-133 6.1E-02 1.2E+00 3.5E-02 3.6E-05 TE-134 4.4E-04 8.5E-03 2.5E-04 2.6E-07 I-134 6.1E-03 1.2E-01 3.5E-03 3.6E-06 CS-134 2.1E-03 4.1E-02 1.2E-03 1.3E-06 I-135 3.0E-02 5.9E-01 1.7E-02 1.8E-05 CS-136 1.5E-03 2.8E-02 8.3E-04 8.7E-07 CS-137 5.8E-03 1.1E-01 3.3E-03 3.5E-06 CS-138 1.2E-02 2.4E-01 7.0E-03 7.3E-06 BA-140 1.2E-04 2.3E-03 6.7E-05 6.9E-08 LA-140 1.1E-04 2.2E-03 6.4E-05 6.7E-08 PR-143 9.7E-05 1.9E-03 5.5E-05 5.8E-08 CE-144 6.1E-05 1.2E-03 3.5E-05 3.6E-08 CR-51 3.4E-05 6.7E-04 2.0E-05 2.1E-09 MN-54 5.7E-06 1.1E-04 3.3E-06 3.4E-10 FE-55 2.0E-05 5.7E-04 1.7E-05 1.7E-09 FE-59 1.9E-05 3.6E-04 1.1E-05 1.1E-09 CO-58 2.9E-04 5.7E-03 1.7E-04 1.7E-08 CO-60 3.7E-05 7.2E-04 2.1E-05 2.2E-09 Note: E-xx denotes powers of 10 T12.2-7 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-8 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

Equipment Chemical Laundry Waste Nuclide Drain Tank Drain Tank Drain Tank Condensate Tank N-16 0. 0. 0. 0.

KR-83M 0. 0. 0. 0.

KR-85M 0. 0. 0. 0.

KR-85 0. 0. 0. 0.

KR-87 0. 0. 0. 0.

KR-88 0. 0. 0. 0.

KR-89 0. 0. 0. 0.

XE-131M 0. 0. 0. 0.

XE-133M 0. 0. 0. 0.

XE-133 0. 0. 0. 0.

XE-135M 0. 0. 0. 0.

XE-135 0. 0. 0. 0.

XE-137 0. 0. 0. 0.

XE-138 0. 0. 0. 0.

BR-83 4.0E-05 5.3E-05 2.8E-05 6.9E-09 BR-84 7.3E-06 5.9E-06 7.0E-06 3.4E-10 BR-85 8.2E-08 6.5E-08 8.5E-08 3.5E-13 I-130 2.8E-05 1.3E-04 1.7E-05 9.4E-09 I-131 4.7E-03 8.1E-02 2.7E-03 2.6E-06 I-132 8.1E-04 1.0E-03 5.7E-04 1.3E-07 I-133 5.5E-03 4.0E-02 3.2E-03 2.2E-06 I-134 2.1E-04 1.8E-04 1.8E-04 1.5E-08 I-135 2.2E-03 6.2E-03 1.4E-03 6.3E-07 RB-86 1.6E-06 2.9E-05 9.1E-07 9.2E-10 RB-88 3.1E-04 2.5E-04 3.2E-04 7.9E-09 CS-134 4.9E-04 9.5E-03 2.8E-04 2.9E-07 CS-136 2.5E-04 4.5E-03 1.4E-04 1.4E-07 CS-137 3.6E-04 6.9E-03 2.0E-04 2.1E-07 SR-89 6.4E-06 1.2E-04 3.6E-06 3.7E-09 SR-90 1.9E-07 3.6E-06 1.1E-07 1.1E-10 SR-91 8.3E-06 3.1E-05 5.0E-06 2.6E-09 Y-90 2.0E-08 2.6E-07 1.1E-08 9.7E-12 Y-91 1.2E-06 2.2E-05 6.7E-07 6.9E-10 Y-91M 1.6E-06 1.3E-06 1.3E-06 1.1E-10 Y-93 4.4E-07 1.7E-06 2.6E-07 1.4E-10 ZR-95 1.1E-06 2.1E-05 6.2E-07 6.4E-10 NB-95 9.0E-07 1.7E-05 5.1E-07 5.3E-10 MO-99 1.4E-03 1.9E-02 8.3E-04 7.2E-07 TC-99M 5.5E-04 1.4E-03 3.4E-04 1.5E-07 RU-103 8.2E-07 1.6E-05 4.7E-07 4.8E-10 RU-106 1.9E-07 3.6E-06 1.1E-07 1.1E-10 RH-103M 2.1E-07 1.9E-07 1.8E-07 1.7E-11 RH-106 4.5E-10 3.6E-10 4.7E-10 0.

Note: E-xx denotes powers of 10 T12.2-8 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-8 (Cont'd)

Equipment Chemical Laundry Waste Nuclide Drain Tank Drain Tank Drain Tank Condensate Tank TE-125M 5.3E-07 1.0E-05 3.0E-07 3.1E-10 TE-127M 5.2E-06 9.9E-05 3.0E-06 3.1E-09 TE-127 1.1E-05 4.0E-05 6.5E-06 3.4E-09 TE-129M 2.5E-05 4.7E-04 1.4E-05 1.5E-08 TE-129 8.7E-06 8.0E-06 6.9E-06 8.4E-10 TE-131M 3.8E-05 3.5E-04 2.2E-05 1.6E-08 TE-131 2.4E-06 1.9E-06 2.4E-06 8.7E-11 TE-132 4.6E-04 6.5E-03 2.6E-04 2.3E-07 BA-137M 3.7E-06 3.0E-06 3.9E-06 1.4E-11 BA-140 4.0E-06 7.1E-05 2.3E-06 2.3E-09 LA-140 2.4E-06 2.7E-05 1.4E-06 1.1E-09 CE-141 1.3E-06 2.4E-05 7.3E-07 7.4E-10 CE-143 6.3E-07 6.1E-06 3.6E-07 2.7E-10 CE-144 6.0E-07 1.2E-05 3.4E-07 3.5E-10 PR-143 9.0E-07 1.6E-05 5.1E-07 5.1E-10 PR-144 5.2E-08 4.2E-08 5.4E-08 1.3E-12 NP-239 1.9E-05 2.5E-04 1.1E-05 9.5E-09 CR-51 3.4E-05 6.4E-04 2.0E-05 2.0E-09 MN-54 5.7E-06 1.1E-04 3.3E-06 3.4E-10 FE-55 2.9E-05 5.7E-04 1.7E-05 1.7E-09 FE-59 1.9E-05 3.5E-04 1.1E-05 1.1E-09 CO-58 2.9E-04 5.6E-03 1.7E-04 1.7E-08 CO-60 3.7E-05 7.2E-04 2.1E-05 2.2E-09 Note: E-xx denotes powers of 10 T12.2-9 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-9 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Nuclide Waste Filter Laundry Filter N-16 0. 0.

KR-85M 0. 0.

KR-85 0. 0.

KR-87 0. 0.

KR-88 0. 0.

XE-131M 0. 0.

XE-133 0. 0.

XE-135 0. 0.

XE-138 0. 0.

BR-84 4.0E-06 6.6E-07 RB-88 3.7E-04 6.1E-05 RB-89 8.9E-06 1.5E-06 SR-89 7.8E-07 1.3E-07 SR-90 2.6E-08 4.3E-09 Y-90 2.6E-08 4.3E-09 SR-91 5.1E-07 8.4E-08 Y-91 8.4E-07 1.4E-07 ZR-95 1.1E-06 1.7E-07 MO-99 6.3E-05 1.0E-05 RU-103 8.8E-07 1.5E-07 RU-106 2.3E-07 3.9E-08 TE-129 1.8E-06 3.0E-07 I-129 4.7E-12 7.7E-13 I-131 5.1E-04 8.4E-05 TE-132 4.8E-05 7.9E-06 I-132 1.1E-04 1.7E-05 I-133 6.3E-04 1.0E-04 TE-134 4.5E-06 7.5E-07 I-134 6.3E-05 1.0E-05 CS-134 2.2E-05 3.6E-06 I-135 3.2E-04 5.2E-05 CS-136 1.5E-05 2.5E-06 CS-137 6.0E-05 1.0E-05 CS-138 1.3E-04 2.1E-05 BA-140 1.2E-06 2.0E-07 LA-140 1.2E-06 1.9E-07 PR-143 1.0E-06 1.7E-07 CE-144 6.3E-07 1.0E-07 CR-51 1.6E-02 9.5E-03 MN-54 1.5E-02 8.4E-03 FE-55 9.1E-02 5.3E-02 FE-59 1.4E-02 8.2E-03 CO-58 3.4E-01 1.9E-01 CO-60 1.2E-01 7.1E-02 Note: E-xx denotes powers of 10 T12.2-10 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-10 LIQUID WASTE MANAGEMENT SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

WASTE LAUNDRY WASTE LAUNDRY NUCLIDE FILTER FILTER NUCLIDE FILTER FILTER N-16 0. 0. Y-91M 9.3E-09 4.0E-09 KR-83M 0. 0. Y-93 2.9E-09 7.9E-10 KR-85M 0. 0. ZR-95 1.1E-08 1.9E-09 KR-85 0. 0. NB-95 9.2E-09 1.5E-09 KR-87 0. 0. MO-99 1.3E-05 2.5E-06 KR-88 0. 0. IC-99M 3.5E-06 1.0E-06 KR-89 0. 0. RU-103 8.4E-09 1.4E-09 XE-131M 0. 0. RU-106 1.9E-09 3.2E-10 XE-133M 0. 0. RH-103M 1.3E-09 5.3E-10 XE-133 0. 0. RH-106 2.7E-12 1.4E-12 XE-125M 0. 0. TE-125M 5.4E-09 9.1E-10 XE-135 0. 0. TE-127M 5.3E-08 8.9E-09 XE-137 0. 0. TE-127 7.2E-08 2.0E-08 XE-136 0. 0. TE-129M 2.6E-07 4.3E-08 BR-83 2.4E-07 8.3E-08 TE-129 5.2E-08 2.1E-08 BR-84 4.4E-08 2.1E-08 TE-131M 3.0E-07 6.7E-08 BR-65 4.9E-10 2.6E-10 TE-131 1.4E-08 7.2E-09 I-130 1.9E-07 5.0E-08 TE-132 4.2E-06 1.9E-07 I-131 4.6E-05 8.2E-06 BA-137M 2.2E-08 1.2E-08 I-132 4.9E-06 1.7E-06 BA-140 4.0E-08 6.8E-09 I-133 4.1E-05 9.7E-06 LA-140 2.0E-08 4.2E-09 I-134 1.2E-06 5.3E-07 CE-141 1.3E-08 2.2E-09 I-135 1.4E-05 4.1E-06 CE-143 5.1E-09 1.1E-09 RB-86 1.6E-08 2.76-09 CE-144 6.2E-09 1.0E-09 RB-88 1.8E-06 9.6E-07 PR-143 9.0E-09 1.5E-09 CS-134 5.1E-06 8.4E-07 PR-144 3.1E-10 1.6E-10 CS-136 2.5E-u6 4.3E-07 NP-239 1.7E-07 3.4E-08 CS-137 3.7E-06 6.1E-07 CR-51 1.6E-02 9.5E-03 SR-89 6.5E-08 1.1E-08 MN-54 1.5E-02 8.4E-03 SR-90 1.9E-09 3.2E-10 FE-55 9.1E-02 5.3E-02 SR-91 5.5E-08 1.5E-08 FE-59 1.4E-02 8.2E-03 Y-90 1.7E-10 3.4E-11 CO-58 3.3E-01 1.9E-01 Y-91 1.2E-08 2.0E-09 CO-60 1.2E-01 7.1E-02 Note: E-xx denotes powers of 10 T12.2-11 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-11 LIQUID WASTE MANAGEMENT SYSTEM MISCELLANEOUS COMPONENTS ESTIMATED MAXIMUM INVENTORIES (CURIES)

(2700 MW)

CONDENSATE WASTE(1)

NUCLIDE ION EXCHANGER CONCENTRATOR N-16 0. 0.

KR-85M 0. 0.

KR-85 0. 0.

KR-87 0. 0.

KR-88 0. 0.

XE-131M 0. 0.

XE-133 0. 0.

XE-135 0. 0.

XE-138 0. 0.

BR-84 2.8E-07 5.4E-03 RB-88 1.4E-05 5.0E-01 RB-89 3.0E-07 1.2E-02 SR-89 1.2E-04 1.1E-03 SR-90 1.6E-05 3.5E-05 Y-90 2.2E-07 3.5E-05 SR-91 6.4E-07 6.9E-04 Y-91 1.5E-04 1.1E-03 ZR-95 2.1E-04 1.4E-03 MO-99 5.5E-04 8.5E-02 RU-103 1.1E-04 1.2E-03 RU-106 1.1E-04 3.2E-04 TE-129 2.7E-07 2.4E-03 I-129 3.0E-09 6.3E-09 I-131 1.3E-02 6.9E-01 TE-132 4.9E-04 6.5E-02 I-132 3.2E-05 1.4E-01 I-133 1.7E-03 8.5E-01 TE-134 4.2E-07 6.1E-03 I-134 7.2E-06 8.5E-02 CS-134 1.2E-02 3.0E-02 I-135 2.8E-04 4.3E-01 CS-136 6.2E-04 2.0E-02 CS-137 3.8E-02 8.2E-02 CS-138 8.9E-06 1.7E-01 BA-140 4.9E-05 1.6E-03 LA-140 6.1E-06 1.6E-03 PR-143 4.3E-05 1.4E-03 CE-144 2.9E-04 8.5E-04 CR-51 3.1E-06 4.8E-05 MN-54 2.8E-06 8.0E-06 FE-55 1.7E-05 4.1E-05 FE-59 2.7E-06 2.6E-05 CO-58 6.4E-05 4.1E-04 CO-60 2.3E-05 5.2E-05 1). The waste concentrator is no longer used T12.2-12 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-12 LIQUID WASTE MANAGEMENT SYSTEM MISCELLANEOUS COMPONENTS ESTIMATED AVERAGE INVENTORIES (CURIES)

(2560 MW)

WASTE CONDENSATE WASTE(1) WASTE WASTE NUCLIDE ION EXCHANGER CONCENTRATOR NUCLIDE CONDENSATE CONCENTRATOR N-16 0. 0. Y-91M 1.0E-09 2.2E-05 KR-83M 0. 0. Y-93 3.9E-09 6.9E-06 KR-85M 0. 0. ZR-95 2.2E-06 2.6E-05 KR-85 0. 0. NB-95 1.0E-06 2.1E-05 KR-87 0. 0. MO-99 1.1E-04 3.0E-02 KR-88 0. 0. TC-99M 2.8E-06 8.2E-03 KR-89 0. 0. RU-103 1.0E-06 2.0E-05 XE-131M 0. 0. RU-106 9.4E-07 4.5E-06 XE-133M 0. 0. RH-103M 1.6E-10 2.9E-09 XE-133 0. 0. RH-106 2.9E-15 6.2E-09 XE-135M 0. 0. TE-125M 9.6E-07 1.3E-05 XE-135 0. 0. TE-127M 1.5E-05 1.2E-04 XE-137 0. 0. TE-127 8.7E-08 1.7E-04 XE-138 0. 0. TE-129M 2.7E-05 6.0E-04 BR-83 7.7E-08 5.7E-04 TE-129 7.8E-09 1.2E-04 BR-84 3.0E-09 1.0E-04 TE-131M 1.2E-06 7.1E-04 BR-65 3.2E-12 1.1E-06 TE-131 7.8E-10 3.4E-05 I-130 3.1E-07 4.5E-04 TE-132 4.2E-05 9.7E-03 I-131 1.2E-03 1.1E-01 BA-137M 1.2E-10 5.2E-05 I-132 1.5E-06 1.2E-02 BA-140 1.6E-06 9.2E-05 I-133 1.1E-04 9.6E-02 IA-140 1.1E-07 4.8E-05 I-134 1.4E-07 2.9E-03 CE-141 1.3E-06 3.0E-05 I-135 1.3E-05 3.4E-02 CE-143 2.2E-08 1.2E-05 RB-86 9.4E-07 3.7E-05 CE-144 2.8E-06 1.4E-05 RB-88 7.1E-08 4.3E-03 PR-143 3.8E-07 2.1E-05 CS-134 2.8E-03 1.2E-02 PR-144 1.2E-11 7.2E-07 CS-136 1.0E-04 5.8E-03 NP-239 1.3E-06 4.0E-04 CS-137 2.3E-03 8.7E-03 CR-51 3.1E-06 8.2E-05 SR-89 1.0E-05 1.5E-04 MN-54 2.8E-06 1.4E-05 SR-90 1.2E-06 4.5E-06 FE-55 1.7E-05 7.0E-05 SR-91 7.0E-08 1.3E-04 FE-59 2.7E-06 4.4E-05 Y-90 1.5E-09 4.1E-07 CO-58 6.3E-05 7.0E-04 Y-91 2.1E-06 2.8E-05 CO-60 2.3E-05 9.0E-05

1) Waste concentrator is no longer used.

T12.2-13 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-13 SOLID WASTE MANAGEMENT SYSTEM COMPONENT MAXIMUM INVENTORIES (CURIES)

(2700 MW)

SPENT RESIN NUCLIDE TANK N-16 0.

KR-85M 0.

KR-85 0.

KR-87 0.

KR-88 0.

KR-131M 0.

XE-133 0.

XE-135 0.

XE-138 0.

BR-84 2.0E-01 RB-88 6.2E+00 RB-89 1.3E-01 SR-89 8.3E+01 SR-90 1.1E+01 Y-90 1.5E-01 SR-91 4.3E-01 Y-91 1.0E+02 ZR-95 1.4E+02 MO-99 3.7E+02 RU-103 7.4E+01 RU-106 7.7E+01 TE-129 2.0E-01 I-129 2.0E-03 I-131 9.1E+03 TE-132 3.6E+02 I-132 2.3E+01 I-133 1.3E+03 TE-134 3.1E-01 I-134 5.3E+00 CS-134 5.5E+03 I-135 2.0E+02 CS-136 2.7E+02 CS-137 1.7E+04 CS-138 3.8E+00 BA-140 3.3E+01 LA-140 4.2E+00 PR-143 2.9E+01 CE-144 1.9E+02 CR-51 2.1E+00 MN-54 1.9E+00 FE-55 1.2E+01 FE-59 1.8E+00 CO-58 4.3E+01 CO-60 1.6E+01 Note: E-xx denotes powers of 10 T12.2-14 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-14 SOLID WASTE MANAGEMENT SYSTEM COMPONENT AVERAGE INVENTORIES (CURIES)

(2560 MW)

SPENT SPENT RESIN RESIN NUCLIDE TANK NUCLIDE TANK N-16 0. Y-91M 3.4E-03 KR-83M 0. Y-93 4.5E-03 KR-85M 0. ZR-95 1.5E+00 KR-85 0. NB-95 6.9E-01 KR-87 0. MO-99 8.9E+01 KR-88 0. TC-99M 3.6E+00 KR-89 0. RU-103 7.1E-01 XE-131M 0. RU-106 6.3E-01 XE-133M 0. RH-103M 4.8E-04 XE-133 0. RH-106 9.2E-07 XE-135M 0. TE-125M 6.6E-01 XE-135 0. TE-127M 1.0E+01 XE-137 0. TE-127 1.1E-01 XE-138 0. TE-129H 1.9E+01 BR-83 1.5E-01 TE-129 2.3E-02 BR-84 1.7E-02 TE-131M 1.2E+00 BR-85 1.8E-04 TE-131 5.5E-03 I-130 3.8E-01 TE-132 3.6E+01 I-131 8.8E+U2 BA-137M 7.7E-03 I-132 2.9E+00 BA-140 1.1E+00 I-133 1.2E+02 LA-140 9.3E-02 I-134 5.1E-01 CE-141 9.1E-01 I-135 1.8E+01 CE-143 2.0E-02 RB 4.3E-01 CE-144 1.9E+00 RB 4.1E-01 PR-143 2.7E-01 CS 1.3E+03 PR-144 1.1E-04 CS 4.7E+01 NP-239 1.0E+00 CS 1.0E+03 CR-51 2.1E+00 SR 7.0E+00 MN-54 1.9E+00 SR 8.2E-01 FE-55 1.2E+01 SR 8.2E-02 FE-59 1.8E+00 Y 1.2E-03 CO-58 4.3E+01 Y 1.5E+00 CO-60 1.6E+00 Note* E-xx denotes powers of 10 T12.2-15 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-15 GASEOUS WASTE MANAGEMENT SYSTEM COMPONENT MAXIMUM INVENTORIES (CURIES)

(2700 MW)

GAS GAS DECAY GAS SAMPLE NUCLIDE TANK TANK ANALYZER VESSEL N-16 0. 0. 0. 0.

KR-85M 1.6E+00 1.6E+00 9.0E-01 2.2E-02 KR-85 4.6E+01 2.6E+03 2.1E+00 5.0E-02 K-87 3.2E-01 3.2E-01 6.2E-01 1.5E-02 KR-88 2.4E+00 2.4E+00 2.1E+00 5.0E-02 XE-131M 6.3E+01 3.6E+02 3.1E+00 7.5E-02 XE-133 3.7E+03 1.1E+04 2.0E+02 4.9E+00 XE-135 1.7E+01 1.7E+01 4.5E+00 1.1E-01 XE-138 3.4E-02 3.4E-02 3.5E-01 8.5E-03 BR-84 2.7E-07 2.7E-07 1.2E-06 3.0E-08 RB-88 2.0E-05 2.0E-05 1.7E-04 4.0E-06 RB-89 4.2E-07 4.2E-07 4.0E-06 9.7E-08 SR-89 5.3E-06 1.1E-04 2.4E-07 5.8E-09 SR-90 1.8E-07 1.0E-05 8.1E-09 1.9E-10 Y-90 1.2E-U7 2.1E-07 8.1E-09 1.9E-10 SR-91 6.2E-07 6.2E-07 1.6E-07 3.8E-09 Y-91 5.7E-06 1.3E-04 2.6E-07 6.2E-09 ZR-95 7.2E-06 1.8E-04 3.3E-07 7.9E-09 MO-99 3.0E-04 5.3E-04 2.0E-05 4.7E-07 RU-103 5.9E-06 1.0E-04 2.7E-07 6.5E-09 RU-106 1.6E-06 7.8E-05 7.2E-08 1.7E-09 TE-129 2.6E-07 2.6E-07 5.5E-07 1.3E-08 I-129 3.2E-11 1.8E-09 1.4E-12 3.5E-14 I-131 3.1E-03 1.2E-02 1.6E-04 3.8E-06 TE-132 2.4E-04 4.7E-04 1.5E-05 3.6E-07 I-132 3.1E-05 3.1E-05 3.3E-05 7.9E-07 I-133 1.5E-03 1.7E-03 2.0E-04 4.7E-06 TE-134 4.1E-07 4.1E-07 1.4E-06 3.4E-08 I-134 7.0E-06 7.0E-06 2.0E-05 4.7E-07 CS-134 2.3E-04 1.2E-02 1.0E-05 2.5E-07 I-135 2.7E-04 2.7E-04 9.8E-05 2.3E-06 CS-136 1.5E-04 9.03-04 7.1E-06 1.7E-07 CS-137 6.3E-04 3.6E-02 2.8E-05 6.8E-07 CS-138 1.3E-05 1.3E-05 5.7E-05 1.4E-06 BA-140 7.6E-06 4.7E-05 3.7E-07 9.0E-09 LA-140 4.4E-06 5.9E-06 3.6E-07 8.7E-09 PR-143 6.4E-06 4.1E-05 3.1E-07 7.5E-09 CE-144 4.3E-06 2.0E-04 2.0E-07 4.7E-09 CR-51 2.4E-06 3.0E-05 1.1E-07 2.7E-09 MN-54 4.1E-07 1.9E-05 1.8E-08 4.4E-10 FE-55 2.1E-06 1.1E-04 9.3E-08 2.2E-09 FE-59 1.3E-06 2.5E-05 5.9E-08 1.4E-09 CO-58 2.1E-05 5.5E-04 9.3E-07 2.2E-08 CO-60 2.7E-06 1.5E-04 1.2E-07 2.9E-09 Note: E-xx denotes powers of 10 T12.2-16 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-16 GASEOUS WASTE MANAGEMENT SYSTEM COMPONENT AVERAGE INVENTORIES (CURIES)

(2560 MW)

GAS GAS DECAY GAS SAMPLE NUCLIDE TANK TANK ANALYZER VESSEL N-16 0. 0. 0. 0.

KR-83M 5.2E-03 5.2E-03 6.8E-03 1.6E-04 KR-85M 6.9E-02 6.9E-02 3.8E-02 9.3E-04 KR-85 1.0E+00 5.8E+01 4.6E-02 1.1E-03 KR-87 9.8E-03 9.8E-03 1.9E-02 4.5E-04 KR-88 7.7E-02 7.7E-02 6.7E-02 1.6E-03 KR-89 2.0E-05 2.0E-05 9.4E-04 2.3E-05 XE-131M 6.1E-01 3.5E+00 3.0E-02 7.2E-04 XE-133M 8.7E-01 1.4E+00 6.2E-02 1.5E-03 XE-133 9.0E+01 2.6E+02 5.0E+00 1.2E-01 XE-135M 2.9E-04 2.9E-04 2.7E-03 6.5E-05 XE-135 3.6E-01 3.6E-01 9.6E-02 2.3E-03 XE-137 4.2E-05 4.2E-05 1.6E-03 3.9E-05 XE-138 8.5E-04 8.5E-04 8.8E-03 2.1E-04 BR-86 1.1E-07 1.1E-07 1.2E-07 2.8E-09 BR-84 1.3E-08 1.3E-08 5.8E-08 1.4E-09 BR-85 1.4E-10 1.4E-10 6.7E-09 1.6E-10 I-130 2.9E-07 2.9E-07 5.8E-08 1.4E-09 I-131 1.9E-04 7.6E-04 9.7E-06 2.3E-07 I-132 2.2E-06 2.2E-06 2.4E-06 5.8E-08 I-133 8.8E-05 9.5E-05 1.1E-05 2.7E-07 I-134 3.8E-07 3.8E-07 1.1E-06 2.6E-08 I-135 1.4E-05 1.4E-05 5.0E-06 1.2E-07 RB-86 1.0E-07 9.1E-07 5.0E-09 1.2E-10 RB-88 7.9E-07 7.9E-07 6.5E-06 1.6E-07 CS-134 3.5E-05 1.8E-03 1.5E-06 3.7E-08 CS-136 1.6E-05 1.0E-04 7.9E-07 1.9E-08 CS-137 2.5E-05 1.4E-03 1.1E-06 2.7E-08 SR-89 2.9E-07 6.1E-06 1.3E-08 3.2E-10 SR-90 8.6E-09 4.9E-07 3.9E-10 9.3E-12 SR-91 6.9E-08 6.9E-08 1.8E-08 4.2E-10 Y-90 5.9E-10 1.0E-09 3.9E-11 9.4E-13 Y-91 5.3E-08 1.2E-06 2.4E-09 5.8E-11 Y-91 2.9E-09 2.9E-09 8.3E-09 2.0E-10 Y-93 3.9E-09 3.9E-09 9.3E-10 2.2E-11 ZR-95 4.9E-08 1.2E-06 2.3E-09 5.4E-11 NB-95 4.0E-08 6.3E-07 1.9E-09 4.5E-11 MO-99 4.4E-05 7.8E-05 2.9E-06 6.9E-08 TC-99M 3.1E-06 3.1E-06 1.2E-06 3.0E-08 RU-103 3.7E-08 6.4E-07 1.7E-09 4.1E-11 RU-106 8.6E-09 4.2E-07 3.9E-10 9.3E-12 RH-103M 4.0E-10 4.0E-10 1.0E-09 2.5E-11 RH-106 7.5E-13 7.5E-13 2.2E-10 5.3E-12 TE-125M 2.4E-08 5.6E-07 1.1E-09 2.6E-11 TE-127M 2.4E-07 8.0E-06 1.1E-08 2.6E-10 TE-127 8.7E-08 8.8E-08 2.3E-08 5.6E-10 T12.2-17 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-16 (Cont'd)

GAS SURGE GAS DECAY GAS SAMPLE NUCLIDE TANK TANK ANALYZER VESSEL TE-129M 1.1E-06 1.7E-05 5.2E-08 1.3E-09 TE-129 1.7E-08 1.7E-08 3.7E-08 8.9E-10 TE131-M 7.9E-07 9.4E-07 7.7E-08 1.8E-09 TE-131 4.1E-09 4.1E-09 2.4E-08 5.9E-10 TE-132 1.5E-05 2.9E-05 9.2E-07 2.2E-08 BA-137M 6.2E-09 6.2E-09 3.6E-07 8.6E-09 BA-140 1.7E-07 1.0E-06 8.1E-09 2.0E-10 LA-140 6.0E-08 8.1E-08 4.9E-09 1.2E-10 CE-141 5.7E-08 8.2E-07 2.6E-09 6.3E-11 CE-143 1.4E-08 1.7E-08 1.3E-09 3.0E-11 CE-144 2.8E-08 1.3E-06 1.2E-09 3.0E-11 PR-143 3.8E-08 2.5E-07 1.8E-09 4.4E-11 PR-144 8.9E-11 8.9E-11 7.5E-10 1.8E-11 NP-239 5.6E-07 9.0E-07 3.9E-08 9.4E-10 CR-51 1.3E-06 1.7E-05 6.3E-08 1.5E-09 MN-54 2.3E-07 1.1E-05 1.0E-08 2.5E-10 FE-55 1.2E-06 6.4E-05 5.4E-08 1.3E-09 FE-59 7.4E-07 1.4E-05 3.4E-08 8.1E-10 CO-58 1.2E-05 3.1E-04 5.3E-07 1.3E-08 CO-60 1.5E-06 8.4E-05 6.8E-08 1.6E-09 Note: E-xx denotes powers of 10 T12.2-18 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-17 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENT MAXIMUM INVENTORIES (CURIES)

(2700 MW)

LETDOWN REGENERATIVE NUCLIDE HEAT EXCHANGER HEAT EXCHANGER N-16 7.6E-01 5.6E-01 KR-85M 1.9E-01 1.8E-01 KR-85 4.0E-01 4.1E-01 KR-87 1.5E-01 1.3E-01 KR-88 4.7E-01 4.4E-01 XE-131M 7.6E-01 7.8E-01 XE-133 5.0E+01 5.1E+01 XE-135 1.1E+00 1.1E+01 XE-138 8.9E-02 5.1E-02 RB-84 5.0E-03 9.1E-04 RB-88 4.7E-01 1.7E-01 RB-89 1.1E-02 4.0E-03 SR-89 9.9E-04 2.1E-04 SR-90 3.3E-05 7.0E-06 Y-90 3.3E-05 7.0E-06 SR-91 6.4E-04 1.3E-04 Y-91 1.1E-03 2.2E-04 ZR-95 1.3E-03 2.8E-04 MO-99 8.0E-02 1.7E-02 RU-103 1.1E-03 2.4E-04 RU-106 3.0E-04 6.3E-05 TE-129 2.3E-03 4.4E-04 I-129 5.9E-09 1.3E-09 I-131 6.4E-01 1.4E-01 TE-132 6.1E-02 1.3E-02 I-132 1.3E-01 2.7E-02 I-133 8.0E-01 1.7E-01 TE-134 5.7E-03 1.1E-03 I-134 8.0E-02 1.5E-02 CS-134 2.8E-02 1.6E-02 I-135 4.0E-01 8.3E-02 CS-136 1.9E-02 1.1E-02 CS-137 7.6E-02 4.4E-02 CS-138 1.6E-01 6.7E-02 BA-140 1.5E-03 3.2E-04 LA-140 1.5E-03 3.1E-04 PR-143 1.3E-03 2.7E-04 CE-144 8.0E-04 1.7E-04 CR-51 4.5E-04 5.9E-05 MN-54 7.5E-05 9.8E-06 FE-55 3.8E-04 5.0E-05 FE-59 2.4E-04 3.2E-05 CO-58 3.8E-03 5.0E-04 CO-60 4.9E-04 6.4E-05 Note: E-xx denotes powers of 10 T12.2-19 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-18 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

REGENERATIVE LETDOWN REGENERATIVE LETDOWN HEAT NUCLIDE HEAT EXCHANGER HEAT EXCHANGER NUCLIDE HEAT EXCHANGER EXCHANGER N-16 5.6E-03 9.1E-02 Y-91M 5.9E-05 1.1E-05 KR-83M 3.3E-03 2.7E-03 Y-93 6.4E-06 1.3E-06 KR-85M 1.76-02 1.6E-02 ZR-95 1.4E-05 3.0E-06 KR-85 1.9E-02 2.0E-02 NB-95 1.2E-05 2.5E-06 KR-87 9.6E-03 7.3E-03 MO-99 1.9E-02 4.0E-03 KR-88 3.1E-02 2.8E-02 TC-99M 8.7E-03 1.8E-03 KR-99 8.0E-04 1.7E-04 RU-103 1.1E-05 2.3E-06 XE-131M 1.7E-02 1.7E-02 RU-106 2.4E-06 5.2E-07 XE-133M 3.5E-02 3.5E-02 RH-103M 7.3E-06 1.3E-06 XE-133 2.8E+00 2.8E+00 RH-106 1.6E-06 2.0E-07 XE-135M 2.1E-03 8.9E-04 TE-125M 7.0E-06 1.5E-06 XE-135 5.6E-02 5.4E-02 TE-127M 6.8E-05 1.4E-05 XE-137 1.4E-03 3.2E-04 TE-127 1.6E-04 3.3E-05 XE-138 7.0E-03 2.8E-03 TE-129M 3.3E-04 7.0E-05 BR-83 8.2E-04 1.6E-04 TE-129 2.6E-04 4.8E-05 BR-84 4.2E-04 7.0E-05 TE-131M 5.2E-04 1.1E-04 BR-85 4.9E-05 6.3E-06 TE-131 1.7E-04 2.8E-05 I-130 4.0E-04 8.3E-05 TE-132 6.1E-03 1.3E-03 I-131 6.3E-02 1.3E-02 BA-137M 2.6E-03 3.4E-04 I-132 1.7E-02 3.3E-03 BA-140 5.2E-05 1.1E-05 I-133 7.6E-02 1.6E-02 LA-140 3.3E-05 7.0E-06 I-134 7.6E-03 1.4E-03 CE-141 1.7E-05 3.5E-06 I-135 3.5E-02 7.2E-03 CE-143 8.5E-06 1.8E-06 RB-86 2.1E-05 1.2E-05 CE-144 7.8E-06 1.7E-06 RB-88 3.1E-02 9.0E-03 PR-143 1.2E-05 2.5E-06 CS-134 6.4E-03 3.7E-03 PR-144 5.4E-06 8.3E-07 CS-136 3.3E-03 1.9E-03 NP-239 2.6E-04 5.5E-05 CS-137 4.7E-03 2.7E-03 CR-51 4.5E-04 5.9E-05 SR-89 8.3E-05 1.8E-05 MN-54 7.5E-05 5.8E-06 SR-90 2.4E-06 5.2E-07 FE-55 3.8E-04 5.0E-05 SR-91 1.2E-04 2.5E-05 FE-59 2.4E-04 3.2E-05 Y-90 2.6E-07 5.5E-08 CO-58 3.8E-03 5.0E-04 Y-91 1.5E-05 3.2E-06 CO-60 4.9E-04 6.4E-05 Note: E-xx denotes powers of 10 T12.2-20 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-19 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Boric Acid Purification Deborating Preconcentrator Condensate Nuclide Ion Exchanger Ion Exchanger Ion Exchanger Ion Exchanger N-16 2.1E-02 0. 0. 0.

KR-85M 1.2E+00 1.2E+00 1.7E-02 1.7E-05 KR-85 2.5E+00 2.5E+00 1.1E+00 1.1E-03 KR-87 9.6E-01 9.6E-01 4.2E-03 4.2E-06 KR-88 2.9E+00 2.9E+00 2.7E-02 2.7E-05 XE-131M 4.7E+00 4.7E+00 1.5E+00 1.5E-03 XE-133 3.1E+02 3.1E+02 6.8E+01 6.8E-02 XE-135 6.9E+00 6.9E+00 1.9E-01 1.9E-04 XE-138 5.5E-01 5.5E-01 4.6E-04 4.6E-07 BR-84 1.8E-01 1.8E-02 4.7E-06 4.2E-10 RB-88 6.2E+00 1.5E+00 4.8E-04 2.7E-08 RB-89 1.3E-01 3.5E-02 8.5E-06 5.6E-10 SR-89 8.2E+01 6.1E-04 5.0E-01 7.6E-08 SR-90 1.1E+01 2.0E-05 7.3E-02 2.8E-09 Y-90 1.5E-01 2.0E-05 3.0E-04 8.4E-10 SR-91 4.3E-01 4.0E-04 1.7E-04 3.2E-09 Y-91 1.0E+02 6.6E-04 6.2E-01 8.3E-08 ZR-95 1.4E+02 8.3E-04 8.7E-01 1.1E-07 MO-99 3.7E+02 5.0E-02 7.7E-01 2.1E-06 RU-103 7.3E+01 6.9E-04 4.4E-01 8.4E-08 RU-106 7.6E+01 1.8E-04 5.0E-01 2.5E-08 TE-129 1.8E-01 1.8E-02 9.7E-06 8.8E-10 I-129 2.0E-03 4.2E-06 1.3E-05 1.2E-09 I-131 8.6E+03 3.5E+02 3.5E+01 3.1E-03 TE-132 3.3E+02 2.4E+01 7.7E-01 7.0E-05 I-132 2.1E+01 2.1E+00 2.2E-03 2.0E-07 I-133 1.2E+03 1.1E+02 9.0E-01 8.1E-05 TE-134 2.8E-01 2.8E-02 9.7E-06 8.7E-10 I-134 4.8E+00 4.7E-01 2.0E-04 1.8E-08 CS-134 5.3E+03 8.6E-02 1.8E+02 6.9E-07 I-135 1.9E+02 1.8E+01 5.3E-02 4.8E-06 CS-136 2.7E+02 5.9E-02 6.5E+00 3.5E-07 CS-137 1.6E+04 2.4E-01 5.5E+02 1.9E-06 CS-138 3.8E+00 5.0E-01 5.4E-04 1.7E-08 BA-140 3.3E+01 9.5E-04 1.6E-01 9.3E-08 LA-140 4.1E+00 9.1E-04 5.7E-03 2.6E-08 PR-143 2.9E+01 7.9E-04 1.4E-01 7.8E-08 CE-144 1.9E+02 5.0E-04 1.3E+00 6.6E-08 CR-51 2.1E+00 3.0E-02 6.6E-03 1.8E-09 MN-54 1.9E+00 5.2E-03 6.8E-03 3.5E-10 FE-55 1.2E+01 2.7E-02 4.3E-02 1.8E-09 FE-59 1.8E+00 1.6E-02 6.1E-03 1.0E-09 CO-58 4.3E+01 2.6E-01 1.5E-01 1.7E-08 CO-60 1.6E+01 3.4E-02 5.8E-02 2.3E-09 Note: E-xx denotes powers of 10 T12.2-21 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-20 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED AVERAGE INVENTORIES (CURIES)

(2560 MW)

Boric Acid Purification Deborating Preconcentrator Condensate Nuclide Ion Exchanger Ion Exchanger Ion Exchanger Ion Exchanger N-16 3.6E-08 0. 0. 0.

KR-83M 2.0E-02 2.0E-02 1.2E-04 1.2E-07 KR-85M 1.1E-01 1.1E-01 1.4E-03 1.4E-06 KR-85 1.2E-01 1.2E-01 5.2E-02 5.2E-05 KR-87 5.9E-02 5.9E-02 2.3E-04 2.3E-07 KR-88 1.9E-01 1.9E-01 1.6E-03 1.6E-06 KR-89 5.0E-03 5.0E-03 8.1E-07 8.1E-10 XE-131M 1.0E-01 1.0E-01 3.0E-02 3.0E-05 XE-133M 2.2E-01 2.2E-01 2.7E-02 2.7E-05 XE-133 1.7E+01 1.7E+01 3.6E+00 3.6E-03 XE-135M 1.3E-02 1.3E-02 1.0E-05 1.0E-08 XE-135 3.4E-01 3.4E-01 9.2E-03 9.2E-06 XE-137 8.9E-03 8.9E-03 1.8E-06 1.8E-09 XE-138 4.3E-02 4.3E-02 3.2E-05 3.2E-08 BR-83 1.4E-01 1.3E-02 8.3E-06 7.4E-10 BR-84 1.5E-02 1.5E-03 2.0E-07 1.8E-11 BR-85 1.7E-04 1.7E-05 2.1E-10 1.9E-14 I-130 3.4E-01 3.4E-02 1.1E-04 1.0E-08 I-131 8.4E+02 3.4E+01 2.9E+00 2.6E-04 I-132 2.7E+00 2.6E-01 1.5E-04 1.4E-08 I-133 1.1E+02 1.1E+01 6.1E-02 5.5E-06 I-134 4.6E-01 4.5E-02 1.0E-05 9.1E-10 I-135 1.6E+01 1.6E+00 2.8E-03 2.5E-07 RB-86 4.2E-01 6.5E-05 1.1E-02 3.9E-10 RB-88 4.1E-01 9.7E-02 2.6E-05 1.5E-09 CS-134 1.2E+03 2.0E-02 4.1E+01 1.6E-07 CS-136 4.6E+01 1.0E-02 1.1E+00 5.6E-08 CS-137 1.0E+03 1.5E-02 3.4E+01 1.2E-07 SR-89 6.9E+00 5.2E-05 4.0E-02 6.2E-09 SR-90 8.1E-01 1.5E-06 5.4E-03 2.0E-10 SR-91 8.1E-02 7.5E-05 2.1E-05 3.9E-10 Y-90 1.2E-03 1.6E-07 1.9E-06 5.3E-12 Y-91 1.4E+00 9.5E-06 8.6E-03 1.1E-09 Y-91M 3.4E-03 3.7E-05 7.3E-08 1.6E-11 Y-93 4.5E-03 4.0E-06 1.2E-06 2.2E-11 ZR-95 1.5E+00 8.8E-06 8.9E-03 1.1E-09 NB-95 6.9E-01 7.3E-06 3.8E-03 8.3E-10 MO-99 8.8E+01 1.2E-02 1.5E-01 4.0E-07 TC-99M 3.6E+00 5.4E-03 5.6E-04 1.7E-08 RU-103 7.1E-01 6.7E-06 4.0E-03 7.7E-10 RU-106 6.3E-01 1.5E-06 4.1E-03 2.0E-10 RH-103m 4.8E-04 4.5E-06 1.1E-08 2.2E -12 RH-106 9.2E-07 9.9E-07 1.9E-13 4.2E-15 T12.2-22 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-20 (cont'd)

Boric Acid(1)

Purification Deborating Preconcentrator Condensate Nuclide Ion Exchanger Ion Exchanger Ion Exchanger Ion Exchanger TE-125M 6.5E-01 4.7E-03 3.9E-03 3.5E-07 TE-127M 1.0E+01 4.7E-02 6.5E-02 5.8E-06 TE-127 1.0E-01 1.0E-02 2.5E-05 2.3E-09 TE-129M 1.9E+01 2.2E-01 1.0E-01 9.3E-06 TE-129 2.1E-02 2.0E-03 6.0E-07 5.4E-11 TE-131M 1.1E+00 1.0E-01 8.7E-04 7.8E-08 TE-131 5.0E-03 4.9E-04 5.2E-08 4.7E-12 TE-132 3.3E+01 2.4E+00 6.2E-02 5.6E-06 BA-137M 7.7E-03 1.6E-03 8.2E-09 3.5E-11 BA-140 1.1E+00 3.2E-05 4.7E-03 2.8E-09 LA-140 9.2E-02 2.0E-05 9.8E-05 4.4E-10 CE-141 9.0E-01 1.0E-05 4.9E-03 1.2E-09 CE-143 1.9E-02 5.3E-06 1.7E-05 9.4E-11 CE-144 1.9E+00 4.8E-06 1.2E-02 6.4E-10 PR-143 2.7E-01 7.3E-06 1.2E-03 6.4E-10 PR-144 1.1E-04 3.3E-06 7.8E-10 4.9E-13 NP-239 1.0E+00 1.6E-04 1.5E-03 4.7E-09 CR-51 2.1E+00 3.0E-02 5.9E-03 1.6E-09 MN-54 1.9E+00 5.2E-03 6.7E-03 3.4E-10 FE-55 1.2E+01 2.7E-02 4.3E-02 1.8E-09 FE-59 1.8E+00 1.6E-02 5.6E-03 9.6E-10 CO-58 4.3E+01 2.6E-01 1.4E-01 1.6E-08 CO-60 1.6E+01 3.4E-02 5.8E-02 2.3E-09 (1) The boric acid condensate ion exchanger is no longer used.

T12.2-23 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-21 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Boric Acid Boric Acid Reactor Drain Volume Control Condensate Holding Nuclide Tank Tank Holdup Tank Tank(1) Tank(1)

N-16 0. 0. 0. 0. 0.

KR-85M 2.1E+01 2.1E+02 2.6E+02 3.7E-06 4.1E-04 KR-85 2.5E+02 4.8E+02 3.0E+03 2.7E-03 9.2E-01 KR-87 7.7E+00 1.4E+02 7.0E+01 2.6E-07 2.9E-05 KR-88 3.7E+01 4.9E+02 4.4E+02 3.8E-06 4.2E-04 XE-131M 2.4E+02 5.1E+02 2.5E+03 3.4E-03 9.7E-01 XE-133 1.5E+04 3.4E+04 1.5E+05 1.4E-01 3.4E+01 XE-135 1.2E+02 7.1E+02 1.2E+03 8.5E-05 9.5E-03 XE-138 2.3E+00 2.1E+01 3.5E+00 5.4E-09 6.0E-07 BR-84 1.1E-01 1.6E-02 2.4E-03 4.7E-11 5.2E-07 RB-88 1.1E+01 6.0E+00 4.0E-01 4.4E-09 4.9E-06 RB-89 2.6E-01 1.4E-01 8.3E-03 7.8E-11 8.6E-08 SR-89 2.9E-02 4.7E-03 1.1E-01 2.0E-06 6.7E-03 SR-90 9.8E-04 1.6E-04 4.1E-03 7.5E-08 2.6E-04 Y-90 9.1E-04 1.6E-04 1.2E-03 1.6E-08 3.0E-05 SR-91 1.6E-02 3.0E-03 4.8E-03 1.7E-08 1.9E-05 Y-91 3.1E-02 5.0E-03 1.2E-01 2.2E-06 7.3E-03 ZR-95 3.9E-02 6.4E-03 1.6E-01 2.8E-06 9.3E-03 MO-99 2.2E+00 3.8E-01 3.1E+00 4.0E-05 7.7E-02 RU-103 3.3E-02 5.3E-03 1.2E-01 2.2E-06 7.2E-03 RU-106 8.7E-03 1.4E-03 3.6E-02 6.6E-07 2.3E-03 TE-129 5.2E-02 8.5E-03 2.3E-03 9.7E-11 1.1E-06 I-129 1.8E-07 2.8E-08 7.4E-07 1.3E-12 4.6E-08 I-131 1.9E+01 3.1E+00 4.9E+01 7.8E-05 2.1E+00 TE-132 1.7E+00 2.9E-01 2.7E+00 3.6E-06 7.4E-02 I-132 3.1E+00 5.6E-01 2.6E-01 2.2E-08 2.5E-04 I-133 2.1E+01 3.7E+00 1.2E+01 8.2E-06 1.0E-01 TE-134 1.3E-01 1.9E-02 3.7E-03 9.7E-11 1.1E-06 I-134 1.8E+00 2.8E-01 6.2E-02 2.0E-09 2.2E-05 CS-134 8.2E-01 6.6E-01 1.0E+01 1.9E-05 6.4E-02 I-135 9.6E+00 1.8E+00 2.2E+00 5.3E-07 5.9E-03 CS-136 5.6E-01 4.5E-01 5.1E+00 8.7E-06 2.5E-02 CS-137 2.3E+00 1.8E+00 2.8E+01 5.2E-05 1.8E-01 CS-138 3.6E+00 2.5E+00 2.5E-01 4.9E-09 5.4E-06 BA-140 4.4E-02 7.3E-03 1.4E-01 2.3E-06 6.8E-03 LA-140 4.0E-02 7.0E-03 3.8E-02 4.0E-07 6.1E-04 PR-143 3.7E-02 6.0E-03 1.2E-01 2.0E-06 5.8E-03 CE-144 2.4E-02 3.8E-03 9.8E-02 1.8E-06 6.1E-03 CR-51 1.3E-02 2.1E-04 2.7E-02 4.7E-08 1.5E-04 MN-54 2.2E-03 3.6E-05 5.1E-03 9.4E-09 3.2E-05 FE-55 1.1E-02 1.8E-04 2.7E-02 4.8E-08 1.7E-04 FE-59 7.2E-03 1.2E-04 1.5E-02 2.7E-08 9.0E-05 CO-58 1.1E-01 1.8E-03 2.5E-01 4.5E-07 1.5E-03 CO-60 1.4E-02 2.3E-04 3.4E-02 6.2E-08 2.1E-04 (1) The boric acid condensate and holding tanks are no longer used.

T12.2-24 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-22 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

Boric Acid Boric Acid Reactor Drain Volume Control Condensate Holding Nuclide Tank Tank Holdup Tank Tank(1) Tank(1)

N-16 0. 0. 0. 0. 0.

KR-83M 3.2E-02 2.8E+00 1.7E-01 1.1E-08 1.2E-06 KR-85M 3.8E-01 1.7E+01 2.1E+00 3.1E-07 3.5E-05 KR-85 7.0E+00 2.1E+01 7.7E+01 1.3E-04 4.4E-02 KR-87 6.3E-02 7.4E+00 3.4E-01 1.5E-08 1.6E-06 KR-88 4.4E-01 2.9E+01 2.4E+00 2.3E-07 2.6E-05 KR-89 2.2E-04 8.2E-02 1.2E-03 2.2E-12 2.4E-10 XE-131M 2.7E+00 1.0E+01 2.1E+01 6.9E-05 2.0E-02 XE-133M 3.1E+00 2.1E+01 1.8E+01 4.3E-05 7.5E-03 XE-133 3.7E+02 1.7E+03 2.5E+03 7.4E-03 1.8E+00 XE-135M 1.6E-03 4.3E-01 7.3E-03 1.4E-10 1.5E-08 XE-135 1.3E+00 3.2E+01 6.4E+00 4.2E-06 4.7E-04 XE-137 2.7E-04 9.7E-02 1.2E-03 5.6E-12 6.2E-10 XE-138 4.9E-03 1.3E+00 2.2E-02 3.7E-10 4.1E-08 BR-83 5.0E-04 2.8E-03 5.0E-04 8.3E-11 9.2E-07 BR-84 5.7E-05 9.0E-04 5.6E-05 2.0E-12 2.3E-08 BR-85 6.4E-07 1.8E-05 6.1E-07 2.1E-15 2.3E-11 I-130 1.1E-03 1.6E-03 1.3E-03 1.1E-09 1.2E-05 I-131 8.4E-01 2.6E-01 2.2E+00 6.5E-06 1.7E-01 I-132 9.8E-03 5.8E-02 9.8E-03 1.5E-09 1.7E-05 I-133 3.2E-01 3.1E-01 4.3E-01 5.5E-07 6.8E-03 I-134 1.7E-03 2.0E-02 1.7E-03 1.0E-10 1.1E-06 I-135 5.6E-02 1.4E-01 6.1E-02 2.8E-08 3.1E-04 RB-86 3.3E-04 4.4E-04 3.1E-03 1.0E-08 3.1E-05 RB-88 2.4E-03 2.4E-01 1.2E-02 2.4E-10 2.6E-07 CS-134 1.2E-01 1.4E-01 1.3E+00 4.3E-06 1.5E-02 CS-136 4.9E-02 6.9E-02 4.5E-01 1.4E-06 4.1E-03 CS-137 8.6E-02 9.9E-02 9.3E-01 3.2E-06 1.1E-02 SR-89 1.4E-03 3.5E-04 4.9E-03 1.6E-07 5.4E-04 SR-90 4.5E-05 1.0E-05 1.6E-04 5.5E-09 1.9E-05 SR-91 2.7E-04 4.9E-04 3.1E-04 2.1E-09 2.3E-06 Y-90 2.3E-06 1.1E-06 4.2E-06 1.0E-10 1.9E-07 Y-91 2.7E-04 6.4E-05 9.2E-04 3.1E-08 1.0E-04 Y-91M 1.3E-05 1.6E-04 1.3E-05 7.3E-12 8.1E-09 Y-93 1.5E-05 2.6E-05 1.7E-05 1.2E-10 1.4E-07 ZR-95 2.5E-04 6.0E-05 8.6E-04 2.9E-08 9.6E-05 NB-95 2.0E-04 5.0E-05 6.6E-04 2.2E-08 7.0E-05 MO-99 1.7E-01 8.0E-02 3.2E-01 7.7E-06 1.5E-02 TC-99M 1.3E-02 3.4E-02 1.4E-02 5.6E-08 6.2E-05 RU-103 1.8E-04 4.5E-05 6.1E-04 2.0E-08 6.6E-05 RU-106 4.4E-05 1.0E-05 1.6E-04 5.4E-09 1.8E-05 RH-103M 1.8E-06 2.0E-05 1.8E-06 1.1E-12 1.3E-09 RH-106 3.5E-09 1.1E-07 3.4E-09 0. 2.1E-14 TE-125M 1.2E-04 2.9E-05 4.2E-04 1.4E-09 4.6E-05 No E-xx denotes powers of 10 T12.2-25 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-22 (Cont'd)

Boric Acid Boric Acid Reactor Drain Volume Control Condensate Holding Nuclide Tank Tank Holdup Tank Tank(1) Tank(1)

TE-129M 5.6E-03 1.4E-03 1.8E-02 6.0E-08 1.9E-03 TE-129 7.7E-05 7.6E-04 7.6E-05 6.0E-12 6.7E-08 TE-131M 2.9E-03 2.2E-03 4.2E-03 7.0E-09 9.6E-05 TE-131 1.9E-05 3.3E-04 1.8E-05 5.2E-13 5.8E-09 TE-132 5.9E-02 2.5E-02 1.2E-01 2.9E-07 5.9E-03 BA-137M 2.9E-05 8.5E-04 2.8E-05 8.2E-13 9.1E-10 BA-140 7.8E-04 2.2E-04 2.2E-03 7.0E-08 2.0E-04 LA-140 2.2E-04 1.4E-04 3.5E-04 6.9E-09 1.1E-05 CE-141 2.8E-04 7.0E-05 9.2E-04 3.0E-08 9.7E-05 CE-143 5.0E-05 3.5E-05 7.5E-05 1.3E-09 1.9E-06 CE-144 1.4E-04 3.3E-05 5.1E-04 1.7E-08 5.9E-05 PR-143 1.8E-04 5.0E-05 5.2E-04 1.6E-08 4.8E-05 PR-144 4.1E-07 8.2E-06 3.9E-07 7.8E-14 8.6E-11 NP-239 2.1E-03 1.1E-03 3.8E-03 8.6E-08 1.5E-04 CR-51 7.5E-03 1.9E-04 1.3E-02 4.2E-08 1.3E-04 MN-54 1.4E-03 3.1E-05 2.7E-03 9.2E-09 3.1E-05 FE-55 7.0E-03 1.6E-04 1.4E-02 4.8E-08 1.7E-04 FE-59 4.2E-03 1.0E-04 7.7E-03 2.6E-08 8.3E-05 CO-58 6.7E-02 1.6E-03 1.3E-01 4.3E-07 1.4E-03 CO-60 8.9E-03 2.0E-04 1.8E-02 6.2E-08 2.1E-04 (1) The boric acid condensate and holding tanks are no longer used.

T12.2-26 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-23 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Nuclide Letdown Filter Preconcentrator Filter N-16 6.5E-03 0.

KR-85M 2.5E-02 4.0E-04 KR-85 5.2E-02 2.6E-02 KR-87 2.0E-02 1.0E-04 KR-88 6.1E-02 6.5E-04 XE-131M 1.0E-01 3.6E-02 XE-133 6.6E+00 1.6E+00 XE-135 1.5E-01 4.5E-03 XE-138 1.2E-02 1.1E-05 BR-84 6.6E-04 3.8E-07 RB-88 6.1E-02 6.5E-05 RB-89 1.5E-03 1.3E-06 SR-89 1.3E-04 1.8E-05 SR-90 4.3E-06 6.7E-07 Y-90 4.3E-06 2.0E-07 SR-91 8.4E-05 7.8E-07 Y-91 1.4E-04 2.0E-05 ZR-95 1.7E-04 2.5E-05 MO-99 1.0E-02 5.0E-04 RU-103 1.5E-04 2.0E-05 RU-106 3.9E-05 5.9E-06 TE-129 3.0E-04 3.7E-07 I-129 7.7E-10 1.2E-10 I-131 8.4E-02 7.9E-03 TE-132 7.9E-03 4.3E-04 I-132 1.7E-02 4.3E-05 I-133 1.0E-01 1.9E-03 TE-134 7.5E-04 5.9E-07 I-134 1.0E-02 1.0E-05 CS-134 3.6E-03 1.6E-03 I-135 5.2E-02 3.5E-04 CS-136 2.5E-03 8.3E-04 CS-137 1.0E-02 4.6E-03 CS-138 2.1E-02 4.0E-05 BA-140 2.0E-04 2.2E-05 LA-140 1.9E-04 6.1E-06 PR-143 1.7E-04 1.9E-05 CE-144 1.0E-04 1.6E-05 CR-51 2.0E+01 6.6E-02 MN-54 1.8E+01 6.8E-02 FE-55 1.1E+02 4.3E-01 FE-59 1.8E+01 6.1E-02 CO-58 4.2E+02 1.5E+00 CO-60 1.5E+02 5.8E-01 Note: E-xx denotes powers of 10 T12.2-27 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-24 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

Nuclide Letdown Filter Preconcentrator Filter N-16 5.0E-08 0.

KR-83M 4.3E-04 2.8E-06 KR-85M 2.3E-03 3.4E-05 KR-85 2.5E-03 1.2E-03 KR-87 1.2E-03 5.5E-06 KR-88 4.1E-03 3.9E-05 KR-89 1.0E-04 1.9E-08 XE-131M 2.2E-03 7.3E-04 XE-133M 4.5E-03 6.4E-04 XE-133 3.6E-01 8.7E-02 XE-135M 2.7E-04 2.5E-07 XE-135 7.3E-03 2.2E-04 XE-137 1.9E-04 4.2E-08 XE-138 9.1E-04 7.6E-07 BR-83 1.1E-04 1.5E-07 BR-84 5.5E-05 1.7E-08 BR-85 6.4E-06 1.8E-10 I-130 5.2E-05 4.0E-07 I-131 8.2E-03 6.5E-04 I-132 2.2E-03 2.9E-06 I-133 1.0E-02 1.3E-04 I-134 1.0E-03 5.1E-07 I-135 4.5E-03 1.8E-05 RB-86 2.7E-06 9.4E-07 RB-88 4.1E-03 3.5E-06 CS-134 8.4E-04 3.8E-04 CS-136 4.3E-04 1.3E-04 CS-137 6.1E-04 2.8E-04 SR-89 1.1E-05 1.5E-06 SR-90 3.2E-07 4.9E-08 SR-91 1.6E-05 9.3E-08 Y-90 3.4E-08 1.3E-09 Y-91 2.0E-06 2.8E-07 Y-91M 7.7E-06 3.8E-09 Y-93 8.4E-07 5.2E-09 ZR-95 1.9E-06 2.6E-07 NB-95 1.5E-06 2.0E-07 MO-99 2.5E-03 9.6E-05 TC-99M 1.1E-03 4.1E-06 RU-103 1.4E-06 1.8E-07 RU-106 3.2E-07 4.8E-08 RH-103M 9.5E-07 5.3E-10 RH-106 2.1E-07 1.0E-12 TE-125M 9.1E-07 1.2E-07 TE-127M 8.9E-06 1.3E-06 TE-127 2.1-05 1.2E-07 Note: E-xx denotes powers of 10 T12.2-28 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-24 (Cont'd)

Nuclide Letdown Filter Preconcentrator Filter TE-129M 4.3E-05 5.5E-06 TE-129 3.4E-05 2.3E-08 TE-131M 6.8E-05 1.3E-06 TE-131 2.3E-05 5.5E-09 TE-132 7.9E-04 3.5E-05 BA-137M 3.4E-04 8.4E-09 BA-140 6.8E-06 6.7E-07 LA-140 4.3E-06 1.1E-07 CE-141 2.2E-06 2.8E-07 CE-143 1.1E-06 2.3E-08 CE-144 1.0E-06 1.5E-07 PR-143 1.5E-06 1.5E-07 PR-144 7.0E-07 1.2E-10 NP-239 3.4E-05 1.1E-06 CR-51 1.7E+01 5.9E-02 MN-54 5.7E+00 6.7E-02 FE-55 3.1E+01 4.3E-01 FE-59 1.2E+01 5.6E-02 CO-58 2.3E+02 1.4E+00 CO-60 40OE+01 5.8E-01 Note: E-xx denotes powers of 10 T12.2-29 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-25 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED MAXIMUM INVENTORIES (CURIES)

(2700 MW)

Nuclide Flash Tank Boric Acid Concentrator(1)

N-16 0. 0.

KR-85M 1.3E+00 4.7E-02 KR-85 2.8E+00 3.0E+00 KR-87 1.1E+00 1.2E-02 KR-88 3.3E+00 7.5E-02 XE-131M 5.4E+00 4.1E+00 XE-133 3.5E+02 1.9E+02 XE-135 7.7E+00 5.2E-01 XE-138 6.1E-01 1.3E-03 BR-84 4.8E-03 4.5E-05 RB-88 1.5E+00 7.5E-04 RB-89 3.5E-02 1.6E-05 SR-89 1.1E-03 2.1E-03 SR-90 3.6E-05 7.8E-05 Y-90 3.5E-05 2.3E-05 SR-91 6.4E-04 9.1E-05 Y-91 1.2E-03 2.3E-03 ZR-95 1.5E-03 2.9E-03 MO-99 8.4E-02 5.9E-02 RU-103 1.2E-03 2.3E-03 RU-106 3.2E-04 6.9E-04 TE-129 2.2E-03 4.3E-05 I-129 6.4E-09 1.4E-08 I-131 6.9E-01 9.2E-01 TE-132 6.4E-02 5.0E-02 I-132 1.3E-01 5.0E-03 I-133 8.1E-01 2.2E-01 TE-134 5.5E-03 6.9E-05 I-134 7.7E-02 1.2E-03 CS-134 9.0E-02 1.9E-02 I-135 3.9E-01 4.1E-02 CS-136 6.1E-02 9.7E-03 CS-137 2.5E-01 5.3E-02 CS-138 4.9E-01 4.6E-04 BA-140 1.6E-03 2.6E-03 LA-140 1.5E-03 7.2E-04 PR-143 1.4E-03 2.2E-03 CE-144 8.7E-04 1.8E-03 CR-51 2.7E-04 5.1E-05 MN-54 4.5E-05 9.7E-06 FE-55 2.3E-04 5.0E-05 FE-59 1.5E-04 2.9E-05 CO-58 2.3E-03 4.7E-04 CO-60 3.0E-04 6.4E-05 (1) The boric acid concentrator is no longer used.

T12.2-30 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-26 CHEMICAL AND VOLUME CONTROL SYSTEM COMPONENTS ESTIMATED AVERAGE INVENTORIES (CURIES)

(2560 MW)

Nuclide Flash Tank Boric Acid Concentrator(1)

N-16 0. 0.

KR-83M 2.2E-02 3.3E-04 KR-85M 1.1E-01 4.0E-03 KR-85 1.3E-01 1.5E-01 KR-87 6.2E-02 6.4E-04 KR-88 2.0E-01 4.6E-03 KR-89 5.2E-03 2.3E-06 XE-131M 1.1E-01 8.5E-02 XE-133M 2.3E-01 7.4E-02 XE-133 1.9E+01 1.0E+01 XE-135M 1.4E-02 2.9E-05 XE-135 3.7E-01 2.6E-02 XE-137 9.4E-03 4.9E-06 XE-138 4.5E-02 8.8E-05 BR-83 4.5E-04 1.8E-05 BR-84 2.2E-04 2.0E-06 BR-85 2.6E-05 2.1E-08 I-130 2.5E-04 4.6E-05 I-131 5.9E-02 7.6E-02 I-132 9.3E-03 3.4E-04 I-133 5.0E-02 1.5E-02 I-134 4.1E-03 5.9E-05 I-135 2.0E-02 2.1E-03 RB-86 6.6E-05 1.1E-05 RB-88 8.3E-02 4.1E-05 CS-134 2.1E-02 4.4E-03 CS-136 1.0E-02 1.6E-03 CS-137 1.5E-02 3.3E-03 SR-89 8.8E-05 1.7E-04 SR-90 2.6E-06 5.7E-06 SR-91 7.3E-05 1.1E-05 Y-90 2.1E-07 1.5E-07 Y-91 1.6E-05 3.2E-05 Y-91M 3.2E-05 4.4E-07 Y-93 3.9E-06 6.1E-07 ZR-95 1.5E-05 3.0E-05 NB-95 1.2E-05 2.3E-05 MO-99 1.5E-02 1.1E-02 TC-99M 5.0E-03 4.8E-04 RU-103 1.1E-05 2.1E-05 RU-106 2.6E-06 5.6E-06 RH103M 3.9E-06 6.1E-08 RH-106 8.5E-07 1.2E-10 TE-125M 7.4E-06 1.5E-05 TE-127M 7.3E-05 1.5E-04 TE-127 9.6E-05 1.4E-05 T12.2-31 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-26 (Cont'd)

Nuclide Flash Tank Boric Acid Concentrator(1)

TE-129M 3.4E-04 6.4E-04 TE-129 1.4E-04 2.7E-06 TE-131M 3.6E-04 1.5E-04 TE-131 9.3E-05 6.4E-07 TE-132 5.0E-03 4.1E-03 BA-137M 1.4E-03 9.8E-07 BA-140 5.1E-05 7.8E-05 LA-140 2.4E-05 1.2E-05 CE-141 1.7E-05 3.2E-05 CE-143 6.1E-06 2.6E-06 CE-144 8.5E-06 1.8E-05 PR-143 1.2E-05 1.8E-05 PR-144 2.9E-06 1.4E-08 NP-239 2.0E-04 1.3E-04 CR-51 2.5E-04 4.5E-05 MN-54 4.5E-05 9.6E-06 FE-55 2.3E-04 5.0E-05 FE-59 1.4E-04 2.7E-05 CO-58 2.2E-03 4.5E-04 CO-60 3.0E-04 6.4E-05 (1) The boric acid concentrator is no longer used T12.2-32 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-27 SAFETY INJECTION SYSTEM COMPONENT - MAXIMUM INVENTORIES (CURIES)

(2700 MW)

SHUTDOWN HEAT EXCHANGERS NUCLIDE SHUTDOWN LOCA N-16 0. 0 KR-85M 3.0E+00 1.8E+02 KR-85 6.4E+00 7.4E+00 KR-87 2.5E+00 3.2E+02 KR-88 7.5E+00 4.6E+02 XE-131M 1.2E+01 4.8E+00 XE-133 8.0E+02 1.5E+03 XE-135 1.8E+01 3.3E+02 XE-138 1.4E+00 1.2E+03 BR-84 8.0E-02 1.4E+04 RB-88 7.5E+00 9.4E+02 RB-89 1.8E-01 1.2E+03 SR-89 1.6E-02 1.3E+03 SR-90 5.3E-04 1.2E+02 Y-90 5.3E-04 1.2E+02 SR-91 1.0E-02 1.6E+03 Y-91 1.7E-02 1.7E+03 ZR-95 2.1E-02 2.5E+03 MO-99 1.3E+00 2.8E+03 RU-103 1.8E-02 2.6E+03 RU-106 4.7E-03 9.0E+02 TE-129 3.6E-02 5.0E+02 I-129 9.4E-08 2:6E-03 I-131 1.0E+01 7.7E+04 TE-132 9.7E-01 2.2E+03 I-132 2.1E+00 1.1E+05 I-133 1.3E+01 1.5E+05 TE-134 9.1E-02 2.2E+03 I-134 1.3E+00 1.6E+05 CS-134 4.4E-01 3.9E+02 I-135 6.4E+00 1.4E+05 CS-136 3.0E-01 1.0E+02 CS-137 1.2E+00 1.7E+02 CS-138 2.5E+00 2.6E+03 BA-140 2.4E-02 2.6E+03 LA-140 2.3E-02 2.7E+03 PR-143 2.0E-02 2.2E+03 CE-144 1.3E-02 1.9E+03 CR-51 7.2E-03 1.1E-05 MN-54 1.2E-03 1.9E-06 FE-55 6.1E-33 9.5E-06 FE-59 3.9E-03 6.0E-06 CO-58 6.1E-02 9.5E-05 CO-60 7.7E-03 1.2E-05 NOTE: E-xx denotes powers of 10 T12.2-33 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-28 SAFETY INJECTION SYSTEM COMPONENTS - AVERAGE INVENTORIES (CURIES)

(2560 MW)

SHUTDOWN SHUTDOWN NUCLIDE HEAT EXCHANGER NUCLIDE HEAT EXCHANGER N-16 0. Y-91M 9.4E-04 KR-83M 5.3E-02 Y-93 1.0E-04 KR-85M 2.8E-01 ZR-95 2.3E-04 KR-85 3.0E-01 NB-95 1.9E-04 KR-87 1.5E-01 MO-99 3.0E-01 KR-88 5.0E-01 TC-99M 1.4E-01 KR-89 1.3E-02 RU-103 1.7E-04 XE-131M 2.6E-01 RU-106 3.9E-05 XE-133M 5.5E-01 RH-103M 1.2E-04 XE-133 4.4E+01 RH-106 2.5E-05 XE-135M 3.3E-02 TE-125M 1.1E-04 XE-135 8.8E-01 TE-127M 1.1E-03 XE-137 2.3E-02 TE-127 2.5E-03 XE-138 1.1E-01 TE-129M 5.3E-03 BR-83 1.3E-02 TE-129 4.1E-03 BR-84 6.6E-03 TE-131M 8.3E-03 BR-85 7.7E-04 TE-131 2.8E-03 I-130 6.4E-03 TE-132 9.7E-02 I-131 9.9E-01 BA-137M 4.1E-02 I-132 2.7E-01 BA-140 8.3E-04 I-133 1.2E+00 LA-140 5.3E-04 I-134 1.2E-01 CE-141 2.7E-04 I-135 5.5E-01 CE-143 1.4E-04 RB-86 3.3E-04 CE-144 1.2E-04 RB-88 5.0E-01 PR-143 1.9E-04 CS-134 1.0E-01 PR-144 8.6E-05 CS-136 5.3E-02 NP-239 4.1E-03 CS-137 7.5E-02 CR-51 7.2E-03 SR-89 1.3E-03 MN-54 1.2E-03 SR-90 3.9E-05 FE-55 6.1E-03 SR-91 1.9E-03 FE-59 3.9E-03 Y-90 4.1E-06 CO-58 6.1E-02 Y-91 2.4E-04 C0-60 7.7E-03 NOTE: E-xx denotes powers of 10 T12.2-34 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-29 SAMPLING SYSTEM COMPONENTS - MAXIMUM INVENTORIES (CURIES)

(2700 MW)

HOT LEG PRESSURIZER PRESSURIZER SHUTDOWN LOOP 2A SURGE LINE STEAM SPACE COOLING 2A COOLING 2A NUCLIDE HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER N-16 6.0E-05 4.0E-05 0. 2.5E-05 KR-85M 2.4E-04 2.4E-04 2.4E-04 2.4E-04 KR-85 5.0E-04 5.0E-04 5.0E-04 5.0E-04 KR-87 1.9E-04 1.9E-04 1.9E-04 1.9E-04 KR-88 5.8E-04 5.8E-04 5.9E-04 5.8E-04 XE-131M 9.5E-04 9.5E-04 4.8E-03 9.5E-04 XE-133 6.3E-02 6.3E-02 3.2E-01 6.3E-02 XE-135 1.4E-03 1.4E-03 7.0E-03 1.4E-03 XE-138 1.1E-04 1.1E-04 5.6E-04 1.1E-04 BR-84 6.3E-06 6.3E-06 1.1E-09 6.3E-06 RB-88 5.8E-04 5.8E-04 1.0E-07 5.8E-04 RB-89 1.4E-05 1.4E-05 2.5E-09 1.4E-05 SR-89 1.2E-06 1.2E-06 2.2E-10 1.2E-06 SR-90 4.1E-08 4.1E-08 7.2E-12 4.1E-08 Y-90 4.1E-08 4.1E-08 7.2E-12 4.1E-08 SR-91 8.0E-07 8.0E-07 1.4E-10 8.0E-07 Y-91 1.3E-06 1.3E-06 2.3E-10 1.3E-06 ZR-95 1.7E-06 1.7E-06 2.9E-10 1.7E-06 MO-99 9.9E-05 9.9E-05 1.7E-08 9.9E-05 RU-103 1.4E-06 1.4E-06 2.4E-10 1.4E-06 RU-106 3.7E-07 3.7E-07 6.4E-11 3.7E-07 TE-129 2.8E-06 2.8E-06 4.9E-10 2.8E-06 I-129 7.3E-12 7.3E-12 1.3E-15 7.3E-12 I-131 8.0E-04 8.0E-04 1.4E-07 8.0E-04 TE-132 7.6E-05 7.6E-05 1.3E-08 7.6E-05 I-132 1.7E-04 1.7E-04 2.9E-08 1.7E-04 I-133 9.9E-04 9.9E-04 1.7E-07 9.9E-04 TE-134 7.1E-06 7.1E-06 1.2E-09 7.1E-06 I-134 9.9E-05 9.9E-05 1.7E-08 9.9E-05 GS-134 3.5E-05 3.5E-05 6.1E-09 3.5E-05 I-135 5.0E-04 5.0E-04 8.7E-08 5.0E-04 CS-136 2.4E-05 2.4E-05 4.2E-09 2.4E-05 CS-137 9.5E-05 9.5E-05 1.7E-08 9.5E-05 CS-138 2.0E-04 2.0E-04 3.5E-08 2.0E-04 BA-140 1.9E-06 1.9E-06 3.3E-10 1.9E-06 LA-140 1.8E-06 1.8E-06 3.2E-10 1.8E-06 PR-143 1.6E-06 1.6E-06 2.8E-10 1.6E-06 CE-144 9.9E-07 9.9E-07 1.7E-10 9.9E-07 CR-51 5.6E-07 5.6E-07 9.8E-11 5.6E-07 MN-54 9.3E-08 9.3E-08 1.6E-11 9.3E-08 FE-55 4.7E-07 4.7E-07 8.3E-11 4.7E-07 FE-59 3.2E-07 3.0E-07 5.3E-11 3.0E-07 CO-58 4.7E-06 4.7E-06 8.3E-10 4.7E-06 CO-60 6.0E-07 6.0E-07 1.1E-10 6.0E-07 NOTE: E-xx denotes powers of 10 T12.2-35 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-30 SAMPLING SYSTEM COMPONENTS - AVERAGE INVENTORIES (CURIES)

(2560 MW)

HOT LEG PRESSURIZER PRESSURIZER SHUTDOWN LOOP 2A SURGE LINE STEAM SPACE COOLING 2A NUCLIDE HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER N-16 5.7E-05 3.8E-05 0. 2.4E-05 KR-83M 4.1E-06 4.1E-06 4.1E-06 4.1E-06 KR-85M 2.2E-05 2.2E-05 2.2E-05 2.2E-05 KR-85 2.4E-05 2.4E-05 2.4E-05 2.4E-05 KR-87 1.2E-05 1.2E-05 1.2E-05 1.2E-05 KR-88 3.9E-05 3.9E-05 3.9E-05 3.9E-05 KR-89 9.9E-07 9.9E-07 1.0E-06 9.9E-07 XE-131M 2.0E-05 2.0E-05 1.0E-04 2.0E-05 XE-133M 4.3E-05 4.3E-05 2.2E-04 4.3E-05 XE-133 3.5E-03 3.5E-03 1.7E-02 3.5E-03 XE-135M 2.6E-06 2.6E-06 1.3E-05 2.6E-06 XE-135 6.9E-05 6.9E-05 3.5E-04 6.9E-05 XE-137 1.8E-06 1.8E-06 9.1E-06 1.8E-06 XE-138 8.6E-06 8.6E-06 4.4E-05 8.6E-06 BR-83 1.0E-06 1.0E-06 1.8E-10 1.0E-06 BR-84 5.2E-07 5.2E-07 9.1E-11 5.2E-07 BR-85 6.0E-08 6.0E-08 1.1E-11 6.0E-08 I-130 5.0E-07 5.0E-07 8.7E-11 5.0E-07 I-131 7.8E-05 7.8E-05 1.4E-08 7.8E-05 I-132 2.1E-05 2.1E-05 3.7E-09 2.1E-05 I-133 9.5E-05 9.5E-05 1.7E-08 9.5E-05 I-134 9.5E-06 9.5E-06 1.7E-09 9.5E-06 I-135 4.3E-05 4.3E-05 7.6E-09 4.3E-05 RB-86 2.6E-08 2.6E-08 4.5E-12 2.6E-08 RB-88 3.9E-05 3.9E-05 6.8E-09 3.9E-05 CS-134 8.0E-06 8.0E-06 1.4E-09 8.0E-06 GS-136 4.1E-06 4.1E-06 7.2E-10 4.1E-06 CS-137 5.8E-06 5.8E-06 1.0E-09 5.8E-06 SR-89 1.0E-07 1.0E-07 1.8E-11 1.0E-07 SR-90 3.0E-09 3.0E-09 5.3E-13 3.0E-09 SR-91 1.5E-07 1.5E-07 2.6E-11 1.5E-07 Y-90 3.2E-10 3.2E-10 5.7E-14 3.2E-10 Y-91 1.9E-08 1.9E-08 3.3E-12 1.9E-08 Y-91M 7.3E-08 7.3E-08 1.3E-11 7.3E-08 Y-93 8.0E-09 8.0E-09 1.4E-12 8.0E-09 ZR-95 1.8E-08 1.8E-08 3.1E-12 1.8E-08 NB-95 1.5E-08 1.5E-08 2.6E-12 1.5E-08 MO-99 2.4E-05 2.4E-05 4.2E-09 2.4E-05 TC-99M 1.1E-05 1.1E-05 1.9E-09 1.1E-05 RU-103 1.3E-08 1.3E-08 2.3E-12 1.3E-08 RU-106 3.0E-09 3.0E-09 5.3E-13 3.0E-09 RH-103M 9.1E-09 9.1E-09 1.6E-12 9.1E-09 RH-106 2.0E-09 2.0E-09 3.5E-13 2.0E-09 NOTE: E-xx denotes powers of 10 T12.2-36 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-30 (Cont'd)

HOT LEG PRESSURIZER PRESSURIZER SHUTDOWN LOOP 2A SURGE LINE STEAM SPACE COOLING 2A NUCLIDE HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER HEAT EXCHANGER TE-125M 8.6E-09 8.6E-09 1.5E-12 8.6E-09 TE-127M 8.4E-08 8.4E-08 1.5E-11 8.4E-08 TE-127 2.0E-07 2.0E-07 3.5E-11 2.0E-07 TE-129M 4.1E-07 4.1E-07 7.2E-11 4.1E-07 TE-129 3.2E-07 3.2E-07 5.7E-11 3.2E-07 TE-131M 6.5E-07 6.5E-07 1.1E-10 6.5E-07 TE-131 2.2E-07 2.2E-07 3.8E-11 2.2E-07 TE-132 7.6E-06 7.6E-06 1.3E-09 7.6E-06 BA-137M 3.2E-06 3.2E-06 5.7E-10 3.2E-06 BA-140 6.5E-08 6.5E-08 1.1E-11 6.5E-08 LA-140 4.1E-08 4.1E-08 7.2E-12 4.1E-08 CE-141 2.1E-08 2.1E-08 3.6E-12 2.1E-08 CE-143 1.1E-08 1.1E-08 1.9E-12 1.1E-08 CE-144 9.7E-09 9.7E-09 1.7E-12 9.7E-09 PR-143 1.5E-08 1.5E-08 2.6E-12 1.5E-08 PR-144 6.7E-09 6.7E-09 1.2E-12 6.7E-09 NP-239 3.2E-07 3.2E-07 5.7E-11 3.2E-07 CR-51 5.6E-07 5.6E-07 9.8E-11 5.6E-07 MN-54 9.3E-08 9.3E-08 1.6E-11 9.3E-08 FE-55 4.7E-07 4.7E-07 8.3E-11 4.7E-07 FE-59 3.0E-07 3.0E-07 5.3E-11 3.0E-07 CO-58 4.7E-06 4.7E-06 8.3E-10 4.7E-06 CO-60 6.0E-07 6.0E-07 1.1E-10 6.0E-07 NOTE: E xx denotes powers of 10 T12.2-37 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-31 SAMPLING SYSTEM COMPONENTS - MAXIMUM INVENTORIES(CURIES)

(2700 MW)

HOT LEG PRESSURIZER LOOP 2A STEAM SPACE NUCLIDE SAMPLE VESSEL SAMPLE VESSEL N-16 1.4E-04 0.

KR-85M 1.1E-03 6.5E-04 KR-85 2.4E-03 1.4E-03 KR-87 9.1E-04 5.2E-04 KR-88 2.8E-03 1.6E-03 XE-131M 4.5E-03 1.3E-02 XE-133 3.0E-01 8.6E-01 XE-135 6.5E-03 1.9E-02 XE-138 5.2E-04 1.5E-03 BR-84 3.0E-05 3.0E-09 RB-88 2.8E-03 2.8E-07 RB-89 6.6E-05 6.6E-09 SR-89 5.8E-06 5.8E-10 SR-90 1.9E-07 1.9E-11 Y-90 1.9E-07 1.9E-11 SR-91 3.8E-06 3.8E-10 Y-91 6.2E-06 6.2E-10 ZR-95 7.9E-06 7.9E-10 MO-99 4.7E-04 4.7E-08 RU-103 6.5E-06 6.5E-10 RU-106 1.7E-06 1.7E-10 TE-129 1.3E-05 1.3E-09 I-129 3.5E-11 3.5E-15 I-131 3.8E-03 3.8E-07 TE-132 3.6E-04 3.6E-08 I-132 7.9E-04 7.9E-08 I-133 4.7E-03 4.7E-07 TE-134 3.4E-05 3.4E-09 I-134 4.7E-04 4.7E-08 CS-134 1.6E-04 1.6E-08 I-135 2.4E-03 2.4E-07 CS-136 1.1E-04 1.1E-08 CS-137 4.5E-04 4.5E-08 CS-138 9.4E-04 9.4E-08 BA-140 9.0E-06 9.0E-10 LA-140 8.7E-06 8.7E-10 PR-143 7.5E-06 7.5E-10 CE-144 4.7E-06 4.7E-10 CR-51 2.7E-06 2.7E-10 MN-54 4.4E-07 4.4E-11 FE-55 2.2E-06 2.2E-10 FE-59 1.4E-06 1.4E-10 CO-58 2.2E-05 2.2E-09 CO-60 2.9E-06 2.9E-10 NOTE: E xx denotes powers of 10 T12.2-38 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-32 SAMPLING SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

HOT LEG PRESSURIZER HOT LEG PRESSURIZER NUCLIDE LOOP 2A STEAM SPACE LOOP 2A STEAM SPACE VESSEL SAMPLE VESSEL SAMPLE VESSEL VNUCLIDE SAMPLE VESSEL SAMPLE N-16 1.4E-04 0. Y-91M 3.5E-07 3.5E-11 KR-83M 1.9E-05 1.1E-05 Y-93 3.8E-08 3.8E-12 KR-85M 1.0E-04 5.9E-05 ZR-95 8.4E-08 8.4E-12 KR-85 1.1E-04 6.5E-05 NB-95 7.0E-08 6.9E-12 KR-87 5.6E-05 3.2E-05 MO-99 1.1E-04 1.1E-08 KR-88 1.8E-04 1.1E-04 TC-99M 5.1E-05 5.1E-09 KR-89 4.7E-06 2.7E-06 RU-103 6.3E-08 6.3E-12 XE-131M 9.7E-05 2.8E-04 RU-106 1.4E-08 1.4E-12 XE-133M 2.0E-04 5.9E-04 RH-103M 4.3E-08 4.3E-12 XE-133 1.6E-02 4.7E-02 RH-106 9.4E-09 9.4E-13 XE-135M 1.2E-05 3.5E-05 TE-125M 4.1E-08 4.1E-12 XE-135 3.3E-04 9.4E-04 TE-127M 4.0E-07 4.0E-11 XE-137 8.5E-06 2.4E-05 TE-127 9.4E-07 9.4E-11 XE-138 4.1E-05 1.2E-04 TE-129M 1.9E-06 1.9E-10 BR-83 4.8E-06 4.8E-10 TE-129 1.5E-06 1.5E-10 BR-84 2.5E-06 2.5E-10 TE-131M 3.1E-06 3.1E-10 BR-85 2.9E-07 2.9E-11 TE-131 1.0E-06 1.0E-10 I-130 2.4E-06 2.4E-10 TE-132 3.6E-05 3.6E-09 I-131 3.7E-04 3.7E-08 BA-137M 1.5E-05 1.5E-09 I-132 9.9E-05 9.9E-09 BA-140 3.1E-07 3.1E-11 I-133 4.5E-04 4.5E-08 LA-140 1.9E-07 1.9E-11 I-134 4.5E-05 4.5E-09 CE-141 9.8E-08 9.8E-12 I-135 2.0E-04 2.0E-08 CE-143 5.0E-08 5.0E-12 RB-86 1.2E-07 1.2E-11 CE-144 4.6E-08 4.6E-12 RB-88 1.8E-04 1.8E-08 PR-143 7.0E-08 6.9E-12 CS-134 3.8E-05 3.8E-09 PR-144 3.2E-08 3.2E-12 CS-136 1.9E-05 1.9E-09 NP-239 1.5E-06 1.5E-10 CS-137 2.8E-05 2.8E-09 CR-51 2.7E-06 2.7E-10 SR-89 4.9E-07 4.9E-11 MN-54 4.4E-07 4.4E-11 SR-90 1.4E-08 1.4E-12 FE-55 2.2E-06 2.2E-10 SR-91 7.2E-07 7.2E-11 FE-59 1.4E-06 1.4E-10 Y-90 1.5E-09 1.5E-13 CO-58 2.2E-05 2.2E-09 Y-91 9.0E-08 9.0E-12 CO-60 2.9E-06 2.9E-10 NOTE: E-xx denotes powers of 10 T12.2-39 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-33 FUEL POOL SYSTEM COMPONENTS MAXIMUM INVENTORIES (CURIES)

(2700 MW)

FUEL POOL FUEL POOL FUEL POOL ION PURIFICATION HEAT NUCLIDE EXCHANGER FILTER EXCHANGER N-16 0. 0. 0.

KR-85M 1.4E-05 3.0E-07 3.3E-05 KR-85 6.2E-02 1.3E-03 1.5E-01 KR-87 6.8E-14 1.5E-15 1.6E-13 KR-88 4.2E-07 8.9E-09 9.9E-07 XE-131M 1.1E-01 2.2E-03 2.5E-01 XE-133 6.0E+00 1.3E-01 1.4E+01 XE-135 4.4E-03 9.3E-05 1.0E-02 XE-138 0. 0. 0.

BR-84 0. 0. 0.

RB-88 0. 0. 0.

RB-89 0. 0. 0.

SR-89 1.7E-01 3.2E-06 3.5E-04 SR-90 6.4E-03 1.1E-07 1.2E-05 Y-90 1.5E-03 6.4E-08 7.2E-06 SR-91 3.9E-04 6.5E-08 7.2E-06 Y-91 1.8E-01 3.4E-06 3.8E-04 ZR-95 2.3E-01 4.3E-06 4.8E-04 MO-99 3.8E+00 1.6E-04 1.8E-02 RU-103 1.8E-01 3.5E-06 3.9E-04 RU-106 5.6E-02 9.7E-07 1.1E-04 TE-129 1.2E-15 0. 0.

I-129 1.1E-06 2.0E-11 2.2E-09 I-131 6.6E+01 1.8E-03 2.0E-01 TE-132 3.4E+00 1.3E-04 1.4E-02 I-132 2.8E-07 1.7E-10 1.9E-08 I-133 5.9E+00 5.2E-04 5.8E-02 TE-134 0. 0. 0.

I-134 0. 0. 0.

CS-134 5.3E+00 9.2E-05 1.0E-02 I-135 3.8E-02 8.7E-06 9.6E-04 CS-136 2.4E+00 5.7E-05 6.3E-03 CS-137 1.5E+01 2.5E-04 2.8E-02 CS-138 0. 0. 0.

BA-140 1.9E-01 4.5E-06 5.0E-04 LA-140 3.8E-02 2.1E-06 2.3E-04 PR-143 1.6E-01 3.8E-06 4.2E-04 CE-144 1.5E-01 2.6E-06 2.9E-04 CR-51 6.9E-03 6.9E-02 1.6E-04 MN-54 1.4E-03 1.4E-02 2.7E-05 FE-55 7.3E-03 7.3E-02 1.4E-04 FE-59 4.0E-03 4.0E-02 8.6E-05 CO-58 6.6E-02 6.6E-01 1.4E-03 CO-60 9.4E-03 9.4E-02 1.8E-04 NOTE: E xx denotes powers of 10 T12.2-40 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-34 FUEL POOL SYSTEM COMPONENTS AVERAGE INVENTORIES (CURIES)

(2560 MW)

FUEL POOL FUEL POOL FUEL POOL ION PURIFICATION HEAT NUCLIDE EXCHANGER FILTER EXCHANGER N-16 0. 0. 0.

KR-83M 7.0E-12 1.5E-13 1.7E-11 KR-85M 1.3E-06 2.7E-08 3.0E-06 KR-85 3.0E-03 6.3E-05 7.0E-03 KR-87 4.2E-15 0. 1.0E-14 KR-88 2.8E-08 5.9E-10 6.6E-08 KR-89 0. 0. 0.

XE-131M 2.3E-03 4.9E-05 5.4E-03 XE-133M 2.9E-03 6.2E-05 6.8E-03 XE-133 3.3E-01 7.0E-03 7.8E-01 XE-135M 0. 0. 0.

XE-135 2.2E-04 4.7E-06 5.2E-04 XE-137 0. 0. 0.

XE-138 0. 0. 0.

BR-83 3.6E-09 2.2E-12 2.4E-10 BR-84 0. 0. 0.

BR-85 0. 0. 0.

I-130 6.5E-04 8.7E-08 9.7E-06 I-131 6.4E+00 1.7E-04 1.9E-02 I-132 3.5E-08 2.2E-11 2.4E-09 I-133 5.6E-01 5.0E-05 5.6E-03 I-134 0. 0. 0.

I-135 3.3E-03 7.5E-07 8.4E-05 RB-86 2.9E-03 6.4E-08 7.1E-06 RB-88 0. 0. 0.

CS-134 1.2E+00 2.1E-05 2.4E-03 CS-136 4.1E-01 9.8E-06 1.1E-03 CS-137 9.1E-01 1.6E-05 1.7E-03 SR-89 1.4E-02 2.7E-07 3.0E-05 SR-90 4.7E-04 8.0E-09 8.9E-07 SR-91 7.4E-05 1.2E-08 1.4E-06 Y-90 1.2E-05 5.1E-10 5.6E-08 Y-91 2.6E-03 4.9E-08 5.5E-06 Y-91M 0. 0. 0.

Y-93 4.9E-06 7.8E-10 8.7E-08 ZR-95 2.5E-03 4.6E-08 5.1E-06 NB-95 1.9E-03 3.8E-08 4.2E-06 MO-99 9.1E-01 3.8E-05 4.2E-03 TC-99M 4.1E-04 1.1E-07 1.2E-05 RU-103 1.7E-03 3.4E-08 3.8E-06 RU-106 4.6E-04 8.0E-09 8.9E-07 RH-103M 0. 0. 0.

RH-106 0. 0. 0.

NOTE: E-xx denotes powers of 10 T12.2-41 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-34 (Cont'd)

FUEL POOL FUEL POOL FUEL POOL ION PURIFICATION HEAT NUCLIDE EXCHANGER FILTER EXCHANGER TE-125M 1.2E-03 2.2E-08 2.5E-06 TE-127M 1.2E-02 2.2E-07 2.5E-05 TE-127 8.2E-05 1.4E-08 1.6E-06 TE-129M 5.2E-02 1.0E-06 1.2E-04 TE-129 0. 0. 0.

TE-131M 8.2E-03 5.6E-07 6.2E-05 TE-131 0. 0. 0.

TE-132 3.4E-01 1.3E-05 1.4E-03 BA-137M 0. 0. 0.

BA-140 6.5E-03 1.5E-07 1.7E-05 LA-140 8.4E-04 4.7E-08 5.2E-06 CE-141 2.6E-03 5.3E-08 5.9E-06 CE-143 1.6E-04 1.0E-08 1.1E-06 CE-144 1.5E-03 2.6E-08 2.9E-06 PR-143 1.5E-03 3.5E-08 3.9E-06 PR-144 0. 0. 0.

NP-239 1.0E-02 4.7E-07 5.3E-05 CR-51 6.9E-03 6.9E-02 1.6E-04 MN-54 1.4E-03 1.4E-02 2.7E-05 FE-55 7.3E-03 7.3E-02 1.4E-04 FE-59 4.0E-03 4.0E-02 8.6E-05 CO-58 6.6E-02 6.6E-01 1.4E-03 CO-60 9.4E-03 9.4E-02 1.8E-04 NOTE: E-xx denotes powers of 10 T12.2-42 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-35 SPENT FUEL GAMMA SOURCE (MEV/WATT-S)

Time after Shutdown Mean Energy (Mev) 50 hr. 200 hr. 500 hr. 1000 hr.

0.30 3.24(+09)* 9.92(+08) 3.38(+08) 1.76(+08) 0.63 9.02(+09) 5.96(+09) 4.46(+09) 3.46(+09) 1.10 1.30(+09) 5.32(+08) 2.58(+08) 1.40(+08) 1.55 3.12(+09) 2.22(+09) 1.10(+09) 3.76(+08) 1.99 1.96(+08) 1.18(+08) 7.04(+07) 4.10(+07) 2.38 1.43(+08) 1.08(+08) 5.60(+07) 1.98(+07) 2.75 2.44(+05) 2.38(+05) 2.34(+05) 2.24(+05) 3.25 9.28(+03) 8.94(+03) 8.70(+03) 8.36(+03)

• Numbers in parenthesis denotes powers of ten T12.2-43 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-36 FISSION PRODUCT GAMMA SOURCE IN CONTAINMENT BUILDING (Mev/sec)

(assuming 100% noble gases, 50% halogens, 1% solids)

Energy Interval (Mev)

Time .1 - .4 .4 - .9 .9 - 1.35 1.35 - 1.8 1.8 - 2.2 2.2 - 2.6 2.6 0 3.08(18)* 1.84(19) 7.30(18) 1.37(19) 8.90(18) 6.41(18) 2.90(18)

.5 hr. 2.93(18) 1.62(19) 6.65(18) 5.02(18) 4.63(18) 5.19(18) 3.71(17) 1 hr. 2.82(18) 1.43(19) 5.81(18) 4.45(18) 3.12(18) 4.30(18) 1.36(17) 2 hr. 2.68(18) 1.14(19) 4.70(18) 3.56(18) 2.14(18) 3.05(18) 4.38(16) 8 hr. 2.09(18) 6.41(18) 2.16(18) 1.58(18) 6.91(17) 5.76(17) 1.56(16) 24 hr. 1.16(18) 4.45(18) 8.68(17) 6.23(17) 2.32(17) 1.03(17) 2.57(14) 1 wk. 3.06(17) 1.07(18) 1.63(17) 1.75(17) 5.88(16) 3.11(16) 1.74(14) 1 mo. 4.10(16) 1.28(17) 3.17(15) 2.42(16) 1.46(15) 1.85(15) 5.52(13) 2 mo. 5.98(15) 7.86(16) 5.37(14) 4.30(15) 7.72(14) 2.85(14) 1.78(13) 4 mo. 7.51(14) 4.35(16) 1.98(14) 4.87(14) 6.45(14) 3.41(13) 7.72(12)

• Denotes power of ten (10)

T12.2-44 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-37 ASSUMPTIONS AND PARAMETERS USED TO CALCULATE AIRBORNE CONCENTRATIONS Leak Rates:

Containment 1% of the noble gas inventory per day

.001% of the iodine inventory per day Reactor Auxiliary Building 160 lb per day Cubicals and other areas Table 12.2-39 Partition Factors:

Reactor Auxiliary Building 0.0075 for iodines, 1 for noble gases Letdown heat exchanger 0.1 for iodines, 1 for noble gases Ventilation Rates (cfm):

Containment Isolated case Fuel Handling Building 19,700 (Fuel Pool) 25,700 (Stack)

Reactor Auxiliary Building 165,000 Volumes (Cu ft):

Containment 2.5 x 106 Fuel Handling Building 3.3 x 104 Reactor Auxiliary Building 1.48 x 106 Other Factors:

Failed fuel fraction 0.12%

Plant load 80%

Outside air condition (winter) 27 F, 65.0% Relative Humidity Fuel Pool Parameters:

Surface temperature 130 F T12.2-45 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-37 (Cont'd)

Fuel Pool Parameters: (Cont'd)

Surface area 567 ft2 Air velocity over surface 10 ft/min T12.2-46 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-38 AVERAGE AIRBORNE C/MPC**** IN REACTOR AUXILIARY BUILDING CONTAINMENT AND FUEL HANDLING BUILDING ISOTOPE CONTAINMENT FUEL HANDLING BUILDING REACTOR AUXILIARY (C/MPC) TURBINE BUILDING (C/MPC) BUILDING (C/MPC)

Kr-83m 8.76 (-2)* -** -*** 2.78 (-4)

Kr-85m 1.82 (-1) - - 2.52 (-4)

Kr-85 2.35 (+3) - - 1.70 (-4)

Kr-87 1.73 (-1) - - 7.85 (-4)

Kr-88 1.26 (0) - - 2.68 (-3)

Xe-131m 3.34 (0) - - 7.35 (-5)

Xe-133m 2.69 (0) - - 3.09 (-4)

Xe-133 5.04 (+2) - - 2.47 (-2)

Xe-135m 4.05 (-2) - - 1.85 (-4)

Xe-135 1.83 (0) - - 1.24 (-3)

Xe-137 1.34 (-3) - - 4.93 (-5)

Xe-138 2.81 (-2) - - 4.52 (-4)

I-131 1.92 (+1) - - 4.64 (-3)

I-133 7.67 (-1) - - 1.70 (-3)

H-3 3.80 - 0.5 1.36 (-4)

• Represents power of 10

• • Concentrations are expected to be negligible since the Turbine Building is of an open design

• • • Indicates negligible C/MPC

• • • • See text note in Section 12.2.2 regarding use of MPC versus DAC T12.2-47 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-39 REACTOR AUXILIARY BUILDING AND FUEL HANDLING BUILDING ROOM BY ROOM Ci/MPC(b) AND WHOLE BODY DOSE COMMITMENT VALUES Elevation Leakage Dose Commitment (mrem/hr occupancy)

Item Location and/or Component (Ft. MSL) Rate Ci/MPC Inhalation External (GPD) Whole Body Whole Body 1 Boric Acid Condensate Tanks 2A & 2B -0.5 1.11 7.28(-4)(a) 7.16(-4)(a) 1.37(-8)(a) 2 Boric Acid Condensate Pumps 2A & 2B -0.5 2.31 3.31(-3) 3.25(-3) 6.21(-8) 3 Boric Acid Holding Pumps 2A & 2B -0.5 3.23 4.62(-3) 4.55(-3) 6.68(-8) 4 Boric Acid Holding Tank -0.5 3.11 6.12(-3) 6.02(-3) 1.15(-7) 5 Spent Resin Tank -0.5 2.41 4.04 1.04 9.64(-4) 6 Gas Decay Tank 2C -0.5 7.95(-5)CFM 3.02 2.81(-2) 1.27(-2) 7 Gas Decay Tank 2B -0.5 7.95(-5)CFM 3.02 2.81(-2) 1.27(-2) 8 Gas Decay Tank 2A -0.5 7.95(-5)CFM 3.02 2.81(-2) 1.27(-2) 9 Gas Surge Tank -0.5 7.36(-5)CFM 1.62(+1) 8.67(-3) 6.88(-2) 10 Waste Gas Compressor 2A -0.5 2.39(-3)CFM 1.50(+2) 8.05(-2) 6.39(-1) 11 Waste Gas Compressor 2B -0.5 2.39(-3)CFM 1.50(+2) 8.05(-2) 6.39(-1) 12 Charging Pumps 2A, 2B and 2C -0.5 8.00(+1) 2.36(+1) 1.75(-2) 1.23(-1) 13 Boric Acid Makeup Tanks and Pumps 2A & 2B -0.5 6.12 3.17(-3) 1.54(-3) 4.81(-6) 14 Holdup Recirc. Pump -0.5 1.30 7.14(-1) 4.08(-3) 3.66(-3) 15 Holdup Drain Pump 2A -0.5 1.24 6.81(-1) 3.89(-3) 3.49(-3) 16 Holdup Drain Pump 2B -0.5 1.30 7.14(-1) 4.08(-3) 3.66(-3) 17 Boric Acid Pre Concentrator Filters 2A and 2B -0.5 8.28 7.72(-3) 5.85(-3) 8.85(-6)

Condensate Recovery Tank and Pumps 2A and 2B 18 Holdup Tank 2C -0.5 1.05 1.92(-1) 1.10(-3) 9.86(-4) 19 Holdup Tank 2D -0.5 1.14 2.09(-1) 1.19(-3) 1.07(-3) 20 Holdup Tank 2B -0.5 1.05 1.92(+1) 1.10(-1) 9.86(-2) 21 Holdup Tank 2A -0.5 1.49 2.75(-1) 1.57(-3) 1.41(-3) 22 Waste Condensate Tanks 2A and 2B -0.5 1.78 1.27(-3) 1.25(-3) 5.20(-12)

Notes: (a) Represents power of 10 (b) See text note in Section 12.2.2 regarding use of MPC versus DAC T12.2-48 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-39 (Cont'd)

Elevation Leakage Dose Commitment (m rem/hr occupancy)

Item Location and/or Component (Ft. MSL) Rate Ci/MPC Inhalation External (GPD) Whole Body Whole Body 23 Waste Condensate Pumps 2A and 2B -0.5 3.23 6.33(-3) 6.25(-3) 2.59(-11) 24 ECCS Area 2B -0.5 6.34 7.29(-1) 3.34((-2) 1.13(-3) 25 ECCS Area 2A -0.5 9.13 1.07 4.88(-2) 1.62(-3)

(Including Reactor Drain Pumps 2A and 2B) 26 Shutdown Cooling Heat Exchanger 2A -0.5 0.89 5.03(-1) 2.29(-2) 7.59(-4) 27 Shutdown Cooling Heat Exchanger 2B -0.5 0.89 1.56(-1) 1.30(-2) 7.33(-4) 28 Aerated Waste Storage Tank -0.5 3.14 6.18(-4) 6.08(-4) 6.84(-10)

Equipment Drain Pumps 2B & 2C 29 Laundry Filter and Waste Filters -0.5 4.34 1.65(-3) 1.57(-3) 5.51(-9)

Laundry Drain Tanks and Pumps 2A and 2B 30 Chemical Drain Tank -0.5 3.68 6.43(-3) 2.12(-3) 4.11(-7)

Equipment Drain Tank Chemical Drain Pump Equipment Drain Pump 2A 31 Pipe Tunnel and Penetration Area -0.5/19.5 7.04 7.97(-1) 3.62(-2) 1.20(-3) 32 Waste Concentrator

• 19.5 8.50 5.80(-3) 5, 73(-3) 2.01(-9) 33 Boric Acid Concentrator 2A

• 19.5 8.50 2.08(-2) 3.33(-3) 8.52(-5) 34 Boric Acid Concentrator 2B

• 19.5 8.50 2.08(-2) 3.33(-3) 8.52(-5) 35 Pre Concentrator Ion Exchanger 2B 19.5 0.86 4.36 3.50 3.58(-3) 36 Pre Concentrator Ion Exchanger 2A 19.5 0.95 3.87 3.11 3.18(-3) 37 Waste Ion Exchanger 19.5 0.98 3.08(-3) 3.05(-3) 2.20(-10) 38 Boric Acid Condensate Ion Exchanger 2B

• 19.5 0.73 9.83(-3) 1.06(-2) 6.92(-6) 39 Boric Acid Condensate Ion Exchanger 2A

• 19.5 0.92 1.55(-2) 1.67(-2) 1.09(-5) 40 CVC Purification Ion Exchanger 2A 19.5 0.76 1.72 7.83(-2) 2.60(-3) 41 CVC Purification Ion Exchanger 2B 19.5 0.76 1.44(+0) 6.56(-2) 2.17(-3) 42 CVC Deborating Ion Exchanger 19.5 1.36 2.58 1.17(-1) 3.89(-3)

• Note: Items are no longer used.

T12.2-49 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-39 (Cont'd)

Elevation Leakage Dose Commitment (m rem/hr occupancy)

Item Location and/or Component (Ft. MSL) Rate Ci/MPC Inhalation External (GPD) Whole Body Whole Body 43 Fuel Fool Ion Exchanger 19.5 0.86 2.07 1.33(-1) 2.72(-4) 44 Volume Control Tank 19.5 0.82 3.16 2.34(-3) 1.64(-2) 45 Volume Control Tank Piping and Valves 19.5 27.53 9.24(+1) 6.83(-2) 4.79(-1) 46 Flash Tank 19.5 1.93 1.62 1.13(-2) 8.16(-3) 47 Flash Tank Pumps 2A and 2B 19.5 3.71 1.56 1.09(-2) 7.84(-3) 48 Purification Filters 2A and 2B 19.5 1.93 4.37 1.99(-1) 6.60(-3) 49 Letdown Heat Exchanger 19.5 1.33 1.08 4.89(-2) 1.62(-3) 50 Fuel Pool Purification Filter 19.5 0.38 8.09(-4) 7.38(-4) 2.69(-7) 51 Fuel Pool Purification Pump 19.5 0.95 8.08(-4) 7.37(-4) 2.68(-7) 52 Fuel Pool Cooling Pumps 2A and 2B 19.5 2.89 1.04(-2) 1.08(-3) 4.04(-5) 53 Fuel Pool Heat Exchanger 19.5 1.33 5.80(-3) 6.03(-4) 2.26(-5)

Notes:

(a) represents powers of 10 T12.2-50 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-40 POST ACCIDENT SAMPLING SYSTEM COMPONENT MAXIMUM INVENTORIES (CURIES)

Reactor Containment Coolant Sample Vessel/ Gas Residence Boron Atmosphere Isotope Tubing Heat Exchanger Chamber Meter Tubing Kr-85m 5.6(1)A 5.0(1) 4.3(2) 1.4(1) 9.7(-2)

Kr-85 1.8 1.6 1.4(1) 4.4(-1) 3.0(-3)

Kr-87 1.0(2) 9.1(1) 7.8(2) 2.6(1) 1.8(-1)

Kr-88 1.5(2) 1.3(2) 1.1(3) 3.6(1) 2.5(-1)

Kr-89 1.8(2) 1.6(2) 1.4(3) 4.5(1) 3.2(-1)

Kr-90 1.8(2) 1.6(2) 1.4(3) 4.4(1) 3.2(-1)

Kr-91 1.3(2) 1.2(2) 1.0(3) 3.2(1) 2.3(-1)

Xe-131m 1.6 1.4 1.2(1) 3.9(-1) 2.7(-3)

Xe-133 4.5(2) 4.0(2) 3.4(3) 1.1(2) 7.7(-1)

Xe-135m 9.1(1) 8.1(1) 6.9(2) 2.2(1) 1.6(-1)

Xe-135 8.1(1) 7.2(1) 6.1(2) 2.0(1) 1.5(-1)

Xe-137 4.0(2) 3.5(2) 3.0(3) 9.8(1) 6.8(-1)

Xe-138 3.6(2) 3.2(2) 2.7(3) 8.9(1) 6.2(-1)

Xe-140 1.8(2) 1.6(2) 1.4(3) 4.6(1) 3.2(-1)

Xe-143 4.4 3.9 3.3(1) 1.1 7.5(-3)

Xe-144 9.8(-1) 8.7(-1) 7.4 2.4(-1) 1.7(-3)

Br-84 2.2(1) 1.9(1) 1.6(2) 5.3 1.9(-2)

Br-85 2.8(1) 2.5(1) 2.1(2) 6.8 2.4(-2)

Br-87 4.5(1) 4.0(1) 3.4(2) 1.1(1) 3.8(-2)

Br-88 4.7(1) 4.2(1) 3.6(2) 1.2(1) 4.0(-2)

Br-89 3.3(1) 2.9(1) 2.5(2) 8.0 2.8(-2)

Br-90 2.2(1) 1.8(1) 1.6(2) 5.1 1.8(-2)

I-127 Stable I-129 2.8(-6) 2.5(-6) 2.1(-5) 6.9(-7) 2.4(-9)

I-131 1.1(2) 9.9(1) 8.4(2) 2.8(1) 1.0(-1)

I-132 1.6(2) 1.4(2) 1.2(3) 4.0(1) 1.4(-1)

I-133 2.3(2) 2.0(2) 1.7(3) 5.6(1) 1.9(-1)

I-134 2.4(2) 2.2(2) 1.8(3) 6.0(1) 2.1(-1)

I-135 2.1(2) 1.9(2) 1.6(3) 5.2(1) 1.8(-1)

I-137 1.0(2) 8.3(1) 7.1(2) 2.3(1) 8.0(-2)

I-138 4.7(1) 4.2(1) 3.6(2) 1.2(1) 4.0(-2)

T12.2-51 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-40 (Cont'd)

Reactor Coolant Sample Vessel/ Gas Residence Boron Isotope Tubing Heat Exchanger Chamber Meter Se-84 4.1(-1) 3.6(-1) 3.1 1.0(-1)

As-85 7.1(-2) 6.3(-2) 5.4(-1) 1.8(-2)

Se-85 2.5(-1) 2.3(-1) 1.9 6.3(-2)

Se-87 4.1(-1) 3.6(-1) 3.1 1.0(-1)

Rb-88 1.5 1.3 1.1(1) 3.7(-1)

Sr-89 2.1 2.1 1.6(1) 5.1(-1)

Rb-90 1.8 1.6 1.4(1) 4.5(-1)

Sr-90 1.4(-1) 1.3(-1) 1.1 3.6(-2)

Y-90 1.5(-1) 1.3(-1) 1.2 3.8(-2)

Rb-91 2.4 2.1 1.8(1) 5.8(-1)

Sr-91 2.5 2.3 1.9(1) 6.3(-1)

Y-91M 1.5 1.3 1.1(1) 3.6(-1)

Y-91 2.7 2.4 2.1(1) 6.7(-1)

Sr-95 2.7 2.4 2.1(1) 6.7(-1)

Y-95 3.6 3.2 2.7(1) 8.9(-1)

Zr-99 3.8 3.3 2.8(1) 9.3(-1)

Nb-95 3.8 3.3 2.9(1) 9.4(-1)

Zr-99 3.7 3.3 2.8(1) 9.2(-1)

Nb-99 3.9 3.4 2.9(1) 9.6(-1)

Mo-99 4.1 3.6 3.0(1) 1.0 Tc-99m 3.5 3.1 2.7(1) 8.7(-1)

Mo-103 3.6 3.2 2.7(1) 8.9(-1)

Tc-103 3.6 3.2 2.7(1) 8.9(-1)

Ru-103 3.6 3.2 2.8(1) 8.9(-1)

Tc-106 1.5 1.3 1.2(1) 3.7(-1)

Ru-106 1.0 9.2(-1) 7.9 2.6(-1)

Sn-129 2.4(-1) 2.1(-1) 1.8 5.9(-2)

Sb-129 7.4(-1) 6.6(-1) 5.6 1.8(-1)

Te-129m 1.9(-1) 1.1(-1) 1.5 4.7(-2)

Te-129 7.0(-1) 6.2(-1) 5.3 1.7(-1)

Sn-131 6.6(-1) 5.8(-1) 5.0 1.6(-1)

Sb-131 1.8 1.6 1.4(1) 4.5(-1)

Te-131m 3.4(-1) 3.0(-1) 2.6 8.3(-2)

Te-131 1.9 1.7 1.5(1) 4.8(-1)

Sn-132 3.8(-1) 3.4(-1) 2.9 9.4(-2)

A - (X) means 10x T12.2-52 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.2-40 (Cont'd)

Reactor Coolant Sample Vessel/ Gas Residence Boron Isotope Tubing Heat Exchanger Chamber Meter Sb-132 1.1 9.5(-1) 8.2 2.7(-1)

Te-132 3.2 2.8 2.4(1) 7.9(-1)

Sn-133 1.3(-1) 1.2(-1) 1.0 3.3(-2)

Sb-133 1.2 1.1 9.1 3.0(-1)

Te-133m 1.6 1.4 1.2(1) 4.0(-1)

Te-133 2.6 2.3 2.0(1) 6.4(-1)

Cs-134 4.2(-1) 3.8(-1) 3.2 1.1(-1)

Sb-134 2.1(-1) 1.9(-1) 1.6 5.3(-2)

Te-134 3.4 3.0 2.6(1) 8.4(-1)

Sb-135 1.3(-1) 1.2(-1) 1.0 3.3(-2)

Te-135 1.8 1.6 1.3(1) 4.4(-1)

Cs-135 5.6(-7) 5.0(-7) 4.3(-6) 1.4(-7)

Cs-136 1.2(-1) 1.1(-1) 9.0(-1) 2.9(-2)

Cs-137 1.9(-1) 1.7(-1) 1.5 4.8(-2)

Ba-137m 1.8(-1) 1.6(-1) 1.4 4.6(-2)

Cs-138 3.8 3.4 2.9(1) 9.4(-1)

Cs-140 3.5 3.1 2.6(1) 8.6(-1)

Ba-140 3.9 3.5 3.0(1) 9.7(-1)

La-140 4.0 3.6 3.1(1) 1.0 Cs-143 7.4(-1) 6.6(-1) 5.7 1.8(-1)

Ba-143 3.0 2.6 2.2(1) 7.3(-1)

La-143 3.4 3.0 2.5(1) 8.3(-1)

Ce-143 3.4 3.0 2.6(1) 8.3(-1)

Pr-143 3.3 2.9 2.5(1) 8.2(-1)

Cs-144 2.3(-1) 2.0(-1) 1.73 5.6(-2)

Ba-144 2.2 2.0 1.7(1) 5.5(-1)

La-144 2.9 2.6 2.2(1) 7.2(-1)

Ce-144 2.7 2.4 2.0(1) 6.6(-1)

Pr-144 2.7 2.4 2.0(1) 6.6(-1)

T12.2-53 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 FACILITY DESIGN FEATURES Compliance with the design feature guidance of Regulatory Guide 8.8 is discussed in Subsection 12.1.2. The layout of plant radiation zones is described in Subsection 12.3.2.1 and Section 12.4. The counting room and other facilities and equipment related to the use of nuclear material is described in Section 12.5. The locations of sampling ports are discussed in Subsection 9.3.2. Instrumentation and control panels are located in shielded areas in as low a radiation area as reasonably achievable for each individual case.

Dose calculations are based on a minimum concrete density of 130 lb/ft3.

General arrangement drawings of the building containing process equipment for treatment of radioactive fluids, and also a site plan are shown in Figures 1.2-1 through -19.

High radiation areas are conspicuously posted in accordance with 10 CFR 20.

Original design of the plant shielding was performed assuming a core power level of 2700 MWt, a 12-month fuel cycle length, and system activity levels stemming from one percent failed fuel (due to fuel cladding defects). The plant shielding was re-evaluated for the extended power uprate assuming a core thermal power of 3030 MWt and an 18-month fuel cycle. Taking into consideration the conservative analytical techniques used to establish the original shielding design and the plant Technical Specifications, which restrict the reactor coolant activity to levels significantly less than 1% failed fuel (due to fuel defects), it is concluded that the increase in the core power level and current operation with an 18-month fuel cycle will have no significant impact on plant shielding adequacy and safe plant operation.

12.3.1.1 Reactor Building Primary Shield The primary shield consists of a minimum of 7'-3" of reinforced concrete surrounding the reactor vessel. The annular cavity between the primary shield and the reactor vessel is air cooled to prevent overheating and dehydration of the concrete shield.

The primary shield limits radiation emanating from the reactor vessel; during operation this radiation consists of both fast and slow neutrons emitted from the core, prompt fission gammas, fission product gammas, and gamma radiation resulting from neutron capture in the core internals and vessel. Following shutdown, only fission product gammas, and gamma radiation from neutron activation of the coolant and corrosion products are present. Neutron and gamma fluxes resulting from the normal, full power operation of the core are used as the basis for determining the primary shield design.

The primary shield arrangement and its thickness (shown in Figures 3.8-38 and 3.8-39) are designed to:

a. attenuate the core neutron flux in order to limit the activation of component and structural materials,

b. limit the radiation level after shutdown in order to permit access to the Reactor Coolant System equipment, 12.3-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2

c. reduce, in conjunction with the secondary shield and the neutron streaming shield, the radiation level from sources within the reactor vessel in order to allow limited access to the containment during normal operation, and

d. permit access during shutdown for inspections.

12.3.1.2 Secondary Shield The secondary shield is reinforced concrete 4'-0" thick surrounding the reactor coolant piping, pumps, steam generators, and pressurizer.

The secondary shield in concert with the primary shield reduces the radiation activity from the Reactor Coolant System to a level which allows limited access to the containment during operation. The controlling radiation source in the design of the secondary shield is N-16 resulting from the (n,p) reaction with the oxygen in the coolant. After reactor shutdown, the fission and corrosion product activities in the Reactor Coolant System become the dominant sources.

In addition, a partial 2'-0" thick reinforced concrete wall is located midway between the primary and secondary shields in the main steam and feedwater lines' penetration region, to prevent streaming of gamma radiation from the Reactor Coolant System through the penetration openings.

The secondary shield system is shown in Figures 3.8-38 and 3.8-39.

12.3.1.3 Shield Building The steel containment structure is enclosed by a reinforced concrete Shield Building with 3'-0" thick cylindrical walls and a 2'-6" thick dome. An annular area separates the structures.

In conjunction with the primary and secondary shields, the Shield Building limits the radiation level from all sources within the containment to less than 0.25 mrem/hr outside the structure during full power operation, assuming one percent failed fuel (due to fuel defects).

Taking into consideration the conservative analytical techniques used to establish the Reactor Building secondary shield and Shield Building and pre-uprate survey dose rates outside the containment wall, the secondary shield and Shield Building were determined to be adequate for extended power uprate operation.

12.3.1.4 Neutron Streaming Shielding The neutron streaming shield, located in the reactor cavity area, serves to reduce exposure doses to personnel in the Reactor Building during operation due to neutron and gamma radiation streaming up the annular gap between the reactor vessel and the cavity walls. The shielding arrangement consists of the following components (refer to Figure 12.3-3):

a. A permanent, one ft thick concrete wall located 8'-3" from the edge of the missile shield at elevation 62 foot, and extending to elevation 68 ft. This wall is oriented perpendicularly to the length of the refueling cavity.

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b. Two removable, one ft. thick concrete walls extending from the steam generator cubicle walls to the wall described in a) above, and extending from elevation 62 ft to elevation 68 ft.

c. Removable, one foot thick steel encased Permali and concrete panels forming a wall extending from elevation 36 ft. of the refueling cavity to elevation 68.50 ft.

This wall is also oriented perpendicularly to the length of the refueling cavity and opposite the permanent wall described in a) above. The top of this wall abuts on the missile shield thus forming a "closed box" to radiation.

12.3.1.5 Fuel Transfer Tube Shielding Fuel transfer tube shielding protects plant personnel from fission product gamma radiation emitted from the spent fuel elements during core refueling operations. The fuel is removed from the reactor through a canal to a water filled spent fuel storage area located in the Fuel Handling Building. After sufficient decay, the spent fuel may be transferred under water to a spent fuel transfer cask.

Consideration has been given to direct radiation from a spent fuel assembly in the transfer tube, as well as to radiation that may stream through the spacings between shielding and structures.

Three shielding areas are identified: (1) between the Fuel Handling and Shield Buildings; (2) between the concrete Shield Building wall and the steel containment; and (3) between the steel containment and the refueling cavity wall.

Area (1) is shielded by a concrete wall between the Fuel Handling Building and the curved surface of the Shield Building. Radiation that may stream through the space between this concrete wall and the Shield Building is reduced by a lead collar around the transport tube. This collar is crescent shaped and extends 300 degrees around the tube thus providing shielding to potentially accessible areas. Access to area (1) is restricted by a fenced-off enclosure during refueling operations.

Area (2) is shielded on all sides by concrete shielding. A reinforced concrete wall to the south of the fuel transfer tube supplements the concrete labyrinth shield in this area. A lead brick hatchway is provided in the concrete wall to permit in-service inspection and maintenance of the fuel transfer tube. The gap between the concrete shielding and the steel containment vessel is closed off by lead-shot-filled steel tubing to the north, south and above the fuel transfer tube.

Area (3) is shielded on all sides by concrete shielding. Access to the transfer tube for inspection and maintenance is provided by a labyrinth. The entrance to the labyrinth is administratively locked during refueling operations.

In addition, a radiation work permit is required to access areas (2) & (3) during all modes of plant operation. The shielding directly above the transfer tube in both areas (2) & (3) consists of removable lead bricks supported by a steel frame. This permits replacement of the transfer tube bellows, if required. Also, the steel containment is thickened from two inches to four inches around the transfer tube to reduce streaming.

12.3.1.6 Fuel Handling Building Shielding Shielding is provided for radiation protection of plant personnel during all phases of spent fuel removal, storage and preparation for offsite shipment. Operations requiring the shielding of 12.3-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 personnel are: spent fuel removal from the reactor, spent fuel transfer through the refueling canal and transfer tube, spent fuel storage, spent fuel transfer cask loading prior to transfer, and maintenance and inspection of the spent fuel pool purification loop.

The refueling cavity above the reactor vessel flange is flooded to elevation 60 ft to provide 24 ft of water shielding above the reactor vessel flange. This height assures 132 in. of water above the active portion of a withdrawn fuel assembly at its highest point of travel. Under these conditions, the dose rate from the spent fuel assembly is less than 2.5 mrem/hr at the water surface. A concrete shield (six ft on north, south and west sides and three ft on east side) around the refueling cavity protects the refueling personnel from radiation from the spent fuel assemblies and reactor internals.

The refueling water and concrete walls also provide shielding from those activated control element assemblies (CEAs) and reactor internals which are removed during refueling.

Spent fuel removal and transfer operations are carried out under a minimum depth of 9'-0" of borated water (under minimum water level) to provide radiation protection. The dose rate at the water surface is less than 2.5 mrem/hr. The concrete walls of the fuel transfer canal and spent fuel pool supplement the water shielding and limit the maximum continuous radiation dose levels in working areas to less than 2.5 mrem/hr from spent fuel sources. The concrete sides of the fuel pool are 6'-0" thick to ensure a dose rate of less than 0.5 mrem/hr at the outer surface of the structure.

The radiation levels are closely monitored during refueling operations to ensure doses for plant personnel do not exceed the integrated dose specified in 10 CFR 20.

Dose calculations assume a source equivalent to the most active spent fuel assembly two days after shutdown. It is further assumed in the calculations that the fuel storage racks are full with 1076 assemblies. Details of spent fuel storage are presented in Subsection 9.1.2.

Taking into consideration the conservative analytical techniques used to establish the fuel handling shielding and pre-uprate survey dose rates in the area, the fuel handling shielding was determined to be adequate for extended power uprate operation.

12.3.1.7 Reactor Auxiliary Building Shielding Reactor Auxiliary Building shielding is designed to protect personnel working near various system components, such as in the Chemical and Volume Control, Waste Management, Shutdown Cooling, and Sampling Systems. Detailed radiation design features appear in Subsection 12.1.2. These features are in agreement with ALARA considerations as presented in Regulatory Guide 8.8, "Information Relevant to Ensuring that ORE at Nuclear Power Stations will be ALARA," June, 1978 (R3) including facilitation of maintenance, ease of access, provisions to minimize crud buildup, provisions for decontamination, cubicle shielding of components, labyrinths, shielded pipe chases, shielded valve stations, etc. Source terms used as a basis for shielding analyses appear in Sections 11.1 and 12.1.

The concrete thickness of compartment shield walls is sufficient to reduce the dose rate outside the compartments in normally accessible areas to less than 2.5 mrem/hr.

Taking into consideration the conservative analytical techniques used to establish compartment shield walls and plant technical specifications that restrict coolant activity levels significantly less 12.3-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 than 1% failed fuel (due to fuel defects), the compartment shield walls were determined to be adequate for extended power uprate operation.

Shielding of the Emergency Core Cooling System room is designed for normal operation and shutdown conditions. Under these operating conditions only the reactor drain pump and the low pressure safety injection pump(s) are sources of radiation. Sufficient shielding is achieved with one foot to two feet thick concrete walls. A two foot thick part height concrete wall divides the rooms into two subcompartments. One subcompartment houses one low pressure safety injection pump, one high pressure safety injection pump, one containment spray pump and two reactor drain tank pumps; the other subcompartment houses one low pressure safety injection pump, one high pressure safety injection pump and one containment spray pump. This arrangement allows inspection and servicing of the safety feature systems during plant operation and shutdown.

Pumps, which frequently require service are segregated from their respective tanks, which are infrequently serviced. These tanks are normally low activity tanks (less than 10 mr/hr contact).

Should crud buildup occur, the infrequent maintenance of these tanks would require the use of part height temporary shields for protection against radiation from the bottom of the tank.

The partial shield walls utilized in radiation Zone V areas, to separate high frequency maintenance items such as pumps and valves from tanks and other potentially radioactive components, are of sufficient height so that maintenance can be performed with a minimum exposure to direct radiation.

Access to the flash tank, ion exchangers and letdown system valves is administratively controlled. Access to the spent resin tank valve operating enclosure is not controlled since there is no unshielded piping carrying activity continuously in the vicinity of the enclosure.

Airborne contamination levels are determined prior to servicing or operation of the valves in controlled access areas by suitable sampling. The level of airborne contamination in valve enclosure areas where access is unrestricted is monitored periodically during health physics surveys.

12.3.1.8 Control Room Shielding The Control Room Radiation Protection System is designed to limit radiation exposure to 0.25 mrem/hr during normal operation, and to a total integrated dose of 5.0 rem for the duration of a DBA, as established in GDC 19 of 10 CFR 50. This protection consists of control room shielding, a HVAC system and a Control Room Emergency Cleanup System (see Subsection 6.4.2.2).

Several postulated radiation sources external to the control room envelope are considered as discussed in Subsection 6.4.2.5; total integrated doses from each source appears in Table 6.4-2. The control room is shielded from each source by at least two feet of concrete, except for a short section of concrete wall separating the control room and air conditioning units, which is one foot thick. Control room shielding limits the direct radiation exposure of personnel, for 30 days post LOCA, to less than 1.4 rem.

12.3-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3.2 SHIELDING 12.3.2.1 Design Objectives The primary design objective of the plant radiation shielding is to protect plant operating personnel and the general public against radiation exposure during normal operation, anticipated operational occurrences, postulated accident conditions, and maintenance.

This objective is accomplished by designing the shielding to perform the following functions:

a. Limit inplant exposure doses of plant personnel, contractors, and authorized site visitors to as far below the limits set forth in 10 CFR 20 as reasonably achievable for normal operation, anticipated operational occurrences, and maintenance in conformance with Regulatory Guide 8.8(R3). Further discussion relative to this regulatory guide is found in Section 12.1.

b. Limit radiation exposure of control room personnel, in the unlikely event of an accident, to allow habitability of the control room as specified in 10 CFR 50, Appendix A, GDC 19, by limiting the total integrated dose over 30 days following the accident to five rem.

c. Limit exposures to the general public (offsite) from direct and air-scattered radiation to a small fraction of the limits set forth in 10 CFR 20 during normal operation and anticipated operational occurrences, and to within the limits specified in 10 CFR 100 for postulated accident conditions.

d. Provide barriers for restricting personnel access to high radiation areas and for controlling the spread of contaminants.

e. Protect certain plant components from excessive radiation damage or activation as follows:

1. Reduce neutron activation of equipment, piping, supports and other materials by the use of suitable shielding around the reactor vessel, by minimizing neutron streaming into the reactor cavity upper reaches, steam generator subcompartments, and general containment spaces.

Subsection 12.3.1.4 describes the neutron streaming shielding.

2. Limit gamma radiation damage to equipment and materials (such as cables and gaskets) to below their individual design-life limits.

To comply with the above objectives, the plant shielding is designed to reduce radiation levels throughout the plant, from direct and scattered neutron and gamma radiation, to the dose limits specified in Table 12.3-1.

Other criteria for shielding, such as zone levels, shielding design, and personnel exposures are discussed in other sections of Chapter 12.

12.3-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3.2.2 Radiation Zone Description For shielding design purposes, the plant has been divided into radiation zones, based on the maximum zone dose rate levels listed in Table 12.3-1.

A description of each radiation zone chosen for design purposes is given below and shown on Figures 12.3-4 through 12.3-12. These descriptions contain references to 10 CFR 20 sections and exposure limits which are outdated; however, this information has intentionally not been revised since it represents the plant's design and licensing basis with respect to the radiation zones.

a. Zone I (<0.25 mrem/hr)

This zone has no restriction on occupancy. Such a zone represents areas in the plant where radiation, based on occupancy on a 40 hr/wk, 50 wk/yr basis, does not exceed the whole body dose of 0.5 rem/yr, as specified in paragraph 20.105 of 10 CFR 20. Most non-employees and visitors to the site receive considerably less than 0.5 rem/yr because of the relatively short time interval during which they are onsite.

b. Zone II (0.25-2.5 mrem/hr)

This zone is a controlled (restricted) area which can be occupied by plant personnel and authorized visitors on a 40 hr/wk, 50 wk/yr basis without exceeding the allowable whole body dose of 1.25 rem/ calendar quarter (10 CFR 20.101). This zone requires no posting.

c. Zone III (2.5-15.0 mrem/hr)

This is a restricted radiation area (10 CFR 20.202) that plant personnel can occupy on a periodic basis. The average radiation level in this zone may vary from 2.5 to 15.0 mr/hr with occasional "hot points" slightly exceeding the 15 mr/hr limit. The zone is posted with "Caution-Radiation Area" signs. However, any areas within this zone remain accessible to plant personnel.

d. Zone IV (15.0-100.0 mrem/hr)

This zone represents a restricted radiation area which is posted with "Caution-Radiation Area" signs. The average radiation level may vary from 15.0 mr/hr to 100 mr/hr. Occupancy is limited. However, qualified personnel who have been issued a radiation work permit for routine access can enter these areas for brief periods of time to operate and inspect components. The length of stay in these areas is determined by the actual radiation level in the area, the past radiation history of the person entering, and the nature of the radiation.

e. Zone V (>100 mrem/hr)

This zone is a high radiation area (10 CFR 20.203). As such, it is posted with "Caution-High Radiation Area" signs and control is exercised over access to it at all times. Access is permitted with the issuance of a radiation work permit.

Locked wire mesh doors or louvered doors are provided to prevent unauthorized 12.3-7 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 entry into areas in which the intensity of radiation is greater than 1000 mr/hr.

Access is controlled through approved procedures.

12.3.2.3 Shielding Design Methods Calculations are performed to determine fluxes, dose rates, activities, etc., which then are used, after applying suitable criteria, to determine design parameters such as shield wall thickness.

Nuclear physics data required in these calculations are obtained from a variety of sources. For example, the ansiotropic, multigroup cross section sets used by the DOT or MORSE transport codes are extracted from data libraries available from the Radiation Shielding Information Center at Oak Ridge National Laboratory, which has processed them from the basic ENDF/B data(3). Decay constants and decay gamma yields are primarily obtained from The Table of the Isotopes(4) ANSI standards, such as that providing flux-to-dose conversion factors(5) are also used. Other sources of information include: Reactor Physics Constants (ANL-5800)(6);

Rockwell's Reactor Shielding Design Manual(7); Schaeffer's Reactor Shielding for Nuclear Engineers(8); as well as many other more specialized reports and books.

Radiation sources, which become input to shielding calculations, are determined as indicated in Section 12.2. Source terms are based on ample conservatisms such as one percent failed fuel, among others.

Radiation gamma shielding calculations are performed using the Rockwell point-kernel integration method,(7) or by using the computer code, ISOSHLD(9), which has automated Rockwell's method. This program calculates the gamma dose rate at a point exterior to the source region for a number of common geometric arrangements of sources and shields as encountered in nuclear power plants. Some geometries which can be specified are: point, linear, spherical, truncated cone, disk, cylindrical, and parallel piped. Slab shields may be used for all cases, while spherical shields can only be used in conjunction with spherical and point sources; and cylindrical shields in conjunction with a cylindrical source.

The actual shielding configuration is approximated by a suitable combination of sources and shields. Tanks and large pipes containing fluids are approximated by cylindrical sources. Gas filled tanks and pipes are simulated by line sources, as are small fluid containing pipes. In cases where the geometry proves to be sufficiently complex to preclude use of the Rockwell method or the ISOSHLD program, an advanced point-kernel integration code, SPAN-4(10) is used. This program calculates the dose rate at a point from any number of sources having complex geometry and complex shield configurations. The geometry of the sources and shields are described by suitable intersection of quadratic surfaces.

ISOSHLD, SPAN-4, and Rockwell's method approximate scattering effects in the shields by using appropriate build-up factors. Concrete gamma scattering problems are treated through use of the Chilton-Huddleston dose-albedo method(11). This method finds use, for example, in calculating scattered doses over partial height walls and through cubicle doorways.

None of the above-mentioned methods considers the energy degradation ("softening") of the source energy spectrum as the radiation penetrates the shield, and thus each predicts conservative values of the dose rate at the point of interest.

Whenever neutrons are involved, or whenever scattering effects are expected to be important, more advanced, transport theory codes are used. For example, the shielding design for the neutron streaming shield configuration (see Subsection 12.3.1.4) in the vicinity of the reactor 12.3-8 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 vessel employed these state-of-the-art methods. Output from the two dimensional, discrete-ordinates DOT transport code was converted into input for the MORSE Monte-Carlo transport Code by the DOMINO program. The MORSE code solves the neutron or gamma ray transport problem in realistic geometries by following a sufficient number of particle random walk flight paths through the system. Various sampling and biasing techniques are used to reduce the number of histories required to produce results of a given accuracy. The combination of DOT-DOMINO-MORSE allows the accurate calculation of primary source neutron and gamma radiation transport, as well as secondary gamma production and transport.

Comparison of the measured dose rates at operating plants, both BWR and PWR, with corresponding calculated dose rates, indicate that the models and method of calculation used predict higher dose rates than actually observed. Therefore, shielding calculations based on such models and methods are conservative.

12.3.2.4 Compliance with Regulatory Guide 1.69 Regulatory Guide 1.69, "Concrete Radiation Shields for Nuclear Power Plants," December, 1973 (R0), generally invokes ANSI Standard N101.6-1972, "Concrete Radiation Shields," as an acceptable method to the NRC for the design of concrete radiation shields for nuclear power plants. St. Lucie Unit 2 complies with the intent of this guide with the following clarifications:

ANSI N101.6-72 Section Clarifications 4.3.1 Concrete shielding pertains to gamma and/or neutron shielding only. There are no significant sources of alpha or beta radiation within the plant which could affect concrete shield design. The maximum temperature of the primary shield wall is 150 F. This wall is designed to afford required shielding at this temperature 4.3.2 There is no significant heat from high nuclear radiation fluxes.

4.3.4 "The possibility of an explosion in the cell" is precluded by design (see Section 11.3).

4.3.5 Applicable assumptions and methods used for accident analyses are those given in Chapter 15.

4.3.6 Regulatory Guide 8.8 (R3) is used as guidance in limiting personnel exposure and in determining shielding practices.

4.7 No design drawings are prepared specifically for formwork.

Concrete design drawings are provided in sufficient detail to allow proper design of formwork according to good construction practice. Formwork specifications are provided which require conformance to ACI-347.

4.8 No heavy aggregates are used.

5.1.2 No high density concrete is used.

5.1.3 No hydrous aggregate is used.

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UFSAR/St. Lucie - 2 ANSI N101.6-72 Section Clarifications (Contd)

(Contd) 5.1.4 No boron containing aggregates are used.

5.1.6 Coatings of clay, silt, gypsum, calcite or caliche on coarse aggregate total no more than three and one half percent of the total weight of the aggregate. Radiation attenuation calculations take this into account.

5.3.4 No pozzolans are used.

5.3.5 No grout fluidifiers are used.

5.4 A maximum slump of four in. is permitted for certain applications where less sump is impracticable.

5.4.2 The pre-planned-aggregate (PA) method is not used.

5.4.4 Heavy aggregates are not used.

6.1 Formwork for shielding is consistent with good construction practice and as required by ACI-347.

6.2.1 ACI-347 is used for the design of formwork.

6.2.2 Approval of concrete forms prior to construction is per ACI-347.

6.4 A detailed thermal stress analysis is made.

6.5 See position for Section 4.7 with regard to shop drawings.

7.2 See position of Section 4.7 regarding shop drawings. Any changes in specifications must be reviewed and approved prior to construction activity. The effects of supplemental tracing on shield adequacy is evaluated at that time.

8.1.3 No high-density concrete is used.

8.1.8 Aggregate is from one source and is continually sampled throughout the construction phase for conformance to project specifications. Considering these controls, bagging and retention of samples is not necessary.

8.2.6 Sufficient spare vibrators are maintained but not necessarily one for every two being used.

8.4 The puddling method is not used.

12.3-10 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 ANSI N101.6-72 Section Clarifications (Contd)

(Contd) 8.6.1 The composition and fluidity of the mortar or grout, when used in pressure grouting, is specified in project specifications.

8.6.2 Filling of forms is done in accordance with good construction practice.

Specifications require that no voids be left in the concrete.

8.7.1 Relocation of and additional construction joints not shown on drawings require the approval of the responsible engineer.

Construction joints are provided with 1/3 or 1/2 keys. Neat cement paste or grout coating of construction joints is a construction option and not a specification requirement.

8.7.2 Concrete is cured for the specified times. The requirements of ACI-347 are not met regarding time limits for removing forms.

8.7.5 Patching and finishing is performed as soon as practicable to ensure a quality product; however, not necessarily within the specified times.

8.7.6 Traffic or other operations is restricted after curing and finishing to prevent damage to the concrete; but not necessarily for the time specified.

9.1 Areas within the steel containment vessel and other significant areas subject to potential contamination by radioactive substances throughout the plant have a tested and qualified protective coating. ANSI Standards N5.12 and N101.2 are utilized for screening and qualifying the coating materials and systems.

Actual application is conducted in accordance with the product manufacturer recommendations. Quality assurance conforms with ANSI Standard N101.4 (see Subsection 6.1.2).

10.1.2 Dimensional tolerances for hatches and openings as specified in ACI-347 are used rather than those given in Table 1 of ANSI N101.6-72. Minimum practicable joint clearances are specified.

10.1.3 Service trenches are not used.

10.2.2 The weight of each block is indicated on the design drawing, not marked on the block.

12.3-11 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 ANSI N101.6-72 Section Clarifications (Contd)

(Contd) 10.2.3 Blocks are cured according to good construction practice, e.g.,

use of wet burlap or curing compound, but not necessarily in the absence of direct sunlight or heat. This sunlight or heat, however, does not result in the loss of shielding efficiency.

10.3.1 There are no present plans for penetrations through shielding plugs. However, if they are required, streaming would be prevented by proper design of the penetration.

10.4 No movable or removable poured walls are used.

10.6 Precast shielding components are fabricated at the site.

11.5.1 Pre-operational tests of shielding are not performed. Normal post-operational tests identify any areas where additional precautions or shielding are necessary.

11.5.2 Containment leak testing is performed in accordance with 10 CFR 50 Appendix J.

12.3.3 VENTILATION Plant heating, ventilation and air conditioning (HVAC) is designed to provide suitable environment for personnel and equipment during normal plant operation and to provide a safe environment for operating personnel and the public during design basis accident conditions when bringing the plant to a safe shutdown condition.

12.3.3.1 Design Objectives Plant HVAC for normal plant operation and design basis accident conditions is designed to meet the requirements of 10 CFR 20, 10 CFR 50 and guidelines established for design basis accidents.

Design criteria for the plant HVAC include the following:

a. During normal operation the maximum airborne radioactive material concentrations in air inhaled by personnel in restricted areas of the plant must be as low as is reasonably achievable and within the limits specified in Appendix B, Table 1 of 10 CFR 20.

The maximum airborne radioactive material concentrations in unrestricted areas of the plant must be within the limits specified in Appendix B, Table II of 10 CFR 20.

b. During normal operation and design basis accident conditions, the dose from concentrations of airborne radioactive material in unrestricted areas beyond the 12.3-12 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 site boundary are as low as is reasonably achievable and within the limits specified in 10 CFR 20 and guidelines established for design basis accidents, respectively.

c. The dose guidelines established for design basis accidents are satisfied following postulated design basis accidents.

d. The dose to main control room personnel does not exceed the limits specified in GDC 19 of Appendix A to 10 CFR 50.

e. Airborne radioactivity monitoring is provided in compliance with GDCs 63 and 64 of Appendix A to 10 CFR 50.

In the design of the ventilation systems, the following guidelines are used:

a. The airflow is directed from areas of lesser potential contamination to areas of greater potential contamination.

b. Airborne radiation monitoring is provided (see Subsection 12.3.4).

c. Ventilation systems are provided with back draft dampers and isolating dampers, where applicable, to allow servicing of redundant equipment without discontinuing system operation.

d. Ventilation fans and filters are provided with adequate space around the units to allow servicing and replacement of sections.

e. Access control and traffic patterns are considered in the basic plant layout to minimize the spread of contamination. In addition, the following concepts are used to minimize the spread of contamination:

3. Potentially radioactive equipment drains are piped directly to the collection device connected to the collections system thus preventing the spread of contamination.

4. Welded ductwork is employed on potentially contaminated systems to the extent possible to reduce system leakage to a minimum acceptable level.

f. Charcoal filters containing radioactivity are remotely changed by a pneumatic system and do not create additional radiation hazard to personnel in normally occupied areas.

12.3.3.2 Design Description The ventilation systems are described for all plant buildings in Section 9.4. The aspects of the design that relate to removal of airborne radioactivity from equipment rooms, corridors and normally occupied areas are discussed in Subsections 6.5.1, 6.5.3 and 11.3.2.

12.3.3.3 Air Cleaning System Design Air cleaning systems are either safety related fission product removal systems which operate following a design basis accident or nonsafety related systems which control airborne 12.3-13 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 radioactivity in normally occupied areas during normal operation. The normal ventilation system of the Fuel Handling Building (FHB) is an example of a non-safety related air cleaning system which functions during normal operation (refer to Subsection 9.4.2).

An example layout of the normal FHB ventilation system showing filter mountings, access doors, aisle space, service galleries and provision for testing, isolation and decontamination is provided in Figure 12.3-13. As shown, removal of related ventilation equipment for maintenance is facilitated by a removable building roof slab.

Periodic testing for all filters and adsorbers will be performed after initial operation. The frequency of changeout of filters and adsorbers, where applicable, will be determined from periodic testing.

12.3.3.4 Ventilation Systems Compliance to Regulatory Guides The engineered safety features ventilation systems with atmosphere cleanup units meet the intent of Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plant," March 1978 (R2) as described in St. Lucie Unit 2 Subsection 6.5.1. Conformance to Regulatory Guide 1.52 (R2) is delineated in Table 6.5-1.

Conformance to Regulatory Guide 1.140, "Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," October 1979 (R1) is delineated in Table 9.4-16.

12.3.4 AREA RADIATION AND AIRBORNE RADIOACTIVITY MONITORING INSTRUMENTATION The Radiation Monitoring System consists of the following:

a. Area Radiation Monitoring System

b. Airborne Radiation Monitoring System

c. Process and Effluent Radiological monitoring The Radiation Monitoring System is a redundant digital computer based system consisting of various monitors located throughout the plant. The system is described in Subsection 11.5.2.1.1 and the system block diagram is shown in Figure 11.5-1. The Process and Effluent Radiological Monitoring System is discussed in Section 11.5.

The Area Radiation and Airborne Radioactivity monitoring instrumentation follows the recommendations of Regulatory Guide 8.8 (R3) relevant in that:

a. Deleted

b. Area monitors have been placed for optimum coverage

c. Component failure is indicated

d. Monitors have local and control room readout and alarm

e. Monitors have clear readout 12.3-14 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2

f. Monitors have a five decade readout range as a minimum

g. Readings are recorded in the computer.

The instrumentation is capable of supporting an administrative program as described in Regulatory Guide 8.2, "Administrative Practices in Radiation Monitoring," February, 1973 (R0) by providing background and airborne monitoring per ANSI N13.2-1969.

Regulatory Guide 1.21 is addressed in Section 11.5.

12.3.4.1 Area Radiation Monitoring System 12.3.4.1.1 Design Objectives The objectives of the Area Radiation Monitoring System during normal operating plant conditions and anticipated operational occurrences are:

a. to measure and to indicate to operations personnel the ambient gamma radiation in specific areas of the plant,

b. to annunciate and warn of abnormal radiation levels in specific areas of the plant,

c. to furnish records of radiation levels in specific areas of the plant,

d. to provide base data in controlling access of personnel to radiation areas,

e. to warn of uncontrolled or inadvertent movement of radioactive material in the plant,

f. to provide local indication and alarms at key points where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area, and

g. to annunciate and warn of possible equipment malfunctions in specific areas of the plant.

The objectives of the Area Radiation Monitoring System during postulated accidents are:

a. to provide the capability to alarm and initiate a containment isolation signal in the unlikely event of high radiation inside the containment,

b. to provide long term post-accident monitoring of conditions inside the containment,

c. to provide a high radiation signal to isolate the Fuel Handling Building Ventilation System and thus divert the air to the Shield Building Ventilation System, and

d. to assist operations personnel in decisions on deployment of personnel in the event of an accident or equipment malfunction resulting in a release of radioactive material in the plant.

For a discussion of post-accident monitoring instrumentation, see Section 7.5.

12.3-15 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3.4.1.2 Criteria for Location of Monitors Considerations for area monitor locations are based on the following:

a. frequency and length of personnel occupancy of specific area,

b. potential for presence of abnormally high radiation doses,

c. potential for equipment malfunction, accidental radiological release, or postulated accidents, and

d. in normally accessible areas where during normal plant operation including refueling, radiation exposure could exceed the radiation limits due to system failure or personnel error.

12.3.4.1.3 System Description - Area Monitoring The Area Monitoring System is an integral part of the Radiation Monitoring System described in Section 11.5. The Area Radiation Monitoring System contains channels that are located at selected places inside the plant to detect and store information on the radiation levels and, if necessary, annunciate abnormal radiation conditions. The areas where the gamma monitors are located are shown in Table 12.3-2. Indication and annunciation for all channels is provided at each of the operator consoles. Records of data and alarms are temporarily stored by the system's computer. The instrument locations, sensitivities and alarm setpoints are also shown in Table 12.3-2.

Those channels identified in Table 12.3-2 as being safety related indicate and record on digital ratemeters and recorders on seismic Category I, Class 1E radiation monitoring panels in the control room. The safety related channels and safety panels are designed to remain functional during and after a safe shutdown earthquake. Through a qualified isolation buffer, the information signals are transmitted to the non-safety computer storage and display consoles.

A typical channel consists of a detector, microprocessor, power supply, local indicator with audio- visual alarm, and a check source.

The detectors are wall mounted beta-gamma or gamma sensitive Geiger-Muller tubes. Their energy dependence is flat (within +/- 15 percent) from 80 Kev to 2.5 Mev. The monitor, and its response, are not affected by the environmental conditions of temperature (ranging from 30 F to 130 F) pressure (+/-) 1 psi from atmospheric) and relative humidity (from 0-100 percent). The monitor performance is not degraded by power supply voltage variations of +/- 10 percent. In the unlikely event of radiation fields in excess of the maximum range of the detector, the maximum signal output of the detector is maintained until the abnormal high radiation level has subsided.

The detectors are not affected by radiation doses of up to 105 rads, and the microprocessors and other electronic components are not affected by a radiation dose of up to 103 rads.

Each area monitor is provided with a dedicated microprocessor through which all channel information is processed. The microprocessor handles the communication between the detector and the local and remote controls, performs various computations, and stores data.

Each area radiation monitor is provided with a low activity Cs-137 check source, attached to the detector assembly in a solenoid operated container. This check source is used as a convenient 12.3-16 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 operational and gross calibration check of the detector and readout equipment. The check source arrangement can be operated either by the local or remote controls.

Six of the area radiation detectors, RD-26-3, RD-26-4, RD-26-5, RD-26-6, RD-26-38, and RD-26-39, may be equipped with a 32 microcurie Cs-137 keep-alive source (corresponding to a dose rate of 15 millirem per hour, nominal) which provides a live-zero for the associated radiation indicating monitors (RIMs). Thus, when the background radiation in the area of the above mentioned detectors falls below that of the keep-alive source, the monitors will indicate an activity equivalent to that of the source, 15 millirem per hour. This eliminates the possibility of the monitors falsely indicating detector error following a detector test using the internal check sources which will result when the activity level drops to a background level below the monitors minimum detection level.

Each monitor channel is provided with three alarms. One alarm level is set high enough above the normal measured radiation levels in the area to prevent spurious alarms, yet low enough to indicate transient radiation level increased. A second alarm is set at a higher level. A third alarm acts as a detector failure, circuit failure or cable disconnect alarm. Alarm points are adjustable and are varied to suit background or plant conditions. Plant health physics personnel specify the alarm points consistent with radiological safety controls for the area. Alarms annunciate locally and at the operator consoles. In addition, the safety monitors annunciate on the seismic safety radiation monitoring panel in the control room. The local alarm consists of both audio and visual signals.

Each safety related area radiation monitoring channel is provided with an independent power supply, designed such that a failure in that channel does not affect any other channel.

Monitoring channels that are identified as safety related are redundant and are supplied power from the station 120 V ac instrument power supply system safety related buses.

The monitors are calibrated and maintained on a routine schedule. As a minimum, the monitors are calibrated during refuelings as discussed in Technical Specification Table 4.3-3 and after any maintenance work is performed on the detector. For calibration purposes, the built-in check source and a portable or built-in signal generator are used.

12.3.4.1.4 Safety Related Area Monitors Area monitors indicated as safety related in Table 12.3-2 are designed to seismic Category I, Class 1E requirements. The monitors are physically and electrically separated from each other in accordance with the criteria set forth in IEEE 279-1971 and IEEE-308 1971. The monitors are qualified in accordance with IEEE 323-1974 and IEEE 344-1975. Refer to Sections 3.10 and 3.11 for further discussion on qualification of Class 1E equipment.

Those monitors whose main function is to measure radiation following a design basis accident meet the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 (R2). The instruments are qualified in accordance with categories 1 and 2 in Regulatory Guide 1.97.

The containment isolation actuation signal (CIAS) monitors consist of four separate gamma-sensitive ion chambers located within the containment at 90 degree intervals along the containment vessel wall. These monitors initiate the CIAS on high radiation. The monitors are 12.3-17 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 fed from four Class 1E instrument power supply system buses (MA, MB, MC and MD). These four detectors, RD-26-3, RD-26-4, RD-26-5 and RD-26-6, are shown on Figure 12.3-13a.

Six spent fuel pool monitors are provided around the spent fuel pool to detect radioactivity in the event of a fuel handling accident in the Fuel Handling Building. A high radiation signal isolatesthe Fuel Handling Ventilation System and diverts the air to the Shield Building Ventilation System.

Two high range post-accident monitors with a maximum range of 108 R/hr (gamma) are located inside the containment. These monitors are widely separated so as to provide reasonable assessment of area radiation conditions inside containment and are powered from an independent Class 1E power supply. Table 12.3-2 provides a tabulation of the basic design description for these monitors. Figure 12.3-13a shows the location of the radiation monitors inside containment. These monitors are gamma sensitive ionization chambers capable of detecting photons with an energy range of 60 KeV to 3 MeV, with a linear energy response of +/-

20%. The detectors are seismic Category I and qualified to normal operating and post-accident environmental conditions inside containment and have a total integrated life of 109 Rads. A self testing radiation source with a continuous reading of 1 R/hr (approximate value, each detector unique) is provided within the radiation monitors for checking the operational availability of the monitors. Also there are two additional safety-related radiation monitors (gamma sensitive GM detectors) located outside the containment which could be referenced by the plant operators in case the redundant displays of the high range in containment monitors disagree.

12.3.4.1.5 New Fuel Storage Area Radiation Monitors On August 14, 1997, FP&L was granted an exemption from the requirements of 10 CFR 70.24 (References 13 & 14) concerning criticality accident monitors. The result of this exemption made the New Fuel Storage Area Monitor (RD-26-28) valid for the intended application. This exemption was superceded when St. Lucie Unit 2 elected to comply with a new regulation, 10 CFR 50.68 (see Section 12.3.4.1.6).

12.3.4.1.6 Criticality Monitoring In December of 1998 the NRC issued 10 CFR 50.68, Criticality Accident Requirements, a new regulation which gave licensees the option of complying with the requirements of 10 CFR 70.24 or 10 CFR 50.68(b). St. Lucie Unit 2 elected to forgo an existing exemption to 10 CFR 70.24 and to comply with 10 CFR 50.68(b) (Reference 15). With respect to radiation monitoring, 10 CFR 50.68(b) states that "radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions."

12.3.4.2 Airborne Radiation Monitoring System 12.3.4.2.1 Design Objectives The objectives of the Airborne Radiation Monitoring System during normal operating plant conditions and anticipated operational occurrences are:

a. To inform operations personnel of airborne particulate, gaseous and iodine activity in the various buildings and structures of the plant, 12.3-18 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2

b. To alarm any abnormal increases in the airborne activity levels,

c. To furnish records of gross airborne trends in the various plant areas and of the amount of radioactive releases to the environment through the plant buildings or structures during normal, or abnormal operational occurrences,

d. To help detect identified or unidentified leaks inside the reactor coolant pressure boundary (as recommended in Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems", May 1973 (R0) and other areas of the plant,

e. To assist personnel in deciding whether or not breathing apparatus is necessary when entering a high activity area, and

f. To provide information for evaluation of the performance of all plant systems that function to minimize the release of radioactivity to accessible areas of the plant and to the environment.

The objectives of the Airborne Radiation Monitoring System during postulated accidents are:

a. To provide the capability to alarm, initiate isolation of the control room normal ventilation system, and actuate emergency ventilation.

b. To provide control room operators with information regarding airborne radiation in the two control room outside air intakes.

Those monitors whose main function is to measure radiation levels following a design basis accident are seismically qualified, Class 1E equipment. For a discussion of post-accident monitoring instrumentation, see Section 7.5.

12.3.4.2.2 Criteria for Location of Monitors Considerations for locating the Airborne Radiation Monitoring System monitors are based on the following:

a. Areas where the airborne radioactivity can abruptly increase and where personnel normally have access to the areas,

b. Inside the containment for the purpose of monitoring unidentified leaks, and

c. In control room outside air intake ducts for post-accident habitability monitoring purposes.

12.3.4.2.3 System Description - Airborne Monitors The Airborne Monitoring System is an integral part of the radiation monitoring system described in Subsection 11.5.2.1.1. Those airborne monitors whose function is to measure radioactive effluents in accordance with Regulatory Guide 1.21 are described in Section 11.5.

The Airborne Radiation Monitoring System provides information, both locally and in the control room, for the purpose of maintaining low in-plant personnel radiation exposure via inhalation of airborne particulates and iodine, in accordance with 10 CFR 20 and Regulatory Guide 8.8 (R3).

12.3-19 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 It also provides a signal to initiate isolation of the control room ventilation system. The Airborne Radiation Monitoring System consists of monitors for the containment atmosphere, control room, and ECCS area vents.

Annunciation and indication is provided for the channels at each of the operator consoles.

Records of data and alarms are temporarily stored by the system's computer. The instrument locations, ranges, sensitivities and alarm set points are addressed in Table 12.3-3.

Those channels identified in Table 12.3-3 as being safety related indicate and record on ratemeters and recorders on a seismic Category I safety radiation monitoring panel in the control room. The safety related channels and control panels are designed to remain functional during and after a safe shutdown earthquake. Through a qualified digital isolation device, the information signals are transmitted to the non-safety computer storage and display consoles.

A typical airborne monitor consists of an off-line sampler assembly, a three-stage gas, particulate and iodine detector, a microprocessor, a power supply, and a local indicator and alarm.

Airborne monitors except the control room outside air intake monitors are housed in off-line shielded sampler assemblies. Samples are continuously withdrawn from ventilation ducts or the ambient atmosphere to the shielded sampler assemblies. The sampler assembly contains the radiation detection equipment, pumps, valves, and check sources. The sample chamber is sized and shielded in a 4 geometry as required to achieve the specified minimum system sensitivities. Samplers are designed so that they have air purge for the noble gas chamber.

Airborne monitors are provided with a solenoid operated check source that simulates a radioactive sample in the detector, and is used for operational and gross calibration checks of the detector and readout equipment.

Each sampler assembly is provided with a dedicated microprocessor through which information is processed. The microprocessor handles communication between detectors and local and remote controls, performs various computations, and stores data; a sampler assembly consisting of more than one detector (channel) i.e., particulate, iodine, gaseous, has one microprocessor.

Monitors are provided with three alarms. Two upscale trips are provided to indicate high radiation levels and one downscale trip to indicate instrument trouble. The alarm points are adjustable and can be varied to suit background or plant conditions. Plant health physics personnel specify the alarm points consistent with the radiological safety function of the detector. Alarms annunciate locally and at the operator consoles. In addition, the safety monitors annunciate on the seismic Category I, Class 1E radiation monitoring panel in the control room.

An airborne radiation monitoring channel is provided with an independent power supply, such that a failure in that channel can not affect any other channel. Channels that are identified as safety related are redundant and are supplied power from the station 120V ac safety related buses.

The monitors are calibrated and maintained on a routine schedule. As a minimum, the monitors are calibrated in accordance with Technical Specifications and after any maintenance work is performed on the detector. For calibration, a standard source, the built-in check source, and a portable or built-in signal generator can be used.

12.3-20 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 The Airborne Radiation Monitoring System follows the general guidance of ANSI 13.1-1969 except that isokinetic sample nozzles are provided only on airborne effluent monitors, as described in Section 11.5.

12.3.4.2.3.1 Containment Atmosphere Radiation Monitors The containment atmosphere radiation monitors are designed to provide a continuous indication in the control room of the particulate, iodine and gaseous radioactivity levels inside the containment. Radioactivity in the containment atmosphere indicates the presence of fission products due to a reactor coolant pressure boundary leak. As discussed in Subsection 5.2.5, continuous monitoring allows unidentified leakage in the containment to be detected in accordance with the intent of Regulatory Guide 1.45 (R0).

Two redundant seismically qualified containment atmosphere radiation monitors are provided.

These monitors are designed to seismic Category I and Class 1E requirements and are qualified in accordance with IEEE 279-1971, IEEE 308-1971, IEEE 323-1974, and IEEE 344-1975.

Monitors are three-stage, gas, particulate and iodine monitors as described in Subsection 11.5.2.1.3C and as shown on Figure 11.5-4 with the addition of redundant pumps. The sampler assembly shown in Figure 12.3-14 extracts air continuously from the Reactor area sampling point. In the event that radioactivity is detected, the leak may be located by sequencing through the sampling points until the highest concentration is found. The sample line temperature is maintained above the dew point upstream of the monitor to eliminate condensation and subsequent particulate losses in the sample line.

This sample system is isolated upon CIAS.

12.3.4.2.3.2 Control Room Air Intake Monitors Redundant (train A and train B) radiation monitors are provided in the two control room air intakes, each monitor with its own dedicated micro-processor. The monitors consist of a pair of beta-gamma scintillation detectors; one shielded against beta radiation, one unshielded. This arrangement provides active background subtraction to allow for changes in the gamma ray background.

Upon detection of high radiation by the monitors, a signal is developed to isolate the main control room and initiate the Control Room Emergency Cleanup System fans. By comparing the readings from the two air intakes, the operator can determine which side of the plant has the lower airborne radiation level and thus allow outside air with the lowest airborne radiation to be drawn in.

These monitors are designed to seismic Category I and Class 1E requirements and are qualified in accordance with IEEE 279-1971, IEEE 308-1971, IEEE 323-1974, and IEEE 344-1975.

The monitors have a solenoid actuated Cs-137 check source. No local alarm or indication is provided.

12.3-21 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3.4.2.3.3 ECCS Area Vent Monitors Two seismically qualified monitors are provided to measure the noble gas effluent from the ECCS areas.

These monitors are designed to seismic Category I and Class 1E requirements and are qualified in accordance with IEEE 279-1971, IEEE 308-1971, IEEE 323-1974, and IEEE 344-1975.

A sample is withdrawn from the ECCS area emergency vents to an off line multistage gas monitor as described in Subsection 11.5.2.1.3d and as shown on Figure 11.5-4. The monitors measure airborne activity and continuously sample particulates and iodine originating from the ECCS area during accident conditions.

12.3-22 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 SECTION 12.3: REFERENCES

1. W. A. Rhoades, F.R Mynatt, "The DOT III Two-Dimensional Discrete Ordinates Transport Code," ORNL-TM-4280 (1973). Contained in code package CCC-276 from Radiation Shielding Information Center, Oak Ridge National Laboratory.

2. E. A. Straker, et al, "The MORSE Code - A Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code,"ORNL-TM-4585 (1970). Contained in code package CCC-203 from Radiation Shielding Information Center, Oak Ridge National Laboratory.

3. D. Garber, C. Dunford, S. Pearlstein, "Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF, "BNL-NCS-50496, (ENDF-102) (1975).

4. C. Lederer, J. Hollander, I. Perlman, Table of Isotopes, 6th Edition, John Wiley & Sons, Inc. (1967).

5. American National Standard - Neutron and Gamma-Ray Flux-to-Dose Rate Factors,"

ANSI/ANS-6.1.1-1977 (N666).

6. Reactor Physics Constants, ANL-5800

7. T. Rockwell, III (ed.), Reactor Shielding Design Manual, TID-7004 (1956).

8. N. M. Schaeffer, Reactor Shielding for Nuclear Engineers, TID-25951 (1973)

9. R. L. Engel, et al, "ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis," BNWL-236 (1066). Contained in Code package CCC-79 from Radiation Shielding Information Center, Oak Ridge National Laboratory.

10. O.J. Wallace, "SPAN-4, A Point Kernel Computer Program for Shielding," WAPD-TM-809(L) (1972).

11. A. Chilton and C. Huddleston, "Semi Empirical Formula for Differential Dose Albedo for Gamma Rays on Concrete", Nuclear Science and Engineering, Volume 17, page 419, 1963.

12. M. B. Emmett, C. E. Burgart, T. J. Hoffman, "DOMINO, A General Purpose Code for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations, "ORNL-4853 (1973). Contained in code package PSR-64 from Radiation Shielding Information Center, Oak Ridge National Laboratory.

13.. NRC letter 8/14/97, L. A. Weins to T. F. Punkett, "ISSUANCE OF EXEMPTION FROM THE REQUIREMENT OF 10 CFR 70.24 ST NRC letter 8/14/97, L. A. Weins to T. F.

Punkett, "ISSUANCE OF EXEMPTION FROM THE REQUIREMENT OF 10 CFR 70.24 ST LUCIE UNITS 1 AND 2".

14. Letter to NRC from J. A. Stall, St. Lucie Plant, to Document Control Desk, U.S. NRC, L-97-171;10 CFR 50.4, 10 CFR 70.14, and 10 CFR 70.24, July 10, 1997.

15. St. Lucie Plant Manager Action Item (PMAI) PM98-11-093.

12.3-23 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3-1 ALLOWABLE DOSE RATES Max Whole Body Location Dose Rate (mrem/hr) a) Site Boundary Normal Operation 0.001 b) Service Building Normal Operation 0.05 c) Turbine Building Normal Operation 0.05 d) Reactor Auxiliary & Fuel Handling Buildings Continuous Occupancy Outside Controlled Access Areas 0.25 Inside Controlled Access Areas 2.5 Controlled Occupancy Occupancy for 6 hr/wk 15.0 Occupancy determined by Health Physics Staff 100.0 e) Limited Access Areas in Contain-ment Structure during operation at full Power 100.0 f) Control Room Normal Operation 0.25 Following DBA 5 rem integrated whole body dose over 30 days after accident.

T12.3-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 Table 12.3-2 AREA RADIATION MONITORS Detector Tag Safety Range(2) Sensitivity Accuracy(2) Typical Alarm(3) Monitor Detector Item Number Monitored Area Classification (mR/hr) (mR/hr) (%) Setpoint (mR/hr) Location Location 1 RD-26-1 Control Room Non-Safety 10-2 1.0 RAB el 67.5' RAB el 67.5 2 RD-26-3 CIAS A 1E 101 104 RAB el 25' Containment el 90' 3 RD-26-4 CIAS B 1E 101 104 RAB el 25' Containment el 90' 4 RD-26-5 CIAS C 1E 101 104 RAB el 25' Containment el 90' 5 RD-26-6 CIAS D 1E 101 104 RAB el 25' Containment el 90' 6 RD-26-7 Spent Fuel Pool A 1E 10-1 10 FHB el 68' FHB el 61' 7 RD-26-8 Spent Fuel Pool B 1E 10-1 10 FHB el 68' FHB el 61' 8 RD-26-9 Spent Fuel Pool C 1E 10-1 10 FHB el 68' FHB el 61' 9 RD-26-10 Spent Fuel Pool D 1E 10-1 10 FHB el 68' FHB el 61' 10 RD-26-11 Spent Fuel Pool E 1E 10-1 10 FHB el 68' FHB el 61' 11 RD-26-12 Spent Fuel Pool F 1E 10-1 10 FHB el 68' FHB el 61' 12 RD-26-38 Containment Post Accident A 1E 101 30 RAB el 64.5 Below FHB roof at el 13 RD-26-39 Containment Post Accident B 1E 101 30 RAB el 67.5 Above RAB roof at el 14 RD-26-32 Personnel Lock Area Non-Safety 10-1 10 Containment Containment el 35.5' el 35.5' 15 RD-26-33 Refueling Canal Area Non-Safety 10-1 10 FHB el 68' FHB el 67.5' 16 RD-26-34 Fuel Pool Pump Area Non-Safety 10-1 10 FHB el 25.5' FHB el 25' 17 RD-26-35 Boric Acid Precon- Non-Safety 10-1 10 RAB el 5' RAB el 5' centrator Filter Area 18 RD-26-36 Waste Filter Area Non-Safety 10-1 80 RAB el 5' RAB el 5' 19 RD-26-37 ECCS Equipment Area Non-Safety 10-1 10 RAB el 10' RAB el 5.5' 20 RD-26-13 Waste Gas Compressor Area Non-Safety 10-1 40 RAB el 5.5' RAB el 9.5' 21 RD-26-14 Charging Pump Area Non-Safety 10-1 40 RAB el 5.5' RAB el 10.5' 22 RD-26-15 Holdup Drain Pump Area Non-Safety 10-1 20 RAB el 4.5' RAB el 4.5' 23 RD-26-16 Sample Room Area Non-Safety 10-2 5 RAB el 25' RAB el 26' T12.3-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3-2 (Cont'd)

Detector Tag Safety Range(2) Sensitivity Accuracy(2) Typical Alarm Monitor(3) Detector Item Number Monitored Area Classification (mR/hr) (mR/hr) (%) Setpoint (mR/hr) Location Location 24 RD-26-17 Ion Exchanger Corridor Non-Safety 10-1 80 RAB el 25' RAB el 26'

-1 25 RD-26-18 Flash Tank Pumps Corridor Non-Safety 10 20 RAB el 25' RAB el 26' 26 RD-26-20 Purification Filter Area Non-Safety 10-1 20 RAB el 25' RAB el 35.5' 27 RD-26-21 Spent Resin Corridor Area Non-Safety 10-1 20 RAB el 5' RAB el 6.5' 28 RD-26-22 ECCS Equipment Area Non-Safety 10-1 20 RAB el 5' RAB el 4' 29 RD-26-23 Decontamination Area Non-Safety 10-1 10 RAB el 5' RAB el 26'

-1 30 RD-26-24 HVAC Room Area Non-Safety 10 10 RAB el 48' RAB el 49' 31 RD-26-25 Chemical Drain Pump Area Non-Safety 10-1 20 RAB el 5' RAB el 12' 32 RD-26-26 Volume Control Tank Corridor Non-Safety 10-1 20 RAB el 25' RAB el 26' 33 RD-26-27 Boronometer Enclosure Area Non-Safety 10-1 10 RAB el 26' RAB el 26' 34 RD-26-28 New Fuel Storage Area Non-Safety 10-1 10 FHB el 67.5' FHB el 53.5'

-1 35 RD-26-29 Aerated Waste Storage Area Non-Safety 10 10 RAB el 5' RAB el 5' 36 RD-26-30 Boric Acid Concentrator Area Non-Safety 10-1 10 RAB el 25' RAB el 28' 37 RD-26-31 Fuel Pool Filter Area Non-Safety 10-1 20 FHB el 25' FHB el 26' 38 RD-26-2 Operating Deck Area Non-Safety 10-1 102 Containment Containment el 35.5' el 85' 39 RD-26-19 Drumming Station Area Non-Safety 10-1 20 RAB el 25' RAB el 25 40 RD-26-40 Containment High Range IE 10-3 104 RAB el 65' RCB el 90' 41 RD-26-41 Containment High Range IE 10-3 104 RAB el 65' RCB el 90' Notes (1) Gamma sensitive detector. Measurement is in Rad/hr (2) Instrument ranges are selected in accordance with standard engineering practices. Instrument accuracies are selected such that existing instrument loop performance and safety analysis assumptions remain valid. Where applicable, instrument accuracies are also evaluated for their impact on setpoints in accordance with the FPL Setpoint Methodology.

(3) Alarm points are adjustable and are varied to suit background or plant conditions. Plant Health Physics personnel specify the alarm points consistent with radiological safety controls for the area.

This information is kept for historical purposes.

T12.3-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3-3 AIRBORNE RADIATION MONITORS Number Monitor/Location of Safety Range (1) Sensitivity Accuracy(1) Typical Alarm (2) of Sample Pt. Type Channels Classification (ci/cc) (ci/cc) (+/-%) Setpoint (ci/cc)

Containment Particulate 2 1E 10-12 10-8 Atmosphere, multiple sample Iodine 2 1E 10-10 10-8 Noble Gas 2 1E 10-7 10-5 Control Room - scint w/ 4 1E 10-8 320 cpm Outside Air Intake bkgrd Subtraction ECCS Area Vents Noble Gas 2 1E 10-7 5 x 10-4 (1) Instrument ranges are selected in accordance with standard engineering practices. Instrument accuracies are selected such that existing instrument loop performance and safety analysis assumptions remain valid. Where applicable, instrument accuracies are also evaluated for their impact on setpoints in accordance with the FPL Setpoint Methodology.

(2) The Typical Alarms Setpoint column is kept for historical purposes. This information is maintained by Plant Procedures.

T12.3-4 Amendment No. 24 (09/17)

DELETED FLORIDAPOWER & LIGHTCOMPANY ST. LUCIEPLANTUNIT2 FIGURE12.3-1 AMENDMENTNO. 18 (01/08)

DELETED FLORIDAPOWER & LIGHTCOMPANY ST. LUCIEPLANTUNIT2 FIGURE12.3-2 Amendment No. 18 (01/08)

Referto Drawing 2998-G-797-SH 20 FLORIDAPOWER & LIGHTCOMPANY ST. LUCIE PLANT UNIT 2 REACTORBLDG MISCELLANEOUS STRUCTURESTEELSH. 20 FIGURE 12.3-3 Amendment No. 18 (01/08)

Withheld Under 10 CFR 2.390 J

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Withheld Under 10 CFR 2.390 AloiiNDMINTII).7~

Fl..ORIDAPOWER & LIGHT COMPANY ST. LUCE PLAMT UNIT 2 RAOIATION ZONES R&AC1'0RCONTAINMENT BUILDING FLOOflELEVATIONS 18'.Zl'AND&2' FIGUR.E 12.3-6

Withheld Under 10 CFR 2.390

--NT-.&~

fLORIDA POWER & LIGHT COMPANY ST. LUCIE PLANT UHJT 2 RAOIATION ZOHES REACTOR CONTAINMENT BUILDING PLOOR ELEVATION 45' FIGURE 12.3-8

Withheld Under 10 CFR 2.390 JIMI-NT NO.I..,_.

FLORIDA POWER & LtGtiT COMI'~Y ST. LUOE PLAMT UHIT 2 RADIATION ZONES REACTOR AUXILIARY BUILDING FLOOR ELEVATION -C.B' FIGURE 12.3-7

Withheld Under 10 CFR 2.390 MlliMIII8fl' NO.7 , _ ,

FLORIDA POWI!R & LIGHT COMPANY ST. WCIE PLAMTUNIT 2 RADIAT10N ZONES IIIACTOR AUXILIARY BUILDING PUJOR I!LI!VATlDN Ul.&'

fiGURE 12.3-8

Withheld Under 10 CFR 2.390

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~NTNO.S~I FLORIDA POWER & LIGHT COMPANY ST. LUCII! PI.AHT UHIT 2 RADIATION ZONES REACTORAUXILIARY BUILDING FLOOR ELEVATION 43' FIGURE 12.3-9

Withheld Under 10 CFR 2.390 TIIO.I 141101 FLORIDA POWER & LIGHT COMPANY ST. LUCIE PLAHTUHIT 2 RADIATIONZONES REACTORAUXILIARY BUILDING FLOOR ELEVATION82' FIGURE12.3-10

Withheld Under 10 CFR 2.390

~

- * - N T110. I 1411111 FLORIDA POWER & LIGHT COMI'AHY ST, LUOE PLANT UNIT l RADIATION ZONES FUEL HANDLING BUILDINC1 FLOORELEIIATIONS 18.6', 48'AHD62' FIGURE 12.3-11

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Withheld Under 10 CFR 2.390 FLORIDA POWER & LIGHT COMPAHY ST. WCIE PLAHT UNIT 2 AADIAllON ZONES FUlLHANDUNG 8UILDING VERT!~ SECTIONS PIOURE12.3-12

Referto Drawing 2998-G-872 FLORIDAPOWER & LIGHTCOMPANY ST. LUCIE PLANT UNIT 2 HVAC-FUELHANDLINGBUILDING PLANS, SECTIONS& DETAILS FIGURE 12.3-13 Amendment No. 18 (01/08)

Withheld Under 10 CFR 2.390 FLORIDAPOWER & LIGHT COMPANY ST. LUCIE PL.A.NTUNIT 2 LOCATIONOF CIS AND POST ACCIDENTRADIATION MONITORS FIGURE 12.3-138

Referto Drawing 2998-G-092-Sh1 FLORIDAPOWER & LIGHTCOMPANY ST. LUCIE PLANT UNIT 2 FLOWDIAGRAMMISCELLANEOUS SAMPLINGSYSTEMS FIGURE 12.3-14 Amendment No. 18 (01/08)

UFSAR/St. Lucie - 2 APPENDIX 12.3A TMI SHIELDING STUDY 12.3A-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 APPENDIX 12.3A TMI SHIELDING STUDY 12.3A.1 INTRODUCTION Following the requirement of NUREG-0737 Item II.B.2 "Plant Shielding", a design review of the St. Lucie Unit 2 plant shielding was performed. This assures safe personnel access to the vital equipment or areas required for mitigation or monitoring of an accident. Equipment qualification to radiation doses resulting from an accident is addressed in Section 3.11.

In compliance with Item II.B.2 of NUREG-0737, radiation source terms are specified, systems assumed to contain high levels of radioactivity as a result of a postulated accident are listed, vital areas requiring access are identified, and dose rates and doses in vital areas are presented. Dose rate zone maps were created to show dose rate images throughout the Reactor Auxiliary Building (RAB) at 1,10,100 and 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> following an accident; they are included as Figures 12.3A-1 to 12.3A-4.

The source term information presented below in Sec. 12.3A.2 and Tables 12.3A-1 through 12.3A-5 represents original licensing basis. Source term scaling factors were used to incorporate the impact of extended power uprate (EPU) on the mission doses. The dose estimates presented in Table 12.3A-7 represent EPU operation. The dose rate zone maps presented in Figures 12.3A-1 to 12.3A-4 are not impacted by the EPU.

12.3A.2 SOURCE TERMS The source terms used in determining dose rates and doses presented in Section 12.3A are consistent with the specifications of NUREG-0737. All source terms are based on the core inventory of nuclides derived for St. Lucie Unit No. 2 from Table 4.3-1 of the Combustion Engineering System 80 Radiation Design Guide.(1) The St. Lucie core inventory, separated for convenience into noble gases, halogens, and other nuclides, is shown in Table 12.3A-1.

Four general sets of multigroup source terms, suitable for input to shielding codes, were created from the core inventory data; two for liquid sources, one for gaseous sources, and one for plateout. The GROUP(2) code was used to transform the isotopic sources into multigroup, energy-dependent gamma sources as functions of time after release from the core.

The two sets of source terms for liquid systems are based on the assumptions of instantaneous release into the reactor coolant of the following percentage of the core inventory:

100% of the noble gases 50% of the halogens 1% of the other nuclides Multigroup source terms calculated under these assumptions for various times after an accident (considering radioactive decay) are presented in Table 12.3A-2. The unit of the source terms is

/sec, which can be converted to a specific source term unit of /(cm3-sec) by dividing by the reactor coolant volume of 2.05 X 108cm3. The resulting set of gamma ray source terms was used in dose rate calculations for systems postulated to contain "undiluted" reactor coolant 12.3A-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 water such as the primary sampling system, which would be required to function in the event of a "small break" LOCA, where the reactor coolant would experience little dilution by nonradioactive water.

Other systems such as the Containment Spray and Safety Injection Systems, would contain radioactive water only after exhausting the supply of nonradioactive water contained in the Refueling Water Tank, and then being switched to the recirculation mode. In that mode, commencing no sooner than 20 minutes after the start of a "large break" LOCA, these systems would draw water from the containment sump located in the Reactor Building. Source terms for recirculated (containment sump) water were created by first eliminating the noble gases from the water, in accordance with the guidelines of NUREG-0737 for recirculated, depressurized water, and then diluting the remaining 50 percent of the core inventory of halogens and one percent of the other nuclides by the combined volumes of the reactor coolant, the Safety Injection Tanks, and the minimum volume of the Refueling Water Tank; the total water volume being approximately 1.62 X 109 cm3. The resulting source terms are listed in Table 12.3A-3.

The gaseous source terms were created using the NUREG-0737 assumption of instantaneous release to the containment atmosphere of the following percentages of the core inventory:

100% of the noble gases 25% of the halogens Time dependent source terms, shown in Table 12.3A-4, resulted from application of appropriate radioactive decay factors, leakage factors, and containment spray removal factors. These source terms were used primarily to obtain dose rates to personnel outside, but in the vicinity of, the containment, and dose rates to personnel in the vicinity of the Hydrogen Analyzers.

The final, general set of source terms was created to model the effect of plateout in the Containment. Accordingly, an instantaneous plateout of 25 percent of the core inventory of iodine was assumed. Time dependent source terms appear in Table 12.3A-5. These source terms were used in determining dose rates to personnel outside the containment.

Several specialized sets of source terms were also created for applications such as determining dose rates from the charcoal adsorbers of the Shield Building Ventilation System, and of the Control Room Emergency Filters.

12.3A.3 RADIOACTIVE SYSTEMS The systems identified as potentially containing high levels of radioactivity in a post accident situation and which were considered in the shielding design review undertaken to assure access to vital areas, are listed in Table 12.3A-6. All other systems, such as the Chemical and Volume Control System (letdown portion is isolated after a LOCA with charging taking suction from the BAM tanks), and the Waste Management System are not necessary for post-LOCA operation.

Degassing of the Reactor Coolant System will be done using the Reactor Head Vent System (see Appendix 1.9A item II.B.1) rather than the letdown portion of the Chemical and Volume Control System, and the Waste Management System will be isolated and, therefore, not employed since radioactive leakages and drains will be routed back into the containment via the ESF Leakage Collection and Return System (see Subsection 9.3.5).

12.3A-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.3A.4 VITAL AREAS REQUIRING OCCUPANCY/ACCESS An extensive review was undertaken to identify vital areas of the plant to which personnel access following an accident must be assured. A list of these areas (with accompanying occupancy, dose rate, and dose information) appears as Table 12.3A-7. Access routes from the control room to the vital areas have been noted on Figures 12.3A-5 to 13.3A-7. No access outside the control room is required for containment isolation reset and instrument panels. The diesel generators are located outside the RAB in a separate building with necessary control and indication provided in the control room. No access is required to the electrical distribution equipment and the Waste Management System post-accident. As indicated in Table 12.3A-7 in cases where the review revealed that high dose rates or accumulated dose would preclude access, means for remote operation, additional shielding or plant modifications were provided.

The result of the review process was to assure that access to vital areas could be accomplished consistent with NUREG-0737 requirements of: (1) less than 15 mrem/hr (averaged over 30 days) for areas requiring continuous occupancy, and GDC-19 requirements of less than five rem for the duration of the accident for areas requiring irregular occupancy.

VITAL AREA ACCESS AFTER EXTENDED POWER UPRATE After implementation of Extended Power Uprate, the two areas that may require operator access post-accident, the Hydrogen Analyzer Cubicle and the ECCS Vent Monitors, will still meet the requirement of GDC-19. Refer to Table 12.3A-7.

12.3A.5 DOSE RATE AND DOSE CALCULATIONS Dose rate calculations were performed in areas identified as vital areas, and along potential access routes. Time-dependent sources were determined as stated in Subsection 12.3A.2 and appropriate geometry factors were applied to pipe and equipment of the systems identified in Subsection 12.3A.3. The shielding effect of the equipment, fluid, and shield wall arrangement were considered. The effect of rebar, embedded plates, or any structural steel was neglected, which, when combined with the very conservative, "worst case" source terms, resulted in very conservative calculated dose rates.

Dose rates were calculated primarily by the ISOSHLD(3)*point-kernel integration code. Radiation dose maps were prepared from the dose rate data, and show dose rate ranges throughout the RAB at 1,10,100 and 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> following a postulated accident. These maps, superimposed on general arrangement drawings are included as Figures 12.3A-1 to 12.3A-4.

The primary volume of the replacement steam generators (RSGs) is greater than that of the original steam generators (AREVA NP Document 77-5069878-01). The calculations documented in this section use a conservatively low RCS volume to determine the source terms. Therefore, these analyses are conservative with respect to RSG primary volume.

An Ebasco version of the code DEV/ISOSHLD was actually used.

12.3A-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 APPENDIX 12.3A: REFERENCES

1. Combustion Engineering, Radiation Design Guide, Rev. 4, SYS80-PE-PG (7/12/79).

2. E. Ochoa, 0. Vories, "GROUP - An Isotopic Source Generation Program." An Ebasco Code (7/10/77).

3. R.L. Engel, et at, "ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis," BNWL-236 (1966). An Ebasco version of the code, DEV/ISOSHLD was actually used.

12.3A-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 0886W-1 TABLE 12.3A-1 CORE INVENTORY (Historical)

Noble Gases Nuclide Ci Nuclide Ci Nuclide Ci Nuclide Ci Kr-85m 1.90+7(a) Kr-89 6.09+7 Xe-133 1.52+8 Xe-138 1.21+8 Kr-85 6.02+5 Kr-90 6.02+7 Xe-135m 3.07+7 Xe-140 6.22+7 Kr-87 3.48+7 Kr-91 4.44+7 Xe-135 2.73+7 Xe-143 1.48+6 Kr-88 4.97+7 Xe-131m 5.30+5 Xe-137 1.34+8 Xe-144 3.31+5 Halogens Nuclide Ci Nuclide Ci Nuclide Ci Nuclide Ci Br-84 1.46+7 Br-89 2.19+7 I-131 7.57+7 I-135 1.41+8 Br-85 1.87+7 Br-90 1.38+7 I-132 1.10+8 I-137 6.33+7 Br-87 3.01+7 I-127 1.19+25(b) I-133 1.52+8 I-138 3.18+7 Br-88 3.18+7 I-129 1.89+0 I-134 1.64+8 Other Nuclides Nuclide Ci Nuclide Ci Nuclide Ci Nuclide Ci Se-84 1.39+7 Nb-95 1.28+8 Te-131m 1.14+7 Cs-137 6.56+6 As-85 2.42+6 Zr-99 1.25+8 Te-131 6.56+7 Ba-137m 6.22+6 Se-85 8.61+6 Nb-99 1.31+8 Sn-132 1.29+7 Cs-138 1.29+8 Se-87 1.38+7 Mo-99 1.38+8 Sb-132 3.63+7 Cs-140 1.17+8 Rb-88 5.05+7 Tc-99m 1.19+8 Te-132 1.08+8 Ba-140 1.32+8 Sr-89 7.01+7 Mo-103 1.21+8 Sn-133 4.49+6 La-140 1.36+8 Rb-90 6.19+7 Tc-103 1.23+8 Sb-133 4.07+7 Cs-143 2.52+7 Sr-90 4.89+6 Ru-103 1.24+8 Te-133m 5.45+7 Ba-143 1.00+8 Y-90 5.13+6 Tc-106 5.11+7 Te-133 8.68+7 La-143 1.13+8 Rb-91 7.97+7 Ru-106 3.50+7 Cs-134 1.43+7 Ce-143 1.14+8 Sr-91 8.61+7 Sn-129 8.04+6 Sb-134 7.26+6 Pr-143 1.12+8 Y-91M 4.96+7 Sb-129 2.50+7 Te-134 1.15+8 Cs-144 7.71+6 Y-91 9.13+7 Te-129m 6.49+6 Sb-135 4.55+6 Ba-144 7.46+7 Sr-95 9.19+7 Te-129 2.37+7 Te-135 5.99+7 La-144 9.84+7 Y-95 1.21+8 Sn-131 2.22+7 Cs-135 1.90+1 Ce-144 9.00+7 Zr-95 1.27+8 Sb-131 6.11+7 Cs-136 4.00+6 Pr-144 9.06+7 (a) Read as 1.90x107 curies (b) I-127 is stable. Number given is total atoms.

T12.3A-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3A-2 UNDILUTED REACTOR COOLANT SOURCE TERMS (Historical)

Effect. Energy Multigroup Gamma Source Terms, /sec, at t = ... After Accident (a) (b)

Grp MeV 0 30 min 1 hr 2 hr 10 hr 1 day 100 hr 7 day 30 day 1000 hr 6 month 1 year 1 .150 9.31+18(c) 5.58+18 5.41+18 5.26+18 4.73+18 4.27+18 2.81+18 1.94+18 1.07+17 3.05+16 4.92+15 3.19+15 2 .250 4.84+18 2.24+18 1.79+18 1.48+18 6.39+17 2.72+17 7.62+16 5.42+16 6.41+15 2.38+15 2.25+13 3.26+12 3 .350 2.41+18 1.94+18 1.70+18 1.44+18 1.15+18 1.07+18 8.06+17 6.28+17 8.66+16 3.18+16 4.23+12 1.60+11 4 .475 1.01+19 4.46+18 3.79+18 3.33+18 2.20+18 1.31+18 1.51+17 5.88+16 2.74+16 2.14+16 1.74+15 1.59+14 5 .650 6.14+18 2.74+18 2.20+18 1.49+18 2.63+17 1.54+17 1.16+17 9.59+16 3.49+16 2.72+16 1.12+16 7.55+15 6 .825 9.67+18 5.96+18 4.44+18 2.61+18 3.86+17 2.06+17 8.67+16 7.30+16 5.00+16 4.23+16 9.31+15 4.23+15 7 1.000 1.84+18 9.86+17 8.88+17 7.43+17 2.57+17 6.72+16 2.48+15 6.11+14 6.90+13 6.57+13 4.97+13 4.15+13 8 1.225 4.52+18 2.17+18 1.86+18 1.45+18 5.92+17 1.93+17 8.55+15 1.97+15 4.68+14 3.19+14 9.67+13 7.31+13 9 1.475 2.83+18 4.96+17 4.30+17 3.34+17 7.28+16 2.63+16 2.19+15 3.92+14 1.79+14 1.77+14 1.56+14 1.31+14 10 1.700 2.33+18 1.40+18 1.08+18 8.13+17 3.20+17 9.59+16 8.63+15 2.67+15 1.96+11 1.57+9 0 0 11 1.900 1.95+17 7.34+16 5.28+16 3.26+16 2.70+15 1.47+14 1.93+13 4.00+12 1.16+7 1.79+4 0 0 12 2.100 9.26+17 3.47+17 1.95+17 1.21+17 1.50+16 4.63+14 3.09+6 0 0 0 0 0 13 2.300 1.19+18 9.30+17 8.20+17 6. 39+17 8.79+16 2.74+15 1. 83+7 0 0 0 0 0 14 2.500 6.32+17 2.26+17 1.84+17 1.28+17 2.95+16 7.38+15 3.07+14 9.44+13 6.94+9 5.55+7 0 0 15 2.700 1.94+16 2.42+15 1.23+15 3.29+14 1.05+10 1.45+2 0 0 0 0 0 0 16 3.000 7.27+17 9.33+14 1.54+12 9.26+6 0 0 0 0 0 0 0 0 17 3.500 1.46+17 6.74+14 8.91+13 1.57+12 0 0 0 0 0 0 0 0 18 4.000 7.11+17 1.83+16 9.50+15 2.58+15 7.40+10 8.40+2 0 0 0 0 0 0 19 5.000 1.77+17 2.62+7 0 0 0 0 0 0 0 0 0 0 (a) 100% core noble gases, 50% halogens, 1% other nuclides (b) To obtain specific source terms /(cm3-sec)), divide by reactor coolant volume of 2.05x108 cm3 (c) Read as 9.31x1018 /sec T12.3A-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3A-3 RECIRCULATED (CONTAINMENT SUMP) WATER SOURCE TERMS (Historical)

Effect. Energy Multigroup Gamma Source Terns, /cm3-sec, at t = ... After Accident (a)

Grp MeV 0b 30 min 1 hr 2 hr 10 hr 1 day 100 hr 7 day 30 day 1000 hr 6 month 1 year (c) 1 .150 0 9.25+7 8.68+7 8.01+7 6.26+7 5.04+7 3.11+7 2.40+7 7.79+6 5.89+6 3.04+6 1.97+6 2 .250 0 9.36+7 9.06+7 8.70+7 8.05+7 7.32+7 4.67+7 3.35+7 3.96+6 1.47+6 1.39+4 2.01+3 3 .350 0 1.14+9 1.00+9 8.44+8 6.96+8 6.59+8 4.98+8 3.88+8 5.34+7 1.96+7 2.61+3 9.88+1 4 .475 0 2.03+9 1.98+9 1.89+9 1.36+9 8.11+8 9.29+7 3.62+7 1.69+7 1.31+7 1.02+6 4.14+4 5 .650 0 1.66+9 1.33+9 9.03+8 1.55+8 9.27+7 7.15+7 5.92+7 2.15+7 1.68+7 6.91+6 4.66+6 6 .825 0 3.46+9 2.56+9 1.48+9 2.23+8 1.27+8 5.35+7 4.51+7 3.09+7 2.61+7 5.75+6 2.61+6 1.000 0 6.08+8 5.48+8 4.59+8 1.59+8 4.15+7 1.53+6 3.77+5 4.26+4 4.06+4 3.07+4 2.56+4 8 1.225 0 1.34+9 1.15+9 8.98+8 3.66+8 1.19+8 5.28+6 1.22+6 2.89+5 1.97+5 5.97+4 4.51+4 9 1.475 0 1.6.5+8 1.42+8 1.09+8 3.16+7 1.58+7 1.35+6 2.42+5 1.10+5 1.09+5 9.63+4 8.09+4 10 1.700 0 7.24+8 6.25+8 4.95+8 1.97+8 5.92+7 5.33+6 1.65+6 1.21+2 0 0 0 11 1.900 0 4.52+7 3.26+7 2.01+7 1.67+6 9.08+4 1.19+4 2.47+3 0 0 0 0 12 2.100 0 0 0 0 0 0 0 0 0 0 0 0 13 2.300 0 3.43+6 1.55+6 4.02+5 1.30+1 0 0 0 0 0 0 0 14 2.500 0 5.38+7 4.83+7 4.11+7 1.77+7 4.56+6 1.90+5 5.83+4 4.28+0 0 0 0 15 2.700 0 1.49+6 7.59+5 2.03+5 6.48+0 0 0 0 0 0 0 0 16 3.000 0 0 0 0 0 0 0 0 0 0 0 0 17 3.500 0 4.16+5 5.50+4 9.69+2 0 0 0 0 0 0 0 0 18 4.000 0 1.13+7 5.86+6 1.59+6 4.57+1 0 0 0 0 0 0 0 (a) 50% core halogens, 1% other nuclides diluted in 1.62 x 109, cm3 water (b) Recirculation does not begin before 20 minutes following accident (e) Read as 9.25 x 107 g /(cm3-sec)

T12.3A-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3A-4 CONTAINMENT ATMOSPHERE SOURCE TERMS (Historical)

Effect. Energy Multigroup Gamma Source Terms, /sec, at t= ... After Accident (a) (b)

Grp MeV 0 2 min 46 min 1 hr 2 hr 8 hr 1 day 4 days 30 day 1 .150 6.44+18(c) 6.37+18 5.64+18 5.57+18 5.38+18 4.82+18 4.20+18 2.79+18 8.71+1 2 .250 8.35+18 8.15+18 6.00+18 5.75+18 5.06+18 2.84+18 7.94+17 8.88+15 2.43+1 3 .350 1.21+18 1.19+18 2.20+17 2.08+17 1.76+17 9.43+16 4.89+16 3.38+16 3.38+1 4 .475 5.81+18 5.53+18 1.35+18 1.08+18 6.01+17 1.37+17 6.23+16 5.35+15 1.12+1 5 .650 1.77+18 1.75+18 2.80+17 2.63+17 2.24+17 1.48+17 8.82+16 3.68+16 4.65+1 6 .825 4.55+18 4.46+18 7.71+17 6.99+17 4.99+17 1.45+17 6.38+16 3.05+16 1.12+1 7 1.000 4.83+17 4.82+17 4.53+16 4.32+16 4.05+16 2.88+16 1.63+16 7.07+15 2.62+1 8 1.225 1.35+18 1.33+18 1.01+17 9.08+16 7.06+16 3.42+16 9.39+15 3.92+14 3.89+5 9 1.475 5.46+17 5.43+17 3.591+17 3.39+17 2.66+17 6.73+16 9.82+15 4.09+15 1.49+1 10 1.700 1.73+18 1.63+18 1.89+17 1.27+17 5.45+16 1.74+16 3.20+15 1.92+12 0 11 1.900 2.33+16 2.33+16 2.29+15 2.20+15 2.18+15 2.07+15 1.79+15 9.40+14 3.51+1 12 2.100 1.14+18 1.06+18 3.61+17 3.00+17 1.96+17 3.95+16 7.11+14 1.05+7 0 13 2.300 1.71+18 1.69+18 1.41+18 1.33+18 1.03+18 2.29+17 4.21+15 6.23+7 0 14 2.500 3.41+17 3.36+17 2.04+17 1.80+17 1.07+17 5.89+15 3.15+14 1.89+11 0 (a) 100% core noble gases, 25% halogens (b) To obtain specific source terms. ( /(cm3-sec)), divide by Containment free volume of 7.10 x 1010 cm3 (2.506 x 106 ft3)

(c) Read as 6.44 x 1018 /sec.

T12.3A-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12. 3A-5 CONTAINMENT PLATEOUT SOURCE TERMS (Historical)

Effect. Energy Multigroup Gamma Source Terms, /sec, at t = ... After Accident (a)

Grp MeV 0 2 min 46 min 1 hr 2 hr 8 hr 1 day 4 days 30 days (b) 1 .150 1.54+16 1.54+16 1.54+16 1. 53+16 1.53+16 1. 49+16 1.41+16 1.09+16 1.17+15 2 .250 3.62+16 3.62+16 3.62+16 3.60+16 3.60+16 3.51+16 3.32+16 2.58+16 2.75+15 3 .350 1.05+18 1. 03+18 8.00+17 7.48+17 6.13+17 4.91+17 4.62+17 3.59+17 3.83+16 4 .475 1.65+18 1.65+18 1.59+18 1.57+18 1.50+18 1.05+18 6.28+17 5.57+16 5.91+7 5 .650 1.59+18 1.57+18 1.26+18 1.19+18 1.01+18 8.17+17 7.14+17 3.90+17 5.28+15 6 .825 3.88+18 3.80+18 2.50+18 2.23+18 1.47+18 7.94+17 6.60+17 3.25+17 1.27+15 7 1.000 4.83+17 4.82+17 4.59+17 4.52+17 4.26+17 1.97+17 1.72+17 7. 5 3+16 2.96+14 8 1.225 1.35+18 1.33+18 1.02+18 9.50+17 7.39+17 1.31+17 9.89+16 4.18+15 4.40+6 9 1.475 1.27+17 1.27+17 1. 25+17 1.25+17 1.24+17 1.13+17 9.26+16 4.35+16 1.68+14 10 1.700 6.96+17 6.87+17 5.30+17 4.95+17 3.86+17 3.43+16 3.37+16 2.07+13 0 11 1.900 2.33+16 2.33+16 2.31+16 2.31+16 2.31+16 2.16+16 1.89+16 1.00+16 3.97+13 12 2.100 0 0 0 0 0 0 0 0 0 13 2.300 0 0 0 0 0 0 14 2.500 3.92+16 3.91+16 3.62+16 3.53+16 3.19+16 3.31+15 3.31+15 2.03+12 0 (a) 25% core iodine (b) Read as 1.54 x 1016 /sec T12.3A-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3A-6 SYSTEMS POTENTIALLY CONTAINING HIGH LEVELS OF RADIOACTIVE MATERIALS Containment Spray System Safety Injection System Low Pressure Safety Injection High Pressure Safety Injection Shutdown Cooling System Post Accident Sampling Systems Liquid Sampling Hydrogen Analyzer Ventilation Systems Shield Building Ventilation System Control Room Emergency Ventilation System ECCS Area Ventilation System Containment Building T12.3A-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.3A-7 AREAS IDENTIFIED IN SHIELDING REVIEW AS REQUIRING ACCESSIBILITY FOLLOWING AN ACCIDENT Maximum Dose Area Location Occupancy Requirements Rate and Dose Remarks (1) Control Room RAB EL 62.00' Continuous <15 mrem/hr Door shown in wall between Figure 12.3A-4 1 hr after accident Room. Emergency Filter Room NW Corner <5 rem for duration (containing HVE-13A,B) and of accident Control Room kitchen is (2) Technical Unit No. 1 Control Room Continuous <15 mrem/hr Units 1 and 2 share the TSC.

Support Center Envelope 1 hr after accident

<5 rem for duration of accident (3) Valve Station: RAB EL-0.50' 2 men for 10 minutes >6,000 rem/hr Manual valves are fitted with LPSI Pump Suction Figure 12.3A-1 1 hr after accident motor operators operated from Isolation Valves ECCS Room ~1,000 rem per person Control Room. Therefore, V3432, V3444 (Historical-not updated for EPU) will no longer be required.

(4) Valve Station: RAB EL 19.50' 2 men for 10 minutes >5,000 rem/hr Same as Item (3)

Containment Spray Figure 12.3A-2 1 hr after accident Discharger Header Penetration Area ~840 rem per person Isolation Valves (Historical-not updated for EPU)

MV-07-3 & 4 (5) Hydrogen Analyzer RAB EL 43.00' Infrequent access up to ~290 mrem/hr Special shielding is designed to Cubicle Figure 12.3A-3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at a time for maintenance 1 hr after accident accommodate the Hydrogen SE part of HVAC Room to one analyzer. ~1.72rem per maintenance Analyzers so that required Also, short (few minutes) period be maintained. Refer to Sub-access to obtain grab sample. section 6.2.5 for a description of the Hydrogen Analyzers.

(6) Post Accident RAB EL 19-50' Infrequent access to <100 mrem/hr Refer to Subsection 9.3.6 Sampling System Figure 12.3A-2 collect grab sample. 1 hr after accident for a description of PASS.

(PASS)* South wall of RAB <100 mrem per visit (Historical-not updated for EPU)

(7) Sample Analysis RAB EL 19.50' Frequent <100 rem/hr 1 hr after High dose rate precludes use of Areas (Health Figure 12.3A-2 accident Health Physics area. Therefore, Physics Area) NW Part of RAB >5 rem for duration of all sample analysis and health accident physics monitoring will be done (Historical-not updated for EPU) from an alternate location as discussed in plant procedures.

(8) ECCS Vent Monitors RAB EL 43.00' Infrequent access to collect < 132 mrem/hr Special shielding is designed Figure 12.3A-3 particulate filters for analysis 1 hr after accident to accommodate the ECCS Vent East side of RAB < 132 mrem per visit Monitors so that required access can be maintained. See Section 11.5 for a description of the monitors.

• Access no longer required since PASS is no longer used.

T12.3A-7 Amendment No. 24 (09/17)

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UFSAR/St. Lucie - 2 12.4 DOSE ASSESSMENT (Historical) 12.4.1 ANTICIPATED DOSE RATES The radiation zone designation for an area is determined by the maximum expected whole body dose rate that should be experienced in that area whether from sources of radiation within the same area, or from sources located external to that area. The radiation zones are shown on Figures 12.3-4 through 12.3-12. A description of the zones is given in Subsection 12.3.2.2.

Maximum zone doses are not expected to occur during normal operation. The highest doses in the plant occur in Zone V areas, such as in rooms containing equipment and piping handling radioactive fluids, or in the containment.

Shielding is used to maintain much of the plant within the lowest zone levels. Also experience in the design of nuclear power plants shows that the actual measured radiation levels are usually considerably less than those used for the shielding design objectives for controlling the radiation doses. Therefore the annual exposure doses received by the plant personnel are well below the limits of 10 CFR 20.

In addition to exposure from continual radioactive sources (direct exposure) radiation exposure within the plant from airborne radionuclides is possible. Occupational exposure from these sources is usually insignificant in comparison to direct radiation exposure. Ventilation systems continually direct air from areas of lesser radioactivity to areas of greater radioactivity then to the plant vent. Area monitors and health physics surveys also aid in precluding exposures so that the contribution of airborne radioactivity to the total man-rem exposure dose is insignificant in comparison with exposure to direct radiation (refer to Subsection 12.2.2).

During normal operation radiation exposures from direct radiation at locations outside the plant structures are insignificant since there are no potential sources of radiation in these areas.

Therefore, the total exposure dose from the plant is given essentially by the occupational exposure to direct radiation within plant structures.

The control room and normally occupied areas are Zone I control access areas, and hence have a maximum allowable dose rate of 0.25 mr/hr. Therefore, annual doses in these areas, considering occupancy factors, are well within the limits of 10 CFR 20, particularly since the actual dose rates are well below 0.25 mr/hr based on PWR operating experience.

12.4.2 ESTIMATE OF EXPOSURE OF PLANT PERSONNEL Nuclear plant operating experience is reviewed in order to estimate the annual exposure to operating personnel. Table 12.4-1 presents data on total number of personnel and total annual dose in operating PWRs. These data, taken from References 1 and 2 are supported by data given in Reference 3.

Table 12.4-1 data is reduced to a weighted average man-rem for operating power plants in Table 12.4-2. Table 12.4-3 (data from Reference 1) shows the distribution of man-rem doses for various functions of operating light water reactors (including BWRs).

Estimated doses are derived from Table 12.4-4 which references data presented in Reference 3 for exposures received during maintenance, in-service inspection, waste handling and refueling.

12.4-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 Based on this operating data, the total man-rem dose for St. Lucie Unit 2, as shown in Table 12.4-5, is estimated to be approximately 440 man-rem/year.

It should be recognized that the predicted dose rates are estimates, and that variations in the working force, radiation levels, unplanned maintenance, and repair work, etc., may result in different true exposures. These estimates provide a reasonable upper bound to the normal operation potential exposures.

Table 12.4-5 summarizes the estimated annual exposure dose to plant personnel separated into the five categories of Table 12.4-3. The assumptions in the estimate are as follows:

a. The plant is operated by three shifts and four crews, each crew consisting of a minimum of 22 personnel: One shift supervisor, three other operators (control center operators, turbine generator operators, etc.), three nuclear operators, three mechanical, three electrical and three instrument and control maintenance personnel, two radio-chemistry personnel, three health physics personnel and one technician.

b. The majority of the maintenance personnel are normally available during the day shift only.

c. Technical staff personnel work on day shift primarily and are not present on weekends unless special occurrences dictate their presence.

The technical staff spend most or all of their time in uncontrolled areas where the radiation levels are expected to be less than 0.03 mr/hr. For instance, the Nuclear Engineer and his associate engineer are involved essentially in office work except during refueling. Likewise the Health Physics Engineer and the Chemical and Environmental Engineer and their associates and technicians spend a considerable part of their shift in office work. The latter however, are engaged in periodic plant surveys and Health Physics Technicians are present during many maintenance operations. The latter may spend approximately one third of their shifts in laboratory work and collecting certain samples.

The nuclear operators are charged with patrolling the containment, Reactor Auxiliary, Fuel Handling and Turbine Buildings. Mechanical maintenance, electrical maintenance, and I & C work is performed solely by the mechanical, electrical, and I & C personnel.

12.4.2.1 Reactor Operation and Surveillance Exposures received for reactor operation and surveillance are estimated by assuming that control room operation, radiochemical laboratory work, and office work result in negligible exposures. Therefore the exposure attributed to reactor operation and surveillance stems from routine patrol of plant areas housing potentially radioactive equipment, periodic tests, operations and work activities (excluding those work activities which entail assembly, disassembly and repair, and are then classified as maintenance activities).

Reactor operations and surveillance activities include local operations such as valve alignment, starting and stopping pumps, patrolling, operation of the radwaste system (but not the actual handling of the solid wastes), sampling, radiation and contamination surveys and occasional lubrication.

12.4-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 The plant is assumed to be operating with 0.1 percent failed fuel, the expected value based on experience with other nuclear plants (rather than the radiation protection design value of 1.0 percent failed fuel). The exposures listed in Table 12.4-6 are predictions based on personnel surveillance of the plant in pairs, with inspection of every active equipment.

The overall exposures for plant operation and surveillance agree with those derived from Reference 1.

12.4.2.2 Maintenance (Routine and Typical)

Estimated dose exposures from maintenance activities reflect what has been experienced by the industry. The St. Lucie Unit 2 plant is designed in such a manner as to minimize exposure doses to personnel, consonant with the ALARA provisions of Regulatory Guide 8.8 "Information Relevant to Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable (Nuclear Power Reactors), June 1978 (R3). The design features are described in detail in Subsection 12.1.2. Equipment which may be radioactive and will require maintenance is housed in individually shielded cubicles, with sufficient space allocated for the erection of portable shields to minimize exposures. In addition, provisions have been made to enable flushing of lines and equipment such as concentrates and resin carrying lines and pumps so that maintenance on those items can be accomplished with a minimum of exposure.

Some maintenance operations, for which historical average exposures under routine maintenance are quoted in Table 12.4-4, may be included in the reactor operations and surveillance estimated exposures for periodic maintenance. Table 12.4-6 lists typical exposure doses received by plant personnel during routine patrol operations, and Table 12.4-7 lists typical exposure doses received during the performance of routine maintenance. Exposures listed in Table 12.4-7 reflect discussions held with operating utilities regarding the time required for the performance of the tasks.

For instance, the estimated exposure for some routine maintenance on the charging pumps (see Table 12.4-7) is given as 5.6 man-rem/year. It is unclear from Reference 3 whether the exposures quoted for the charging pumps (item 12) and routine maintenance (item 13) include the exposure for the activities listed in Table 12.4-7. Therefore it is possible that these exposures have been accounted for twice in Table 12.4-5; once under operations and surveillance and once under maintenance. Thus the exposures quoted for maintenance in Table 12.4-5, which reflect operating experience, represent an upper bound.

12.4.2.3 In-service Inspection Exposures for in-service inspection (ISI) are estimated from operating plant experience. Where ISI needs to be performed, the plant layout is designed to provide adequate access to locations and is shielded from nearby sources. The allocation of sufficient space and compartmentalization of individual major sources of activity should minimize exposures resulting from ISI activities.

12.4.2.4 Waste Processing The exposure estimated for radwaste handling and operations is deduced from information contained in Reference 3, which is summarized in Table 12.4-4. However, certain St. Lucie Unit 2 features should result in even lower exposure. For instance, liquid filter removal is 12.4-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 accomplished remotely, thus avoiding the relatively high exposures normally associated with the routine handling filter cartridges.

Furthermore, some of the exposures resulting from the operation and maintenance of liquid radwaste systems, as quoted in Table 12.4-4 item 1, are included in Tables 12.4-6 (routine patrol) and 12.4-7 (periodic maintenance). In this sense, the figure of 35 man-rem for radwaste operation and handling given in Table 12.4-5 represents an upper bound of the exposure which may be expected from such activities in the plant.

12.4.2.5 Refueling Exposures for refueling have been estimated from data presented in Reference 3. Separate estimates performed for other plants confirm that exposures from refueling operations can range to approximately 50 man-rem. This is based on data from actual refueling outages at nuclear plants which have similar refueling features. Design features, such as additional shielding around the spent fuel transfer tube reduce exposures, however, estimates derived from other operating plants show approximately 1/3 of the exposure was accrued by health physics personnel providing coverage for the refueling. Therefore only partial improvement on the exposure is expected as a result of such design features.

12.4-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 SECTION 12.4: REFERENCES

1. Ninth Annual Occupational Radiation Exposure Report, 1976, NUREG-0322, USNRC Office of Management Information and Program Control (October 1977).

2. T.D. Murphy, et. al, Occupational Radiation Exposure at Light Water Cooled Power Reactors, NUREG-75/032, USNRC, Radiological Assessment Branch (June 1975).

3. Compilation and Analysis of Data on Occupational Radiation Experienced at Operating Nuclear Power Plants, National Environmental Studies Project, SAI Services (September 1974).

12.4-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-1 DATA FROM OPERATING PWR PLANTS(1)

Designed Power Level Total No. of Total Annual Year Plant (MWe) Personnel Dose (Man-Rem) 1970 Connecticut Yankee 575 734 689 San Onofre - Unit 1 450 251 155 1971 Connecticut Yankee 575 289 342 Ginna 490 340 430 San Onofre - Unit 1 450 121 50 1972 Connecticut Yankee 575 355 325 Ginna 490 677 1,032 Point Beach - Unit 1 497 NA 580 Robinson 707 245 215 San Onofre - Unit 1 450 326 256 1973 Connecticut Yankee 575 841 673 Ginna 490 421 244 Palisades 821 901 1,109 Point Beach - Units 1 497, 497 729 570

& 2 (2nd Unit 4/73)

Robinson 707 831 695 San Onofre - Unit 1 450 878 329 1974 Connecticut Yankee 575 550 201 Fort Calhoun 457 327 71 Ginna 490 884 1,225 Haddam Neck 575 550 201 Maine Yankee 790 619 420 Oconee - Unit 1 886 844 517 Palisades 821 774 627 Point Beach - 497, 497 400 295 Units 1 & 2 Robinson 707 853 672 San Onofre 450 219 71 Surry - Units 1 & 2 823, 823 1,715 884 (Unit 2 5/73)

Turkey Point - 745, 745 794 454 Units 3 & 4 (Unit 4 9/73)

(1)These are taken from data for operating PWR plants given in References 1 and 2. In compiling this table, generally data form the first year of plant operation has not been considered. Only data from those PWR plants that are designed to operate at power levels greater than or equal to 450 MWe were chosen.

T12.4-1 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-1 (Cont'd)

Designed Power Level Total No. of Total Annual Year Plant (MWe) Personnel Dose (Man-Rem) 1975 Arkansas 1 850 147 21 Calvert Cliffs 1 1,065 783 97 Fort Calhoun 457 469 294 Ginna 490 685 538 Haddam Neck 575 795 703 Kewaunee 560 104 28 Maine Yankee 790 440 319 Oconee Unit 1 886 829 497 Palisades 821 495 306 Point Beach 1 & 2 497, 497 339 459 Robinson 707 849 1,142 San Onofre 450 424 292 Surry 1 & 2 823, 823 1,948 1,649 Turkey Point 3 & 4 745, 745 1,176 876 1976 Arkansas 1 850 476 289 Calvert Cliffs 1 1,065 507 74 D C Cook 1,090 395 116 Fort Calhoun 457 516 313 Ginna 490 758 636 Haddam Neck 575 644 449 Kewaunee 560 381 270 Maine Yankee 790 244 85 Oconee 1 & 2 886 1,215 1,026 Palisades 821 742 696 Point Beach 1 & 2 447 313 370 Prairie Island 1 & 2 NA 818 447 Robinson 707 597 715 San Onofre 450 1,330 880 Surry 1 & 2 823, 823 2,753 3,165 Three Mile Island 819 819 286 Turkey Point 3 & 4 745, 745 1,647 1,184 Zion 1 & 2 1,015, 1,015 774 571 NA: Not Available T12.4-2 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-2 YEARLY AVERAGES AND GRAND AVERAGE FOR NUMBER OF PERSONNEL AND MAN-REM DOSES FOR OPERATING PWR PLANTS(1)

Total Total Average Average No. of No. of Man-Rem No. of Man-Rem Year Units Personnel Dose Personnel Dose/Unit 1970 2 985 844 493 422 1971 3 750 822 250 274 1972 5 1,603(2) 2,408 401(2) 482 1973 7 4,601 3,620 657 517 1974 15 8,529 5,538 568 369 1975 16 9,483 7,794 593 487 1976 23 15,416 12,031 670 523 1970- 71 41,367 33,057 583 466 1976 (1) This table is based on the data given in Table 12.4-1.

(2) The entry corresponds to four plants only, since no information on personnel is available for Point Beach, Unit 1.

T12.4-3 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-3 DISTRIBUTION OF MAN-REM DOSES FOR VARIOUS FUNCTIONS FOR OPERATING LIGHT WATER REACTORS*

(Historical Information)

Percentage of Total Man-Rem Reactor Operation & Surveillance 10.4 Maintenance 71.2 (Routine & Typical)

In-service Inspection 5.7 Waste Processing 4.8 Refueling 7.9

• Includes BWRS; information is derived from Reference (1).

T12.4-4 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-4 OCCUPATIONAL RADIATION EXPOSURES EXPERIENCED AT OPERATING NUCLEAR POWER PLANTS* BY TYPE OF WORK (Historical Information)

1. Operation and maintenance of liquid radwaste systems 22.7 man-rem/yr, including a contractor's exposure for construction work of new radwaste building and other systems. Otherwise, 9.4 man-rem/yr, although there is some evidence to indicate that the overall figure should range from 10-30 man-rem/yr.

2. Solid waste handling 9.5 man-rem/yr. There is no clear correlation with plant age or the activities of materials handled due to the random occurrence of broken or otherwise unusable objects requiring handling.

3. Gaseous radwaste systems 1.0 man-rem/yr, based on only one year's experience at only one plant, and it is due to work on fans and charcoal filters. The dose rate for work on vacuum pumps, holdup systems, and decay tanks is not known.

4. Reactor vessel head removal and installation 29 man-rem per removal-and-installation cycle, due mainly to accumulation of crud in the control-rod drive mechanisms.

5. Fuel handling 11.7 man-rem/yr, including exposure to a reactor vendor's personnel. Otherwise, 8.4 man-rem/yr. The probable range is from 6.4 to 33 man-rem/yr, apparently due to random difficulties arising from removal and replacement of fuel assemblies. Such difficulties are considered common, and a more likely personnel dose rate would be 20 man-rem/yr.

6. Instrument work including calibration 4.5 man-rem/yr, including in-core work. There seems to be no connection with plant age.

7. In-service inspection Probably 50 - 60 man-rem/yr, but data is difficult to isolate for just this category alone

• Summarized from Reference 3.

T12.4-5 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-4 (Cont'd

8. Major equipment failures Three incidents are reported for PWRs two of them at the same plant: 1100 man-rem/7 months to replace a thermal sleeve in one of the main coolant loops after eight years of operation; and 3500 man-rem/6 months to replace a section of pipe in eight downcomers to horizontal U-tube steam generators after 11 years of operation at the same plant; and, at another plant, 68 man-rem/yr to partially repair a leak in the spent fuel pit after four years of operation.

9. Steam generators 113 man-rem/yr for many maintenance and repair operations on steam generators.

Because of the diversity of operations, no age dependence is apparent, but should be expected as crud builds up. Sophisticated methods of tube-plugging lower personnel doses, but tube-plugging seems to be a minor contributor. Decontamination does not seem to lower overall doses.

10. Reactor Coolant Pumps 13.4 man-rem/yr for a variety of maintenance and repair operations. The data are insufficient to establish an age-dependence, but again, such dependence should be expected as crud accumulates.

11. Main coolant loops 19.0 man-rem/yr for a variety of work at two PWR plants. The data suggests, but does not establish age-dependence; however, such dependence should be expected. The personnel dose is dependent more on loop layout and on the necessity to work on valves.

12. Charging pumps 4.6 man-rem/yr, including a special job to install a pressure-pulse filter. Otherwise 2.5 man-rem/yr.

13. Routine Maintenance Reference 3 to Section 12.4 quotes that average exposures from routine operations and maintenance account for up to 33 percent of the total exposure. Since reactor operations and surveillaance represent approximately 10 percent, the routine maintenance accounts for 23 percent of the exposure.

T12.4-6 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-5 ESTIMATED ANNUAL EXPOSURE DOSES TO PLANT PERSONNEL (Historical Information)

Type of Work Exposures Man-rem

1. Reactor Operation and Surveillance 45

2. Maintenance (routine and typical)(a) 255

3. In-Service Inspection 55

4. Waste Processing(b) 35

5. Refueling 50 Reactor Vessel Head Removal and Installation(c) (30)

Fuel Handling(d) (20)

Total 440 man-rem Notes:

a. Items 6, 9, 10, 11, 12, 13 of Table 12.4-4

b. Items 1, 2, 3 of Table 12.4-4.

c. Item 4 of Table 12.4-4

d. Item 5 of Table 12.4-4 T12.4-7 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-6 PREDICTED DOSES TO PLANT PERSONNEL DURING ROUTINE PATROL (Historical Information)

DATA Time Req'd. Max. Dose Dose per Frequency (Man-Min/ Rate Shift Activity (per shift) Activity) (mR/hr) (man-mR)

Walking 3300 ft.

@ 120 ft/min. 3 27.3 0.3 0.824 Walking 300 ft.

@ 120 ft/min.

(containment) 1 2.5 12 0.100 Spent Resin Storage Pump & Valve Inspection 1 0.1 24 .08 Floor Drain Tank Pump & Valve Inspection 1 0.2 24 .16 Concentrates Transfer Pump & Valve Inspection 1 0.1 24 .08 Detergent Waste Pump & Valve Inspection 1 0.2 6 .04 Intake Cooling Water Pump & Valve Inspection 1 0.2 0.3 .002 Charcoal Delay Beds Valve Inspection 1 0.1 3.0 .010 High Conductivity Sample Pump & Valve Inspection 1 0.2 0.3 .002 Waste Gas Compressors Inspection 1 5.0 24 4.0 Condensate Valves & 1 0.5 .3 .0050 Pumps Sump Pumps & Valves 3 5.0 6 1.5 Boric Acid Make-up Pump & Valve Inspection 1 0.2 6 .04 Volume Reduction System Pump & Valve Inspection 1 0.2 6 .04 Detergent Waste Sampling Pump &Valve Inspection 1 0.2 6 04 T12.4-8 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-6 (Cont'd)

DATA Time Req'd. Max. Dose Dose per Frequency (Man-Min/ Rate Shift Activity (per shift) Activity) (mR/hr) (man-mR)

Floor Drain Evaporator Pump & Valve Inspection 1 0.2 12 .08 Floor Drain Demineralizer Pump & Valve Inspection 1 0.2 6 .04 H.P. Safety Injection Pump & Valve Inspection 1 0.2 12 .08 Containment Spray Pump & Valve Inspection 1 0.2 12 .08 L.P. Safety Injection Pump & Valve Inspection 1 0.2 12 .08 Reactor Drain Pump & Valve Inspection 1 0.2 12 .08 I.C.W. Evaporator Recirc.

Pump & Valve Inspection 1 0.2 6 .04 Low Purity Sump Pump & Valve Inspection 1 0.4 6 .08 High Purity Pump & Valve Inspection 1 0.4 6 0.08 Gas Stripper Valve Inspection 1 0.5 6 .10 Shutdown Heat Exchanger Valve Inspection 1 0.2 3 .02 Equipment Drain Tank Pump & Valve Inspection 1 0.2 6 .04 High Purity Sump Pumps Inspection 1 0.2 6 .04 Letdown Hx Valve Pump & Valve Inspection 1 1.5 12 0.4 Charging Pump & Valve Inspection 1 1.5 12 0.6 Fuel Pool. Cleanup Pump & Valve Inspection 1 1.0 6.0 0.2 T12.4-9 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-6 (Cont'd)

DATA Time Req'd. Max. Dose Dose per Frequency (Man-Min/ Rate Shift Activity (per shift) Activity) (mR/hr) (man-mR)

Boric Acid Evap.

Valve Inspection 1 1.0 6 0.2 Volume Control Tank Valve Inspection 1 0.5 6.0 0.10 Ion Exchangers Valve Inspection 1 2.0 6.0 0.4 Health Physics Survey, 0.3 120 5 10.0 etc.

Blowdown Equipment Pump & Valve Inspection 1 2.0 6 0.4 Radwaste Control Reading

& Checking 1 5.0 .03 .004 Inside Containment Instrumentation 0.3 15 6.0 3.0 Total Routine Patrol = 23.463 man-mR/Shift Total Routine Patrol = 25.7 man-rem/yr T12.4-10 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-7 ESTIMATED DOSES TO PLANT PERSONNEL DURING PERIODIC MAINTENANCE DATA Frequency Time Req'd. Max. Dose Crew Dose (Per Unit (Man-Hr/ Rate (Man-Rem/

Activity Shown) Activity) (mR/hr) year)

Spent Resin Storage Pump Maintenance Year 20 24 .48 Floor-Drain Condensate Pump Maintenance Year 20 24 0.48 Concentrator Maintenance Year 90 24 2.16 Concentrator Heat Exchanger Cleaning Year 40 24 0.96 Concentrate Metering Pump Maintenance Year 10 24 .24 Detergent Waste Pump Maintenance Year 20 24 .48 Gas Stripper Instrument Adjustment Month 2 24 .576 Gas Stripper Heat Exchangers Cleaning Year 16 24 .384 Gas Stripper Pump & Other Maintenance Year 40 24 0.96 Waste Gas Compressors Maintenance Year 20 24 0.48 Boric Acid Make-up Pump Maintenance Year 20 .3 .006 Spent Resin Metering Pump Maintenance Year 10 24 0.24 T12.4-11 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-7 (Cont'd)

DATA Frequency Time Req'd Max. Dose Crew Dose (Per Unit (Man-Hr/ Rate (Man-Rem/

Activity Shown) Activity) (mR/hr) year)

Boric Acid Condensate Sampling Pump Maintenance Year 20 24 0.48 Inorganic Chemical Waste Pump Maintenance Year 20 0.6 .012 Floor Drain Pump Maintenance Year 20 12 .24 H.P. Safety Injection Pumps a) Mechanical Seal Maintenance 2 Year 16 10 .080 b) Bearing Oil Change 2 Year 8 10 .040 c) Coupling Alignment Check 6 Months 4 10 .080 d) Coupling Lubrica-tion Check 6 Months .5 10 .010 e) Coupling Lubrica-tion Change 2 Year 8 10 .040 f) Motor Air-Filter Cleaning 3 Months 4 10 0.16 g) Motor Lubrication Change 2 Year 8 10 04 Containment Spray Pump Maintenance Year 20 10 0.2 Shutdown Cooling Heat Exchangers Cleaning Year 10 10 0.1 Letdown HX Valve Maintenance Year 40 25 1.00 T12.4-12 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-7 (Cont'd)

DATA Frequency Time Req'd. Max. Dose Crew Dose (Per Unit (Man-Hr/ Rate (Man-Rem/

Activity Shown) Activity) (mR/hr) year)

Letdown Heat Exchanger Cleaning Year 40 50 2.00 Regenerative HX Valve Maintenance Year 40 25 1.00 Regenerative Heat Exchanger Cleaning Year 10 50 0.5 L.P. Safety Injection Pumps a) Mechanical Seal Maintenance 2 Year 8 10 .04 b) Bearing Oil Change 2 Year 4 10 .02 c) Coupling Alignment Check 6 Months 2 10 .04 d) Coupling Lubrication Check 6 Months .25 10 .005 e) Coupling Lubrication Change 2 Year 4 10 .02 f) Motor Air Filter Cleaning 3 Months 2 10 .08 g) Motor Lubrication Change 2 Year 4 10 .02 Reactor Drain Pump Maintenance Year 20 24 .48 I.C.W. Evaporator Recirculation Pump Maintenance Year 20 .6 .012 T12.4-13 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 TABLE 12.4-7 DATA Frequency Time Req'd. Max. Dose Crew Dose (Per Unit (Man-Hr/ Rate (Man-Rem/

Shown) Activity) (mR/hr) year Equipment Drain Pump Maintenance Year 10 .6 .006 I.C.W. Condensate Pump Maintenance Year 20 .6 .012 Floor Drain Evaporator Pump Maintenance Year 10 0.6 .006 Charging Pumps a) Plunger Packing Replacement 3 Months 8 50 1.60 b) Pump Oil &

Filter Change 6 Months 8 50 0.8 c) Motor Bearing Repacking Year 16 50 0.8 d) Lubrication Pump Lubrication 6 Months 4 50 0.4 e) Gear Reducer Oil Changing 3 Months 8 50 1.60 f) Coupling Alignment Year 4 50 0.2 g) Coupling Lubrication Year 4 50 0.2 Spent Fuel Pool Heat Exchangers Cleaning Year 20 5.0 .100 Spent Fuel Pool Cooling Pump Maintenance Year 20 5.0 .100 Spent Fuel Pool Clean-up Pump Maintenance Year 20 5.0 .100 Total - Periodic Jobs = 20.07 man-rem/yr T12.4-14 Amendment No. 24 (09/17)

UFSAR/St. Lucie - 2 12.5 HEALTH PHYSICS PROGRAM 12.5.1 ORGANIZATION 12.5.1.1 Program Objectives The program objectives are:

a. To implement a radiation protection program for protecting the health and safety of plant personnel and the public.

b. To establish and maintain a comprehensive record system which will demonstrate the adequacy of the radiation protection program to ensure that occupational exposures areas low as reasonably achievable and which complies with all applicable Federal and State regulations.

c. To ensure that all radioactive effluent releases and waste shipments meet guidelines and established in plant procedures.

d. To ensure that radioactive effluent releases are within technical specification requirements and maintained "as low as is reasonably achievable" (ALARA).

A detailed discussion on ALARA responsibilities, and management policy is provided in Section 12.1.

To ensure that the program objectives are implemented, the health physics program ensures that:

a. All radiation workers receive radiation protection training commensurate with their respective responsibilities.

b. Respiratory protection equipment training is provided to workers who may use the equipment.

c. Emergency plan training is provided as necessary for personnel who may be assigned to radiation emergency teams.

d. Appropriate personnel dosimetry is available.

e. Internal and external dose assessment is provided for workers.

f. Workers' internal and external exposure history are known prior to establishing allowable exposure limits.

g. Respiratory protection equipment is provided to keep internal exposure ALARA.

h. Restricted areas are segregated to control exposure potential.

i. Access to restricted areas is through approved procedures which control exposure potential.

j. Radiological instrumentation is provided and maintained to assess exposure potential.

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k. Incoming shipments of radioactive material are received and surveyed properly.

l. Outgoing shipments of radioactive material are packaged, surveyed, and labeled properly.

m. Necessary measures are taken and guidelines followed to keep exposures and effluents ALARA while safely supporting a reliable source of power to the public.

12.5.1.2 Staff Organization The health physics program is established to ensure that the radiation protection and training requirements of 10 CFR 19, 10 CFR 20, 10 CFR 50 Appendix I, and Regulatory Guide 8.2, "Guide for Administrative Practices in Radiation Monitoring," February 1973 (R0), Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Plants will be As Low As Is Reasonably Achievable," June 1978 (R3), Regulatory Guide 8.10 "Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably Achievable," September 1975 (R1) and Regulatory Guide 1.8, "Personnel Selection and Training," May 1977 (R1-R) are met.

The Radiation Protection Manager has the specific responsibility and authority to ensure that EC291105 the radiation protection program maintains exposures ALARA. He reports to the Site Director.

He has direct access to the Site Director on matters related to radiological health and safety of employees and the public. He fulfills the responsibilities of the radiation protection manager as defined in the Technical Specifications.

All individuals in the Health Physics Department meet the qualification criteria as outlined in the health physics instructions on qualification and training of health physics personnel before they are delegated responsibility in the radiation protection program (see Sections 13.1 and 13.2).

Exposure reduction training and job application are covered in detail in the Radiation Protection Training Program. Foremen and supervisors are given extensive training in addition to continuing communication with Health Physics in the form of written information or instructions and meetings, so that the foremen and supervisors are fully able to cope with specific jobs order EC291105 to minimize radiation exposure. ANSI/ANS-3.1-1978 (which supersedes ANSI 18.1) qualified Health Physics Technicians are provided to man each shift. Radiation Protection Training for all other employees is referenced in Section 13.2.

12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES The design of facilities including restricted areas or equipment for use in restricted areas are periodically reviewed by the health physics staff to ensure that provisions are included to maintain ALARA exposures during maintenance, in-service inspection, refueling, and non-routine operations. The specific provisions are listed in Subsection 12.1.2.

12.5.2.1 Counting Room A low background counting room is located on elevation 19.5 ft in the Reactor Auxiliary Building (refer to Figure 1.2-13). This facility includes both laboratory and shielded counting room. It is equipped to analyze routine air samples and contamination swipe surveys. Portable radiation survey instruments, respiratory protection equipment and contamination control supplies are stored in the Reactor Auxiliary Building.

Counting room equipment is listed on Table 12.5-1.

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UFSAR/St. Lucie - 2 The criteria for selection of instrumentation ensures reliable and expeditious sample counting, low background and sensitivities. All laboratory equipment is checked and calibrated at regular intervals per plant procedures using standard radioactive sources traceable to a National Institute of Science and Technology (NIST) source.

12.5.2.2 Portable Instruments The portable instruments used for radiation monitoring are listed on Table 12.5-2.

The portable gamma survey instruments are calibrated in accordance with plant procedures utilizing a gamma calibration unit. The dose rates on the surface of the calibrator do not exceed 2.0 mR/hr., while the dose rates inside the calibrator range from 2 mR/hr to 600 R/hr.

Additionally, the use of a pulse generator may be used for count rate instrument calibration.

The criteria for selection of portable instrumentation was to obtain accurate and reliable instrumentation that is easily serviced and maintains a high degree of operability under adverse conditions and covers the entire spectrum of radiation measurements made during normal operation, during shutdown, and during accident conditions.

12.5.2.3 Personnel Monitoring Instrumentation Records of radiation exposure history and current occupational exposure are maintained for each employee for whom personnel dosimeters are issued. In addition to the routine review of EC291105 dosimeter of legal record (DLR) data, a periodic review of airborne exposure, beta gamma area radiation and contamination level data is made by Health Physics to evaluate exposure trends.

a. Portal monitors or friskers are used for checking possible contamination upon leaving a highly contaminated area and before leaving the Radiation Controlled Area.

b. Self-Reading Dosimeters EC291105 Quantities of self-reading dosimeters available are listed in Table 12.5-3.

Self-reading or digital alarming dosimeters are issued to all personnel entering the Radiation Controlled Area. Exposure results are normally recorded by the individual on a daily basis.

EC291105 Self-reading dosimeters are calibrated semi-annually.

c. Dosimeters of Legal Record (DLR)

The official permanent record of accumulated external radiation exposure EC291105 received by individuals is obtained from DLRs.

All persons subject to occupational radiation exposure are issued beta-gamma and neutron sensitive DLRs and are required to wear them at all times inside the EC291105 RCA. The DLRs are processed on a routine basis by a NVLAP accredited processor. The DLR of any individual is processed by special handling whenever it appears that an overexposure may have occurred. Special or additional personnel monitors are issued as may be required under unusual conditions.

These devices are issued at the discretion of health physics personnel.

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UFSAR/St. Lucie - 2 Visitors accessing the Radiation Controlled Area are escorted by qualified personnel and are issued appropriate personnel monitoring devices. An escort is not required for those who have received the necessary radiation protection training when this arrangement is approved by the Radiation Protection Manager and authorized by the Plant Manager.

d. Bioassay and Whole Body Counting EC291105 An initial bioassay sample, whole body count or equivalent is performed for radiation workers entering the Radiation Controlled Area to determine a base-line for further reference. Any time there is an incident which produces unknown concentrations of airborne radioactivity or when it is suspected that personnel have taken radionuclides into their body by some other means, a bioassay sample or whole body count is analyzed for uptake of radionuclides. If a significant increase of radionuclides present in the body is reported, a second bioassay sample and/or whole body count may be required.

A bioassay sample, whole body count or equivalent is analyzed for radiation EC291105 workers who are terminating employment. Whenever necessary, a service is contracted with a reliable vendor to provide a complete bioassay analysis program and submit a report after the sample is analyzed. A whole body counting facility is available at the St. Lucie Plant.

e. Hand and Foot Monitors Portable personnel radiation monitors are used as determined by Health Physics in accordance with plant procedures.

f. Contamination Monitors Portal monitors are located at the main guard station exits. These monitors provide a final radiation survey of all personnel leaving the generating station area.

External monitoring equipment for detection of personnel contamination is selected to best determine and reduce the potential for the spread of contamination. This equipment includes friskers as well as personnel contamination monitors that under normal operation have the sensitivity of a whole body frisk. Whole body contamination monitors are typically used by EC291105 personnel when exiting the Radiation Controlled Area.

The whole body counter is selected and located to ensure an accurate and immediate determination of any potential internal uptake of radionuclides.

12.5.2.4 Protection Equipment

a. Protective Clothing The fundamental purpose of protective clothing is to keep contamination away from the body of the wearer and thereby reduce the potential for internal deposition of radioactive materials and to help prevent the spread of 12.5-4 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 contamination. The protective clothing required for a particular job is prescribed by health physics and a radiation work permit, based on actual or potential radiological conditions.

EC291105 An adequate inventory of protective clothing is maintained. Supplies may be located at other areas to facilitate job activities. Protective clothing may consist of cap, hood, lab coat, coveralls, cotton glove liners, rubber gloves, cotton booties/plastic bag shoe covers, disposable shoe covers, rubber shoe covers, plastic suits, plastic hoods. Other equipment may be specified by health physics, such as mops, absorbent paper, plastic sheets and bags, ropes, signs, and labels.

b. Respiratory Protection Equipment Respirator devices available for use include:

1) Full-face respirator (filter, gas canister, or supplied air)

2) Self-contained breathing apparatus EC291105

3) Powered air purifying respirator (PAPR)

4) Bubble hoods Self-contained or supplied air breathing apparatus is used when respiratory protective devices are required in situations involving exposure to gaseous activity or oxygen deficient atmospheres.

The following tabulation of airborne concentrations expected to be in excess of Derived Air Concentrations (DAC) specified in 10 CFR 20 is used to determine the appropriate type of respiratory protection equipment required.

Airborne Concentration Type Particulate activity less than 50 times Full-face respirator (filter type)

DAC EC291105 Particulate activity greater than 50 times Self-contained, supplied air DAC breathing apparatus, or PAPR Respirators are maintained by checking for mechanical defects, contamination, and cleanliness by health physics trained personnel.

All individuals requiring the use of respirators have an annual medical exam and are properly fitted every twelve months for a specific respirator. Checking of respirator fit in the field is performed according to procedures.

12.5.2.5 Support Facilities

a. An area is designed, located, and shielded such that radiation in the calibration area does not interfere with low level monitoring or counting systems.

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b. The general arrangement of the locker room facilities in the Reactor Auxiliary Building is designed to provide adequate personnel decontamination and change areas as shown on Figure 1.2-13. The cold locker room is used as a change and storage area for clothing and personal items not required or allowed in the radiation controlled areas. The hot locker room is employed as a change area and storage area for potentially contaminated clothing. Personnel contamination monitors are located at the access point(s). All personnel monitor themselves on leaving the Radiation Controlled Area. Showers, sinks and necessary monitoring equipment also are provided in the hot locker room to aid in the decontamination of personnel.

c. The equipment decontamination area is located near the hot machine shop since most of the need for decontamination occurs prior to utilization of the shop facilities. The decontamination area (Figure 1.2-13) is equipped to handle the decontamination of small and medium sized equipment and tools.

EC291105 Respirators are cleaned and decontaminated as necesscary.

d. The Radiation Controlled Area is a restricted area as defined by 10 CFR 20 and is shown on Figures 12.3-4, 12.3-7 through 10 and 12.3-13. This area includes that in which radioactive materials and radiation above 2 mrem/hr may be present. Traffic is routed into and out of the Radiation Controlled Area in a path that will minimize exposure and spread of contamination.

Control points allow adequate control of access and occupancy for radiation protection purposes. All radioactive materials are checked prior to exit from the Radiation Controlled Area.

e. A health physics office is located next to the Control Point to accommodate the health physics staff, permanent records, and plant procedures. Monitoring equipment is maintained in the health physics counting room.

EC291105

f. The Health Physics Records are computerized and centralized.

12.5.3 PROCEDURES The health physics staff prepares written procedures and letters of instructions to implement the Radiation Protection Program.

Time-radiation dose schedules are followed to formalize operations to be executed in unusually high gamma radiation fields in such a manner that exposure to operating and maintenance personnel is minimized.

Health Physics personnel periodically observe jobs in progress in the Radiation Controlled Area and make daily radiation surveys to ensure that exposure to radiation and contamination levels are kept ALARA.

Administrative exposure guidelines are designed to evenly distribute each individual's annual exposure. When personnel are assigned to a job or a location where there exists the possibility that administrative guidelines may be exceeded, Health Physics investigate the exposure records of the personnel involved and make an authorization accordingly. This authorization can 12.5-6 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 be given by Health Physics only after the individual's current exposure history, the amount of guidelines will be exceeded, and the alternatives that are available to complete the job under consideration have all been considered. EC291105 The events or incidents which necessitate special radiation surveys include, but are not limited to the following:

a. spills involving radioactive materials.

b. a significant increase in radiation levels as noted on area monitors, portable radiation detection instruments, or air monitors.

c. radiation work permit request in areas where the potential change of radiation and contamination levels are high.

All personnel entering contaminated areas are required to wear protective clothing. The nature of the work to be done is the governing factor in the selection of protective clothing to be worn by individuals. The protective apparel available includes shoe covers, head covers, gloves, and coveralls or lab coats. Additional items of specialized apparel such as plastic or rubber suits, face shields, safety glasses and respirators are also available. Health physics-trained personnel evaluate the radiological conditions and specify the required items of protective clothing to be worn on the radiation work permit.

Appropriate written procedures govern the proper use of protective clothing, where and how it is to be worn and removed, and how the change room and decontamination facilities for personnel equipment, and plant areas are to be used.

Provisions are made for decontamination of work areas throughout the plant. The Cask Handling Facility has services to decon large equipment but the Cask Handling Facility was specifically designed to support Dry Cask Storage. A decontamination room external to the machine shop in the Reactor Auxiliary Building is used for the decontamination of hand tools and small equipment.

Respiratory protective devices are required in any situation arising from operations in which airborne radioactivity exists or is expected in unknown concentrations or in known concentrations which would cause personnel to exceed exposure limits specified by 10 CFR 20.

In such cases the actual or potential airborne activity is evaluated by health physics trained personnel and the necessary protective devices specified.

Survey instruments are calibrated periodically and maintenance records are maintained for each instrument.

In order to protect personnel from access to high radiation areas that may exist temporarily as a result of plant operations and maintenance, warning signs, visual indicators, barricades, and locked doors are used as necessary.

The requirements for controlling entry into high radiation areas are contained in the Technical Specifications. These requirements are implemented by plant procedures.

Contaminated areas and equipment are decontaminated as soon as practicable after detection or use by personnel trained in decontamination procedures. Prior to decontamination, each contaminated area in which the contamination levels are equal to or greater than the plant 12.5-7 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 administrative limits are barricaded and conspicuously posted as a contaminated area and entrance hereto is controlled by requiring issuance of a Radiation Work Permit.

Decontamination is performed under the direction of health physics personnel.

12.5.3.1 Radiation Surveys Health physics personnel and health physics qualified personnel normally perform radiation surveys, the techniques of which are delineated in plant health physics procedures. Surveys are performed on frequencies that vary with the potential radiological hazards associated with a given area. Frequencies are also delineated in plant health physics procedures.

Surveys are normally for radiation level, contamination level, and/or airborne radionuclide concentration determination. Records of all surveys are maintained including current survey EC291105 information for the restricted areas. Survey information is factored into exposure stay time determination and radiation work permit specifications (see Subsection 12.5.3.4).

Radiation level surveys may be performed for alpha, gamma, beta, and/or neutron exposure rates. Contamination surveys are normally performed to establish gross beta-gamma contamination level, but may be processed for specific types of radiation (beta-alpha-gamma) or specific radionuclides (via gamma spectroscopy). Air samples are normally taken to establish airborne concentrations of particulates and/or radioiodine, but specific nuclide information may also be obtained. Availability of current survey information will aid in keeping exposures ALARA.

12.5.3.2 Procedural Methods To Maintain Exposures ALARA The ALARA policy is delineated in Section 12.1. ALARA considerations are incorporated into various types of plant procedures in addition to health physics procedures.

12.5.3.2.1 Refueling Some examples of procedural methods of maintaining exposures ALARA during refueling are:

a. refueling cavity water is filtered to remove radioactive material,

b. prior to removing the vessel head, the primary system is degassed and sampled to minimize expected airborne levels when the head is removed,

c. movement of irradiated fuel assemblies is accomplished with the assembly maintained under water,

d. work performed in the controlled area is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments,

e. current survey information is used,

f. ventilation is provided to minimize airborne radioactive material, and

g. the radiation work permit system is used to maintain positive radiological control over work in progress.

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UFSAR/St. Lucie - 2 12.5.3.2.2 In-service Inspection Some examples of procedural methods of maintaining exposures ALARA during in-service inspection are:

a. Equipment is calibrated and checked prior to entry into the radiation area.

b. Portable shielding is used where practicable.

c. Work performed in the controlled area is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.

d. Current survey information is used.

e. Ventilation is provided to minimize airborne radioactive material.

f. The radiation work permit system is used to maintain positive radiological control over work in progress.

12.5.3.2.3 Radwaste Handling Some examples of procedural methods of maintaining exposures ALARA during radwaste handling are:

a. The volume of radwaste handling is minimized by station design.

b. Radwaste systems are shielded and located so that operator and other personnel exposure is minimized.

c. Extension tools are used when practical.

d. Portable shielding is available for use as necessary.

e. Ventilation is provided to minimize airborne radioactive material from waste handling operations.

f. Current survey information is used.

g. Drums are labeled prior to filling with waste.

12.5.3.2.4 Spent Fuel Handling, Loading, and Transfer Some examples of procedural methods of maintaining exposures ALARA during spent fuel handling are:

a. The spent fuel pool water is filtered to remove radioactive material.

b. The spent fuel pool water is cooled and surface air ventilation is provided to minimize airborne radioactive material.

c. Loading of the transfer cask is performed under water.

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d. Fuel handling cranes and extension tools are used to handle transfer casks, fuel assemblies, and inserts.

e. Movement of irradiated fuel assemblies is accomplished with the assembly maintained under water.

f. Work performed in the controlled area is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.

g. Current survey information is used.

h. The radiation work permit system is used to maintain positive radiological control over work in progress.

i. Ventilation is provided to minimize airborne radioactive material.

j. After the transfer cask is loaded, it is decontaminated using a pressurized water washing device.

12.5.3.2.5 Normal Operation Some examples of procedural methods of maintaining exposures ALARA during normal operation are:

a. The station is designed so that sources of radiation are shielded.

b. An area radiation monitoring system is available and provides indication of radiation levels and, as applicable, local and/or remote alarms.

c. Work performed in the controlled area is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.

d. Current survey information is used.

e. Ventilation is provided to minimize airborne radioactive material.

f. The radiation work permit system is used to maintain positive radiological control over work in progress.

g. During initial start-up, neutron and gamma dose rate surveys are performed to verify shielding adequacy.

h. Areas are conspicuously posted in accordance with 10 CFR 20.

12.5.3.2.6 Maintenance Some examples of procedural methods of maintaining exposures ALARA during maintenance are:

a. Equipment is moved to areas with lower radiation and contamination levels for maintenance when practicable.

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b. Extension tools are used when practical.

c. Portable shielding is used as practical.

d. Work performed in the controlled area is staged, i.e., workers are briefed on assignments and familiar with procedures and equipment needed to complete assignments.

e. Current survey information is used.

f. The radiation work permit system is used to maintain positive radiological control over work in progress.

g. Routine maintenance is proceduralized and precautions specified.

h. Required tools are specifically listed in procedures where practical.

12.5.3.2.7 Sampling Some examples of procedural methods of maintaining exposures ALARA during sampling are:

a. Sampling hoods are provided in the radiochemistry laboratory. Ventilation minimizes airborne radioactive material. The sampling hoods are located to reduce the exposure from sampling of radioactive liquids and gases.

b. Procedures specify protective clothing and proper sampling methods.

c. Radiation levels of samples are checked.

d. Extension tools are used when practicable.

12.5.3.2.8 Calibration Some examples of procedural methods of maintaining exposures ALARA during calibration are:

a. The instrument calibrator is heavily shielded.

b. An interlock is provided so that the calibrator door cannot be opened while sources are exposed.

c. Portable sources used to calibrate fixed instruments are transported and maintained in shielded containers.

d. The radiation work permit system is used, where applicable, to maintain positive radiological control over calibration.

12.5.3.3 Access Control Access to the radiological controlled area is normally through the main access control point.

Control points are established in plant areas as necessary (such as at the containment entrance during refueling outages), and are also manned as delineated in plant procedures. These control points provide positive radiological control over personnel entering controlled areas.

Only personnel with current authorization, such as a radiation work permit (see Subsection 12.5.3.4), may pass control points.

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UFSAR/St. Lucie - 2 High radiation areas (as defined in 10 CFR 20) also have access control features delineated in plant procedures.

12.5.3.4 Radiation Work Permit All work in radiation areas, high radiation areas, or any radiologically hazardous areas (as determined by health physics personnel) requires a radiation work permit. The radiation work permit (RWP) provides appropriate radiological controls and instructions for work to be performed safely, efficiently, and within the ALARA commitment.

12.5.3.5 Contamination Control Contamination limits for personnel, equipment, and areas are delineated in plant procedures.

Surveys for contamination control are performed by health physics personnel on a routine basis at various locations in the plant (see Subsection 12.5.3.1). Nonroutine surveys are performed in areas whenever a change in contamination levels is likely and may be important for radiation protection. Areas found contaminated are posted, isolated (with ropes, barriers, etc.) and decontaminated as practical. Since the complete removal of surface contamination from some plant areas is not practical, these areas may be designated as contamination areas. The level of contamination and number of such areas is minimized. Control points are established with step-off pads, and the area is posted and isolated. Entrance to such an area normally requires authorization of, and adherence to the specifications of a radiation work permit.

Tools and equipment used in potentially contaminated areas are surveyed for removable contamination and fixed contaminated tools are bagged for transportation. If tools or equipment do not meet the clean area limits, they are decontaminated before leaving the controlled area.

Some tools and equipment are for use only in a contaminated area. These items are also surveyed periodically and decontaminated as appropriate. They are considered to be contaminated and are used only by personnel in anti-contamination clothing.

Personnel are protected from contamination by the protective clothing and equipment specified in radiation work permits. Personnel survey themselves for contamination upon exiting a contamination area. In addition, when personnel pass through the access control point they pass through a personnel contamination monitor. Contaminated personnel are decontaminated at the decontamination facility under the supervision of health physics personnel.

12.5.3.6 Radiation Protection Training Plant personnel, both permanent and temporary, whose duties require such training, are instructed in the fundamentals of radiation protection. Personnel must be acceptably cognizant of fundamentals presented in training to enter the radiologically controlled areas unescorted.

Training topics include: instructions in applicable station and NRC exposure limits, station procedures, instructions to women concerning prenatal exposure, properties of radiation and radioactivity, biological effects of exposure, techniques of radiation protection, ALARA, emergency and fire alarm response, and other topics as pertinent. More detail on the training program is given in Section 13.2. Additional training is given to plant personnel whose duties involve greater degrees of radiological hazard, such as health physics personnel and operators.

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UFSAR/St. Lucie - 2 12.5.3.7 Personnel Monitoring 12.5.3.7.1 External Radiation Exposure The official permanent record of accumulated external radiation exposure received by EC291105 individuals is obtained from dosimeter of legal record (DLR).

All persons subject to occupational radiation exposure are issued beta-gamma and neutron sensitive DLRs and are required to wear them at all times inside the RCA. Extremity DLRs are issued to workers when extremity doses are anticipated to exceed doses specified in approved procedures. DLRs are processed on a routine basis by a NVLAP accredited processor. EC291105 The DLRs of any individual is processed by special handling whenever it appears that an over exposure may have occurred or whenever a self-reading dosimeter result is considered questionable.

Special or additional personnel monitors are issued as may be required under unusual conditions. These devices are issued at the discretion of health physics personnel.

Visitors, accessing the Radiation Controlled Area are escorted by qualified personnel and are issued appropriate personnel monitoring devices.

12.5.3.7.2 Internal Radiation Exposure Internal Radiation Exposure assessment at the plant is performed using ANSI N343 (1978),

"Internal Dosimetry for Mixed Fission and Activation," as described in Regulatory Guide 8.26 (R0).

EC291105 An internal bioassay sample, whole body count or equivalent is performed for radiation workers entering a radiation controlled area to determine a base-line for further reference. Anytime there is an incident which produces unknown concentrations of airborne radioactivity or when it is suspected that personnel have taken radionuclides into their body by some other means, a bioassay sample or whole body count is analyzed for uptake of radionuclides. If a significant increase of radionuclides present in the body is reported, a second bioassay sample and/or a whole body count may be required.

EC291105 All plant employees requiring access to a radiation controlled area require a whole body count, a bioassay sample collected or equivalent on an annual basis. All other individuals have the required exam performed upon termination.

12.5.3.8 Airborne Radionuclide Control, Assessment, and Personnel The plant ventilation systems (refer to Subsection 6.5.1 and Section 9.4) provides the means for removing airborne radioactive material from the in-plant atmosphere. Airborne radionuclide concentrations are controlled by minimizing loose surface contamination levels and providing containment of sources.

Concentrations of airborne radionuclides are routinely assessed by fixed continuous air monitors and air sample surveys. Air sample surveys are taken routinely at specified frequencies and non- routinely when the potential for personnel exposure exists (as determined by health physics personnel). Radiation work permits may specify air sampling prior to the start 12.5-13 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 of work in a given area. Continuous air monitors alarm when airborne radionuclide concentrations exceed present values in a given area. Thus internal exposures are minimized by this assessment and follow up control.

Routine iodine monitoring is accomplished by use of the continuous air monitors and by grab EC291105 sampling using charcoal as the absorption media. Grab Sample Analysis of radioiodines is by use of a GeLi or high purity Ge detector and multi-channel analyzer.

Monitoring of post-accident radioiodines consist of the same systems as described for routine sampling except that Silver Zeolite cartridges are used as the absorption media and these are purged with clean air prior to counting.

There may exist areas in which airborne radionuclide concentrations cannot be maintained below applicable station limits (normally, these are the Derived Air Concentrations found in 10 CFR 20). If personnel entry to these areas is required, either stay time is restricted (such as for exposures to airborne radionuclides) or respiratory protection is provided. The radiation work permit for the work to be performed specifies the applicable equipment and/or stay time to maintain exposures ALARA.

Respiratory protection equipment is available at the access control point and other locations.

For normal use, equipment may be taken from the supplies at the access control point.

Equipment is maintained, inspected and used in accordance with Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection" October 1976 (R0).

To assure an adequate program for respiratory protection, the following controls are incorporated into the program:

a. Each respirator user is advised that he may leave an airborne radioactivity area for psychological or physical relief from respirator use. Each user must leave the area in the case of respirator malfunction or any other condition that might cause reduction in the protection afforded the user.

b. Sufficient air samples and surveys are made to identify the various radionuclides present and to estimate the individual exposures so that selection of appropriate respiratory equipment can be made in accordance with 10 CFR 20.1703.

c. Training procedures are established to assure correct fitting, use, maintenance, and cleaning of the various types of respiratory equipment. Each employee EC291105 wearing a face-sealing separator is individually fitted prior to use.

d. Bioassays, urinalysis and whole body counts (or equivalent) may be made in accordance with plant procedures and, as required, to evaluate individual body burdens of radionuclides and to assess the overall effectiveness of the respiratory protection program.

12.5.3.9 Radioactive Material Safety Program Radioactive material may be used by station personnel for calibration and other purposes. This material includes both sealed sources and unsealed materials (gaseous, liquid, or solid).

Calibration of radiochemistry counting, fixed monitoring, and portable survey instrumentation is the most common use of such material. Sealed sources that are exempt quantities do not 12.5-14 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 require special handling procedures for radiation protection purposes. Plant procedures control the use, inventory and leak testing of all sealed sources. This also applies to exempt quantities or exempt concentrations of unsealed material.

Recognized methods for the safe handling of radioactive materials, such as those recommended by the National Council of Radiation Protection and Measurement, are proceduralized to ensure proper usage. Procedures specifying handling techniques, storage, and other safety considerations, are listed below:

a. minimizing distances that large radioactive sources are transported,

b. use of shielded transporters,

c. storage of sources in appropriately shielded containers,

d. proper labeling of all radioactive material (per 10 CFR 20),

e. inventorying of all radioactive sources in accordance with plant procedures,

f. leak testing of sources at six month intervals in accordance with license conditions, and

g. monitoring of all packages received containing radioactive material in accordance with the requirements of 10 CFR 20.

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UFSAR/St. Lucie - 2 TABLE 12.5-1 COUNTING ROOM EQUIPMENT Minimum Quantity Description 1 Pulse Height Analyzer System with GeLi or high purity Ge Detector for isotopic analysis 1 Liquid Scintillation Counter for tritium analysis (Service provided EC291105 by Chem. Dept.)

2 Gas Flow Proportional Counters for gross beta gamma-alpha analysis 1 End Window G-M type Counter T12.5-1 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 TABLE 12.5-2 PORTABLE INSTRUMENTS FOR RADIATION MONITORING APPROXIMATE DETECTION QUANTITY RANGE OR CAPACITY TYPE(S) OF RADIATION EFFICIENCIES OR ENERGIES 32 Count Rate Meters G.M. 10% for - EC291105 0-400,000 cpm Beta & Gamma 0.15 MeV to 2.2 MeV 6 Intermediate Range G.M. Gamma 60 KeV - 1.5 MeV 0-1,000 R/hr Gamma 12 Intermediate Range Ion Beta & Gamma -12 KeV - 3.0 MeV Chambers 0-50 R/hr EC291105 6 Detector and Moderator Neutron 0.025eV - 10 MeV 0-5,000 mRem/hr 50 cpm 1 mRem/hr 6 High Range G.M. Gamma 60 KeV - 3.0 MeV EC291105 0.1 mR/hr - 12,000 R/hr Multiple Ranges EC291105 4 Not Applicable Beta Air Detect 0.3 DAC for 60Co Monitoring 1.0x10-12Ci/cc for 90Sr 3 0 - 400,000 cpm Alpha 30% for 239Pu EC291105 Multiple Range EC291105 6 0-8 SCFM Air Sampler >90% for Particulates and Iodines T12.5-2 Amendment No. 26 (09/20)

UFSAR/St. Lucie - 2 TABLE 12.5-3 SELF-READING DOSIMETERS EC 291 105 APPROXIMATE QUANTITY RANGE TYPE ENERGY EC 291 400 1mrem- Low to High 80KeV-2MeV 105 999 Rem Range Digital Alarming Dosimeter T12.5-3 Amendment No. 26 (09/20)