View source for L-2016-198, Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 11, Waste Disposal and Radiation Protection System

{{Adams
| number = ML16330A235
| issue date = 10/29/2016
| title = Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 11, Waste Disposal and Radiation Protection System
| author name = 
| author affiliation = Florida Power &amp; Light Co
| addressee name = 
| addressee affiliation = NRC/NRR
| docket = 05000250, 05000251
| license number = 
| contact person = 
| case reference number = L-2016-198
| package number = ML16330A191
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 121
}}

=Text=
{{#Wiki_filter:TABLE OF CONTENTS 

Section       Title       Page

11 WASTE DISPOSAL AND RADIATION PROTECTION SYSTEM 11.1-1 

11.1 Waste Disposal System 11.1-1 11.1.1  Design Bases 11.1-1    Control of Releases of Radioactivity    to the Environment 11.1-1 11.1.2  System Design and Operation 11.1-2    System Description 11.1-3 Liquid Processing 11.1-3 Gas Processing 11.1-5 Solids Processing 11.1-7 Components 11.1-9 Laundry and Hot Shower Tanks 11.1-9 Reactor Coolant Drain Tanks 11.1-9 Waste Holdup Tanks 11.1-10 Spent Resin Storage Tank 11.1-10 Gas Decay Tanks 11.1-10 Compressors 11.1-10 

Waste Monitor Tanks 11.1-11     Monitor Tanks 11.1-12     Waste Disposal Demineralizer 11.1-12       Nitrogen Manifold 11.1-13     Hydrogen Manifold 11.1-13     Gas Analyzer 11.1-14  

11-i Rev. 16  10/99 

TABLE OF CONTENTS (Continued) 

Section     Title        Page

Pumps  11.1-14     Piping 11.1-14     Valves 11.1-14 11.1.3  Design Evaluation 11.1-15 Liquid Releases 11.1-15 Liquid Wastes Without Primary - Secondary Leakage 11.1-15 Liquid Wastes With Primary - Secondary Leakage 11.1-17 Gaseous Wastes 11.1-18 Gaseous Release Rate 11.1-20 Solid Wastes 11.1-22 

11.2 Radiation Protection 11.2-1 11.2.1  Design Bases 11.2-1    Monitoring Radioactivity Releases 11.2-1 Monitoring Fuel and Waste Storage 11.2-1 Fuel and Waste Storage Radiation Shielding 11.2-2 Protection Against Radioactivity Release     from Spent Fuel and Waste Storage 11.2-2 11.2.2  Primary and Secondary Shielding 11.2-3    Design Basis 11.2-3 Primary Shield 11.2-4 

Secondary Shield 11.2-5     Accident Shield 11.2-5     Fuel Handling Shield 11.2-5     Auxiliary Shielding 11.2-5    Shielding Design 11.2-6     Primary Shield 11.2-6     Secondary Shield 11.2-6     Accident Shield 11.2-7     Fuel Handling Shield 11.2-8     Auxiliary Shielding 11.2-9 11.2.3  Radiation Monitoring System 11.2-10    Process Radiation Monitoring System 11.2-11     Containment High Range Radiation Monitors (RaD-3-6311A & B, RaD-4-6311A & B) 11.2-15     Containment Air Particulate Monitors    (R3-11 & R4-11) 11.2-15     Containment Radioactive Gas Monitors (R3-12 & R4-12) 11.2-17     Plant Vent Gas Monitors (R-14 & RaD 6304) 11.2-18  

11-ii Revised 04/17/2013 C26 TABLE OF CONTENTS (Continued)
Section     Title       Page

Condenser Air Ejector Monitors (R3-15, R4-15, RaD-3-6317 & RaD-4-6317) 11.2-19     Component Cooling Liquid Monitors (R3-17A, R3-17B, R4-17A & R4-17B) 11.2-20     Waste Disposal System Liquid Effluent Monitor (R-18)  11.2-20     Steam Generator Liquid Sample Monitors (R3-19 &
R4-19)  11.2-20     Main Steam Line Monitors (RAD 6426) 11.2-21     Reactor Coolant Letdown Line Activity Monitors  (R3-20 & R4-20) 11.2-22     Spent Fuel Pool Vent Monitor - Unit 3 (RAD-3-6418) 11.2-22 Area Radiation Monitoring System 11.2-23    System Description 11.2-23     The Detector 11.2-24     The Local Indicator 11.2-24     The Remote Indicator 11.2-25     Radiation Monitoring System Cabinet 11.2-25    Health Physics Program 11.2-26     Facilities and Access Provisions 11.2-26     Personnel Monitoring 11.2-28     Personnel Protective Equipment 11.2-29     Monitoring Instrumentation 11.2-31 11.2.4  Evaluation 11.2-31    Evaluation of Vital Area Access Outside the Control Room 11.2-32  11.2.5  Tests and Inspection Capability 11.2-35  11.2.6  References 11.2-36  11.3 Radioactive Materials Safety 11.3-1 11.3.1  Materials Safety Program 11.3-1 11.3.2  Facilities and Equipment 11.3-3 11.3.3  Personnel and Procedures 11.3-4 11.3.4  Required Materials 11.3-5 11.4 Radiological Administrative Controls 11.4-1 11.4.1  In-Plant Radiation Monitoring 11.4-1 11.4.2  Radiological Environmental Monitoring Program 11.4-1 11.4.3  Radiation Protection Program 11.4-1 11-iii Revised 04/17/2013 C26 LIST OF TABLES 

Table       Title

11.1-1 Waste Disposal System Performance Data (Two Units) 

11.1-2 Waste Disposal Components Code Requirements 

11.1-3 Component Summary Data 

11.1-4 Estimated Liquid Discharge to Waste Disposal 

11.1-5 Estimated Liquid Release by Isotope (Two Units) 

11.1-6 Estimated Annual Gaseous Release by Isotope (Two Units) 

11.2-1 Radiation Zone Classifications 

11.2-2 Primary Shield Neutron Flux and Design Parameters 

11.2-3 Original Secondary Shield Design Parameters 11.2-4 Deleted 11.2-5 Deleted 

11.2-6 Principal Auxiliary Shielding 

11.2-7 Radiation Monitoring System Channel Sensitivities 

11.2-7a Deleted 

11.2-8 Detecting Medium Conditions 

11.2-9 Portable Radiation Survey Instruments 

11.2-10 Deleted 11.2-11 LOCA Activity Source in Circulating Residual Heat Removal Loop and Associated Equipment 11.2-12 Control Room Direct Shine Shielded Dose Results Using AST 11.2-13 Vital Area Access Mission Doses 

11.3-1 Byproduct, Source and Special Nuclear Materials; Radioactive Sources Listing  

11-iv Revised 04/17/2013 

C26C26 LIST OF FIGURES Figure                               Title 11.1-1 Liquid Waste Disposal System - Reactor Coolant Drain Tank and Pumps  (Unit 3) 11.1-1a Deleted 

11.1-1b Deleted 

11.1-1c Deleted 

11.1-2 Liquid Waste Disposal System - Containment Drains (Unit 3) 

11.1-2a Deleted 

11.1-2b Deleted 

11.1-3 Disposal of Radioactive Liquids 

11.1-4 Liquid Waste Disposal System - Reactor Coolant Drain Tank and Pumps  (Unit 4) 11.1-5 Primary - Secondary Activity Relations Study 

11.1-6 Deleted 

11.1-7 Radwaste Solidification System - Cement Handling and Container Filling 

11.1-8 Liquid Waste Disposal System - Containment Drains (Unit 4) 

11.1-9 Liquid Waste Disposal System - Waste Holdup & Transfer 

11.1-10 Liquid Waste Disposal System - Laundry Waste 

11.1-11 Liquid Waste Disposal System - Drain Headers and Sumps 

11.1-12 Liquid Waste Disposal System - Polishing Demineralizer 

11.1-13 Liquid Waste Disposal System - Waste Evaporator Feed 

11.1-14 Liquid Waste Disposal System - Waste Evaporator Package 

11.1-15 Liquid Waste Disposal System - Liquid Sampling, Monitoring,  and Chemical Addition 

11.1-16 Liquid Waste Disposal System - Waste Monitor Tanks 

11.1-17 Solid Waste Disposal System - Spent Resin Storage 

11.1-18 Solid Waste Disposal System - Holdup and Mixing 

11.1-19 Solid Waste Disposal System - Container Fill 

11.1-20 Gaseous Waste Disposal System - Waste Gas Compressors 

11-v Rev. 13  10/96 

LIST OF FIGURES 

Figure       Title

11.1-21 Gaseous Waste Disposal System - Waste Gas Decay Tanks 

11.1-22 Gaseous Waste Disposal System - Gas Waste Analyzers 

11.2-1 General Arrangement Ground Floor Plan El. 18'-0" 

11.2-2 Area Radiation Zone Plan - Full Power Operation with 1% Failed Fuel 

11.2-3 General Station Area 

11.2-4 LOCA Recirculating Piping T=0 hr Dose rate and 31-day Integrated Dose  11.2-5 Deleted 

11-vi Revised 04/17/2013 C26 11 WASTE DISPOSAL AND RADIATION PROTECTION SYSTEM 11.1 WASTE DISPOSAL SYSTEM

The system is designed to process wastes from both Units 3 and 4 and the term "plant" refers to these two nuclear units. 

11.1.1 DESIGN BASES

Control of Releases of Radioactivity to the Environment

Criterion: The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment (10CFR Part50, Appendix A, Criterion 60)*.  

The limits placed on plant radioactive effluent release by 10 CFR 20 and 10 

CFR 50.67 have been considered in the design and operating plans for the 

plant, with the objective to maintain release concentration at the site 

boundary below natural background activity and thus only a minute fraction of 

10 CFR 20 limits. To achieve these objectives the facility has been designed 

and will be operated as follows: 
: 1. Liquid wastes will be collected in tanks and processed by the waste disposal demineralizers. The waste process provided can reduce activity 

well below established limits and represents a design for reducing 

activity to the lowest practicable value. Analyses of liquid prepared 

for release will be made to determine that activity levels have been 

minimized before release is permitted. The resulting activity after 

mixing with the
* Letter L-83-499, Amendment 103 and 97 - Radiological Effluent Technical Specifications and Radiological Environmental Monitoring", dated September 26, 1983. 

11.1-1 Revised 04/17/2013 C26C26C26 circulating water will be near to or equal to natural background. The tritium is expected to be about 1% of MPC. 
: 2. Gaseous wastes will be stored in decay tanks for natural decay. Gases will be released through the monitored plant vent, and at the site boundary the annual dose will be a small fraction of 10 CFR 20 limits.
Cover gases in the nitrogen blanketing system will be reused to minimize the number of tanks released. 

The quantity of radioactivity contained in each gas decay tank is restricted to provide (a) assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem, and (b) assurance that the concentration of potentially explosive gas mixtures contained in the Gas Decay Tank System is maintained below the flammability limits of hydrogen and oxygen. 
: 3. Solid radioactive wastes will be packaged to minimize the number of containers shipped. Low level waste packaged for shipment may be stored on site in the Low Level Waste Storage Facility while awaiting shipment. 

11.1.2  SYSTEM DESIGN AND OPERATION 

The Waste Disposal System Process Flow Diagrams are shown in Figures 11.1-1 through 11.1-18 and Performance Data are given in Table 11.1-1. The Waste Disposal System is common to Units 3 and 4 with the exception of the reactor coolant drain tanks and reactor coolant drain tank pumps. 

11.1-2 Revised 12/05/2014 C27 The Waste Disposal System collects and processes potentially radioactive reactor plant wastes prior to release or removal from the plant site within 

limitations established by applicable governmental regulations. The fluid 

wastes are sampled prior to release using an isotopic identification as 

necessary. Radiation monitors are provided to maintain surveillance over the 

release operation. Permanent record of Waste Disposal System releases is 

provided by radiochemical analysis of known quantities of waste. The system 

is capable of processing all wastes generated during continuous operation of 

the Reactor Coolant System assuming that fission products escape from one per 

cent of the fuel elements into the reactor coolant. 

At least two valves must be opened to permit discharge of liquid or gaseous 

waste from the Waste Disposal System. One of these valves is normally locked 

closed. During release, the effluent is monitored, and the release 

terminated if the radioactivity level exceeds a predetermined value.  

Activity release limits are given in the Offsite Dose Calculation Manual in 

accordance with the Technical Specifications. 

11.1-2a Rev. 16  10/99 As secondary functions, system components supply hydrogen and nitrogen to RCS components as required during normal operation and provide facilities to transfer fluids from inside the containment to other systems outside the containment. 

The waste disposal system is controlled primarily from a local control board in the auxiliary building and four local control boards in the radwaste facility with appropriate indicators and alarms. Off normal conditions are annunciated in the control room. All system equipment is located in or near the auxiliary building and in the radwaste facility except for the reactor coolant drain tank and pumps, which are located in the containment. 

System Description 

Liquid Processing 

During normal plant operation the Waste Disposal System can process liquids from the following sources: 

a) Equipment drains, floor drains, tank overflows, containment sumps, and leak-offs b) Laboratory drains c) Radioactive laundry and shower drains d) Decontamination area drains e) Resin transfer flush water  

f) Refueling water from fuel transfer canal and/or reactor cavity g) Holdup Tanks h) Miscellaneous sources via molybdate holding tank Additionally, each Unit's blowdown tank can be connected by a hose to the Waste Disposal System via the Waste Holdup Tank. 

The system also collects and transfers liquids directly from the following sources to the Chemical and Volume Control System for processing: 

a) Reactor coolant loop drains.
b) Reactor coolant pump seal leakage. c) Excess letdown during startup.
d) Accumulators.
e) Valve and reactor vessel flange leakoffs. 

11.1-3 Revised  07/15/2016 C28 These liquids flow to the reactor coolant drain tank and are discharged directly to the CVCS holdup tanks by the reactor coolant drain tank pumps 

which are operated automatically by a level controller in the tank. These 

pumps also return water from the refueling canal and cavity to the refueling 

water storage tank. There are one reactor coolant drain tank and two reactor 

coolant drain tank pumps inside each containment. 

Waste liquids are collected by various drains and sumps. The liquid drains 

flow by gravity, or are pumped, to the waste hold up tank (See Figure 

11.1-9). The activity level of waste liquid from the laundry area will 

usually be low enough to permit discharge from the site without processing.  

The liquid is pumped to one of the waste monitor tanks or monitor tanks where 

its activity can be determined for record before it is discharged through a  

radiation monitor. The liquid waste in the molybdate holding tank (item h, page 11.1-3) is typically pumped directly to the waste monitor tanks. 

The liquids requiring cleanup before release are processed by the waste 

disposal demineralizer. The liquid from the waste disposal demineralizer is 

routed directly to one of three radwaste facility waste monitor tanks or one  

of two monitor tanks. 

When one of the waste monitor tanks is filled, it is isolated, recirculated 

and sampled for analysis while one of the other two tanks is in service. If 

analysis confirms the activity level is suitable for discharge, the liquid is 

pumped through a flowmeter and a radiation monitor and then released to the 

circulating water system. 

11.1-4  Revised  5/10/2004 Otherwise, it can be returned to a waste holdup tank for reprocessing.
Although the radiochemical analysis forms the basis for recording activity 

releases, the radiation monitor provides surveillance over the operation by 

automatically closing the discharge control valve if the liquid activity 

level exceeds a preset value. 

Gas Processing

During plant operation, gaseous wastes originate from: 

a) degassing reactor coolant discharge to the Chemical and Volume Control System, b) displacement of cover gases as liquids accumulate in various tanks, c) miscellaneous equipment vents and relief valves, and d) sampling operations and gas analysis for hydrogen and oxygen in cover gases. 

During normal operation the Waste Disposal System supplies nitrogen from a 

Dewar vessel and hydrogen from a tube trailer to waste disposal components.  

Dual manifolds are provided, one for operation and one for backup. The 

system is sufficiently instrumented and alarmed to ensure continuous supply 

of gas.   

Most of the gas received by the Waste Disposal System during normal operation 

is cover gas displaced from the Chemical and Volume Control System holdup 

tanks as they fill with liquid (see Figures 11.1-16, 11.1-17 and 11.1-18).  

Since this gas must be replaced when the tanks are emptied during processing, facilities are provided to return gas from the decay tanks to the holdup 

tanks. A backup supply from the nitrogen header is provided for makeup   

11.1-5 Rev. 16  10/99 if return flow from the gas decay tanks is not available. To prevent 

hydrogen concentration from exceeding the combustible limit during this type 

of operation, components discharging to the vent header system are restricted to 

those containing no air or aerated liquids and the vent header itself is   

designed to operate at a slight positive pressure (1.0 psig minimum to 4.0 

psig maximum) to prevent in-leakage. On the other hand, out-leakage from the 

system is minimized by using Saunders patent diaphragm valves, bellows seals, self contained pressure regulators and soft-seated packless valves throughout 

the radioactive portions of the system.   

Gases vented to the vent header flow to the waste gas compressor suction 

header. One of the two compressors is in continuous operation with the 

second unit instrumented to act as backup for peak load conditions or failure 

of the first compressor. From the compressors, gas flows to one of the gas 

decay tanks. The control arrangement on the gas decay tank inlet header 

allows the operator to place one tank in service and to select one tank for 

backup if the tank in operation becomes fully pressurized. When the tank in 

service becomes pressurized to approximately 100 psig, a pressure transmitter 

automatically   

closes the inlet valve to that tank, opens the inlet valve to the backup tank 

and sounds an alarm to alert the operator of this event so that he may select 

a new backup tank. Pressure indicators are supplied to aid the operator in   

selecting the backup tank.   

Gas held in the decay tanks can either be returned to the Chemical and Volume 

Control System holdup tanks, or discharged to the atmosphere if it has 

decayed sufficiently for release. Generally, the last tank to receive gas 

will be the first tank emptied back to the holdup tanks in order to permit 

the maximum decay time before releasing to the environment.   

However, the header arrangement at the tank inlet gives the operator freedom 

to fill, re-use or discharge gas to the environment simultaneously without   

restricting operation of the other tanks. During degassing of the reactor   

coolant prior to a refueling shutdown, it may be desirable to pump the gas   

purged from the volume control tank into a particular tank and isolate that 

tank for decay rather than re-use the gas in it. This is done by aligning 

the control to open the inlet valve to the desired tank and closing the 

outlet valve to the re-use header.   

11.1-6 However, one of the other tanks can be opened to the re-use header at this time if desired, while still another might be discharged to atmosphere. 

Before a tank can be emptied to the environment, it must be sampled and analyzed to determine the activity to be released. Once the activity has been recorded the gas can be discharged to the plant vent at a controlled rate through a radiation monitor. Samples are taken manually by opening an isolation valve from the gas decay tank discharge to the gas analyzer and collecting the gas in one of the sampling system gas sample vessels. If sampling has shown that sufficient decay has occurred, the isolation valve in the line from the tank to the gas analyzer is closed, the isolation valve in the plant vent discharge line is opened and the tank contents are released through the plant vent. During release a trip valve in the discharge line is closed automatically by loss of air flow from auxiliary building exhaust fans. In the event of a high activity level in the discharge line, the plant vent isolation valve RCV-014 will either be closed automatically (PVGM R-14 in service) or manually (RAD 6304 in service). 

During operation, a gas sample is drawn from  the particular gas decay tank being filled at the time, and analyzed to determine its hydrogen and oxygen content. The hydrogen analysis is for surveillance since the concentration range can vary considerably from tank to tank. Also, the capability exists for manual grab sample analysis of cover gases from tanks discharging to the waste gas vent header. 

Solids Processing 

The Waste Disposal System is designed to package all solid wastes in High Integrity Containers (HICs) for removal to disposal facilities. The HICs are designed to be placed into transfer casks for shipment off-site for disposal.
The HICs are also designed to be stored in the Low Level Waste Storage Facility while awaiting shipment off-site for disposal. Refer to Figures 11.1-17, 11.1-18 and 11.1-19 for the spent resin processing flow diagrams. 

11.1-7 Revised 12/05/2014 C27 The spent resins from the CVCS demineralizers are normally deposited in the 

spent resin storage tank. After resin in the spent resin storage tank has 

been agitated by bubbling nitrogen through the tank to the vent header, water 

is pumped through the tank at a controlled rate to sluice the slurry to the 

container area. There it is received in shielded containers and dewatered 

for disposal. 

Provisions for dry bulk packaging of liquid waste system spent resins also 

exist. Spent resin is pumped as a water-resin slurry into a disposable 

container, which has connections for a dewatering line. The sluice water is 

removed by using a dewatering pump, which is piped to the waste hold-up tank 

through the floor drains. 

11.1-8 Rev. 15  4/98 All system components and piping can be internally decontaminated with flushing water from the primary water system. The permanently installed 

flushing water pipes can be isolated with manually operated valves.   

Control valves and pumps handling radioactive fluids are functionally   

grouped together and located behind shield walls. The equipment is installed 

to permit easy access for maintenance work, tests,  inspections, and 

replacement with minimum exposure to personnel.   

Shielding is provided for each container as necessary to reduce the work area 

dose rates. The basis for all dose rate calculations is for one cycle of core 

operation with one percent defective fuel in each unit.   

Components

Codes applying to components of the Waste Disposal System are shown in Table 

11.1-2. Components summary data are shown in Table 11.1-3.   

Laundry and Hot Shower Tanks

Three stainless steel tanks collect liquid wastes originating from the 

laundry. When a tank has been filled, its contents are pumped to one of the 

monitor tanks or waste monitor tanks after passing through a strainer and 

filter. If the radioactivity level is within permissible limits, the liquid 

is released to the circulating water system. 

Reactor Coolant Drain Tanks

The reactor coolant drain tanks are all-welded austenitic stainless steel.  

There is one tank inside the containment of each of the two units. This tank 

serves as a drain collecting point for the Reactor Coolant System and other 

equipment located inside the containment. 

11.1-9 Rev. 16  10/99 Waste Holdup Tanks The two waste holdup tanks can receive radioactive liquids from the Chemical 

and Volume Control System, floor drains, chemical drains, reactor coolant 

drain tanks, and laundry and hot shower tanks. The tanks are of stainless 

steel welded construction. The 24,300 gallons and 10,000 gallons waste 

hold-up tanks are located in the auxiliary building and radwaste facility, respectively. Contents of the auxiliary building tank can be transferred to 

the radwaste facility tank, but not vice-versa.   

Spent Resin Storage Tank

The spent resin storage tank retains spent resin discharged from some of the 

demineralizers. Normally, the tank is filled over a long period of time, the 

contents are allowed to decay. A layer of water is maintained over the resin 

surface to prevent resin degradation due to heat generation from decaying 

fission products. The tank is all welded austenitic stainless steel.   

Gas Decay Tanks

Six welded carbon steel tanks are provided to contain compressed waste gases 

(hydrogen, nitrogen, and fission gases). After a period for radioactive 

decay, these gases may be released at a controlled rate to the atmosphere 

through the plant vent. All discharges to the atmosphere will be monitored. 

Compressors

Two compressors are provided for removal of gases to the gas decay tanks from 

all equipment that contains or can contain radioactive gases. These 

compressors are of the water-sealed centrifugal displacement type. The 

operation of the compressors is automatically controlled by the gas manifold 

pressure. Construction is primarily carbon steel. A mechanical seal is 

provided to minimize leakage of seal water. While one unit is in operation, the other serves as a standby for unusually high flows or failure of the 

first unit.   

11.1-10   Rev. 16  10/99  

Waste Monitor Tanks

The contents of one of the three waste monitor tanks are analyzed for levels 

of radioactivity. If the activity is sufficiently low, the contents of the 

tanks are released to the circulating water system by one of two waste 

monitor pumps. Otherwise, the contents are returned to the waste holdup 

tanks for reprocessing. These tanks, are fabricated from stainless steel and 

meet the requirements of ASME Section VIII. Each tank provides the 

capability of storing 5,000 gallons of water. 

11.1-11 Rev. 16  10/99 Monitor Tanks See description in Section 9.2.2.   

Waste Disposal Demineralizer

Waste water in the waste holdup tank is processed primarily by the waste   

disposal demineralizer to reduce the level of activity. The liquid passes   

through a portable demineralization system which provides filtration and ion 

exchange before it is conveyed to the waste monitor tanks or monitor tanks.  

11.1-12 Rev. 16 10/99 

Nitrogen Manifold

A dual manifold supplies nitrogen to purge the vapor space of various 

components to reduce the hydrogen. concentration or to replace fluid that has 

been removed. A large volume Dewar vessel which is maintained above a preset 

level, assures a continuous supply of gas. Additionally, bottled gas is 

provided for short-term maintenance and backup requirements. 

Hydrogen Manifold

A dual manifold supplies hydrogen to the volume control tank to maintain the 

hydrogen partial pressure as hydrogen dissolves in the reactor coolant. A  

pressure controller, which is manually switched from one manifold to the 

other, assures a continuous supply of gas. 

11.1-13 Rev. 16  10/99 Gas Analyzer Manual sampling and laboratory analysis is conducted to monitor the 

concentrations of oxygen and hydrogen in the cover gas of various Waste 

Disposal System tanks, Chemical and Volume Control System tanks and the 

pressurizer relief tank. Upon indication of a high oxygen level, provisions 

are made to purge the equipment to the gaseous waste system with an inert 

gas. 

Continuous sampling of the gas decay tank being filled is performed by on-

line equipment. A local alarm warns of a potentially explosive condition.   

Pumps Pumps used throughout the system for draining tanks and transferring liquids 

shown in Figures 11.1-1a and 11.1-1b are either canned motor or mechanically 

sealed types to minimize leakage. The wetted surfaces of all pumps are   

stainless steel or other materials of equivalent corrosion resistance.   

Piping In general, the permanent piping which carries liquid wastes is stainless 

steel. All gas piping is carbon steel. Piping connections are welded except 

where flanged connections are necessary to facilitate equipment maintenance. 

Valves All valves exposed to gases are carbon steel. All other valves are stainless 

steel. All valves have stem leakage control. Globe valves are installed 

with flow over the seats when such an arrangement reduces the possibility of 

leakage. Stop valves are provided to isolate equipment for maintenance, to 

direct the flow of waste through the system, and to isolate storage tanks for 

radioactive decay.   

Relief valves are provided for tanks containing radioactive wastes if the 

tanks might be overpressurized by improper operation or component 

malfunction. Tanks containing wastes which are normally of low radioactivity 

level are vented locally.   

11.1-14 6/4/2001 11.1.3  DESIGN EVALUATION The following section was prepared as part of the licensing process for the plant. This section is historical and has not been updated in consideration of revisions to 10 CFR Part 20. Reference to 10CFR Part 20 refer to the pre-1990 version of 10 CFR.
Liquid Releases

Based on the estimated total liquid discharge to the Waste Disposal System in Table 11.1-4 and the capacity of the waste monitor and monitor tanks, the   

estimated number of yearly releases is 1000. This evaluation was performed 

for original plant licensing and is conservative with respect to actual 

operations.   

The estimated annual liquid release is indicated in Table 11.1-5. The maximum activity discharge rate will be controlled to assure that the 

circulating water concentration during releases is as low as practicable 

below the requirements of 10CFR20.   

The liquid waste processing facilities have been evaluated and demonstrated 

to be in compliance with 10CFR50, Appendix I requirements. This is addressed 

in supplementary licensing documents*.   

Liquid Wastes (Without Primary - Secondary Leakage)

Liquid wastes are generated primarily by plant maintenance and service operations, and consequently, the quantities and activity concentrations of   

influents to the system, Tables 11.1-4 and 11.1-5, are estimated values.   

Therefore, considerable operational margin has been assigned between the   

estimated system load and the design capability as indicated by Table 11.1-4. 

A conservative estimate of activity released in the liquid phase is 

summarized in Table 11.1-5. This tabulation is generated as follows:   
: 1.        All liquid waste is initially at peak reactor coolant activity   

concentrations based on continuous full power operation with   

1% defective fuel clad in each unit.   
: 2.        Allow 500 minutes for decay, the time required to process a   

1000 gallon batch at 2 gallons per minute.**   
: 3.        Concentrate the waste to a bottoms activity concentration of   

40 uc/cc, the packaging facility design limit.**   
  **These values are based on original system design and operating   

characteristics. While changes have been made to the original system,     

actual releases continue to meet the guidelines of 10CFR20.   

11.1-15 Rev. 16  10/99 

*Letter L-76-212, "Appendix I Evaluation" dated June 4, 1976 from R.E. 

Uhrig   of Florida Power and Light to D. R. Muller of the USNRC.  
: 4. Divide demineralizer combined DF of at least 10 6 which yields 4 x 10
-5 uc/cc in the waste condensate. 
: 5. Multiply by the quantity released from both units, listed in Table 11.1-4, to obtain the total estimated annual release in Table 11.1-5.   
: 6. Add to this the activity released through waste disposal by the CVCS monitor tanks. This is estimated to be less than 2 mc/yr. 
: 7. The tritium estimate in Table 11.1-5 assumes that one percent of the tritium that is formed in the fuel (the predominant source) diffuses    

through the zircaloy clad and enters the reactor coolant. Tritium      

discharges will be evaluated and accounted for by analyzing a composite 

sample. All of the sources of tritium accumulating in the            

reactor coolant, shown in Table 9.2-6, are included in the annual 

release.  
: 8. When the liquid in the waste monitor or monitor tanks has been properly determined to have an activity level low enough for discharge according 

to the release requirements of 10CFR20 for unidentified isotopes, the 

monitor tank pumps or waste monitor pumps are started and the liquid can 

be discharged to the seal wells of either Unit 3 or Unit 4 or both. The 

valves at the seal wells are electrically interlocked with the 

circulating water pumps to prevent liquid from being discharged into an 

inactive well, thus ensuring complete mixing at all times. Discharge 

piping is shown schematically in Figure 11.1-3. 
: 9. A radiation monitor (described in Section 11.2.3) automatically closes the discharge from the waste monitor tank pumps or monitor tank pumps if 

the activity level exceeds the monitor set point. This ensures that the 

activity in the circulating water discharge canal will be below the 

release requirements of 10CFR20. 

11.1-16 Rev. 16  10/99 Liquid Waste With Primary - Secondary Leakage The isotopic equilibrium activity concentration in the secondary coolant for 

any given radioisotope is related to the reactor coolant activity, the steam 

generator blowdown (cleanup) flow, the isotopic natural decay and the primary 

to secondary leakage flow by the following equation:
where: C si = Secondary coolant activity, c/cc C pi = Primary coolant activity, c/cc L ps = Primary to secondary leakage, gpm F s  = Secondary blowdown flow, gpm i  = Isotopic natural decay constant, min. 
-1 , and V s  = Liquid volume of the secondary coolant, gal. 

The relationship assumes that the reactor coolant equilibrium is independent 

of leakage rate. Consideration is given to I-131 as the major contributor to 

environmental activity release, because noble gas concentrations in the 

secondary will be quite low, being continuously entrained with the normal 

steam flow and released to the atmosphere through the air ejector. 

The above relation is plotted in Figure 11.1-5 for I-131 as the primary to 

secondary leak rate versus the ratio of secondary to primary activity as a 

function of various blowdown flows. 

The steam generators blowdown system, shown in Figures 10.2-41 and 10.2-42, consists of three independent blowdown lines (one per steam generator) which 

tie into a common blowdown flash tank. High activity liquid contained in the 

flash tank can be directed to the radioactive liquid waste system through 

manual valve alignment and a portable hose connection. The flashing 

component is discharged to the atmosphere or the shell side of the associated 

number 4 feedwater heaters. The blowdown tank liquid overflow discharge goes 

to the circulating water discharge canal. Upon indication of a high-

radiation level, a radiation monitor provided in the header of the steam 

generators' blowdown sampling lines on each unit will actuate solenoid valves 

on the associated unit to automatically isolate the blowdown including the 

sampling lines, close the valve in the blowdown flash tank discharge line to 

the circulating water discharge canal and sound an alarm in the control room. 

Because the iodine preferentially remains in the liquid phase, the air 

ejector monitor would be less sensitive than the liquid effluent monitor to 

iodine activity. 

11.1-17 Rev. 16  10/99 F s+V sic/cc C p i L ps=C si The steam generators blowdown flow is normally maintained at 1% or less of feedwater flow. This flow is adjusted as required to control the chemistry 

in the steam generators secondary side within established requirements.  

During startup or abnormal conditions the steam generator blowdown flow may 

be increased, as required, to approximately 6% of the feedwater flow. 

The setpoints of the steam generator blowdown radiation monitors are selected 

to isolate the blowdown, as previously indicated, at an activity 

concentration that will limit the combined secondary coolant and radwaste 

releases to below 10CFR20 requirements. The alarm setpoints for these 

monitors are determined by and set in accordance with the methodology  and 

parameters of the Turkey Point Offsite Dose Calculation Manual (ODCM). ODCM 

implementation is required by Technical Specification 6.8.
Blowdown was analyzed for radiological considerations in the original FSAR to 

occur routinely, typically on a daily basis over a one to several hour period 

at which times a flowrate of approximately 50 gpm is maintained. Assuming a 

permissible limit for the 624,000 gpm condenser cooling water iodine 

concentration at ten times the concentration limit for a one-hour blowdown, then, for several values of percent failed fuel, the allowable maximum 

primary to secondary leak rates can be read from Figure 11.1-5. At these 

limiting values, the combined secondary coolant and radwaste releases would 

be below 10CFR20 requirements provided blowdown did not exceed 2.5 hours per 

day. In addition to the ranges of normal operating conditions with tolerable 

amounts of failed fuel and primary to secondary leak rates, the site boundary 

I-131 equivalent dose is estimated under the following assumptions:   
: a. Steam line break outside the containment under no load conditions,   
: b. Releasing the contents of one steam generator,   
: c. Secondary I-131 activity = Primary I-131 activity =
1.5 uCi/cc from 1% failed fuel, and   
: d. One tenth the iodine content in the steam generator reaches the site boundary.
The site boundary thyroid dose equals approximately 1.78 rem.
Gaseous Wastes Gaseous wastes consist primarily of hydrogen stripped from coolant discharged to the CVCS holdup tanks during boron dilution, nitrogen and hydrogen gases 

purged for the CVCS volume control when degassing the reactor coolant, and 

nitrogen from the closed gas blanketing system. The gas decay tank capacity 

will permit 45 days decay of waste gas before discharge. Table 11.1-6 

contains an estimate of annual noble gas activity release based on the 

following assumptions:   

11.1-18                 Rev. 16  10/99 For Xe-133:   
: 1. The quantity of Xe-133 removed from the plant over a core cycle is determined assuming all gaseous waste is initially at peak reactor  

coolant activity concentration based on 1% defective fuel clad, and 

each unit at 2300 Mwt power with daily load reduction to 15% power. 
: 2. Using the same reactor coolant activity concentrations as in (1), the total Xe-133 removed to the Waste Disposal System by degassing the 

Reactor Coolant System for three cold shutdowns are combined. The 

Cold shutdowns are assumed to occur at the following times: (a) during 

the second week of operation, (b) at the peak xenon level and (c) 

during refueling. 
: 3. Using the same reactor coolant activity concentrations as in (1) the total Xe-133 removed from the reactor coolant to the Waste Disposal 

System as a result of 4 hot shutdowns occurring at equal intervals in 

the core cycle. 
: 4. Sum items 1, 2 and 3 for two units to obtain the total Xe-133 removed to the Waste Disposal System and allow for 45 days decay to obtain the 

total estimated annual release of Xe-133. 

For Kr-85:   

Since there is not significant decay of Kr-85 during the operating periods   

involved, the total Kr-85 that enters the reactor coolant during the core 

cycle is assumed to be eventually released through the Waste Disposal System. 

In comparison to Kr-85 and Xe-133, there will be no significant activity 

release after 45 days of decay from the remaining gaseous wastes since the  

isotopes half lives are short and/or the quantities present in the reactor 

coolant are small. 

11.1-19 Rev. 16  10/99 Gaseous Release Rate

In order to illustrate the conservatism that is available for gaseous 

releases from Turkey Point, an estimate has been made of the maximum release 

rate that would conform to 10 CFR 20. Considering Xe-133 and Kr-85 as the 

only nuclides, Table 2, Column 1, in Appendix B of 10 CFR 20 gives effluent 

concentration values of 5 x 10
-7 &#xb5;Ci/ml and 7 x 10
-7 &#xb5;Ci/ml Respectively, applicable at the site boundary.   

The average annual dilution factors for all 10 degree sectors of the site   

boundary are given in Figure 2D-1 and Table 2D-1, both in Appendix 2D of 

Section 2. For the three years of wind data taken at the site the largest 

dilution factor (X/Q) occurs in the 360 degree sector. The average value for 

the three year period, 1968-1970, is 1.02 x 10
-6 sec/m 3. For purposes of calculating the allowable gaseous routine release rate limit, this value is   

used.   

Using the above given effluent concentration value and X/Q value, the 

allowable average annual routine gaseous release limit is 0.49 Ci/sec, for 

Units 3 and 4 combined. In Table 11.1-6 the estimated release of Xe-133 and 

Kr-85 is 14,758 Ci/yr for Units 3 and 4 combined, and is equivalent to an 

average annual release rate of 0.47 x 10
-3) Ci/sec (which is much less than the 10 CFR 20 limit) using the   

conservative assumptions above.   

The estimated annual releases are as follows:   

No.        Ci/release       Release time, hrs.   

Min. 6            2460                   7   

Max. 20             760                   2.1   

The maximum release rate would be 97m Ci/sec. The site boundary effluent 

concentration will not be exceeded. Hold up for further natural decay of 

xenon for an additional month as feasible, and proportionally fewer releases 

per year, would about halve the total activity released.   

11.1-20 Rev. 16  10/99 The iodine activity release to the atmosphere from the secondary system and from the waste processing system under the limiting operating conditions of 1% failed fuel and the steam generator tube leakage (0.135 gpm) described in this section, and an expected 0.8 plant availability factor for Turkey Point Units 3 and 4 is estimated to be 259 millicuries per year. The corresponding maximum thyroid dose at the site boundary would be 0.38 millirem. The computations are based upon the data in Table 9.2-4, (un-updated FSAR), use a stripping and plateout fraction for iodine of 4 x 10
-3, include a 45 day gas decay tank holdup and yield an annual release from the steam system of 252 millicuries and from the waste processing system of 7 millicuries. Under normal expected operating conditions these activity releases will be less than one-one hundredth of those indicated.   

The exposure of minors within the restricted area; if continuously present at the Scout Camp would be considerably below the limits established by 10 CFR 20.104, and 10 CFR 20.202 (pre-1990 10 CFR 20).   

The maximum probable exposure for this on site facility would be for an individual in the Scout Camp area during a release when the wind is blowing into this sector. That exposure would be 0.007 rem, assuming:
X/Q = 1.9 x 10
-4 sec/m 3      _    E = 0.205 (based on 48% contribution from Xe-133 and 52%     from Kr-85 after 45 day gas storage)      S = 0.105 Ci/sec.      The provisions for monitoring iodine release paths are as follows:      
: 1. Both the plant vent and Unit 3 spent fuel building exhaust vent have fixed filter iodine monitors.      
: 2. The iodine release via the blowdown tanks will be calculated from the integrated flow through the blowdown flow meters and the quantity of iodine measured in the secondary side of the steam generators. 

11.1-21 Revised 12/05/2014 
: 3. The iodine release from the hogging jets, main steam safety valves, and waterbox priming jets will be calculated from steam flow and the iodine measured weekly in the main steam samples.
Steam flow will be calculated from time in use times maximum flow capacity of the device. 
: 4. The iodine release via the steam jet air ejectors on the main condenser will be calculated from a weekly sampling for iodine at the air ejector discharge and a concurrently made air flow measurement or from time in use multiplied by maximum flow capacity of the device. As described in Section 3.0 of the Offsite Dose Calculation Manual (ODCM), iodine sampling of the condenser air ejectors is permitted to be performed using developed compensation factors which estimate the iodine release concentrations from the air ejectors as a function of the noble gas concentrations emitted.
The testing and/or measurements outlined in 2, 3, and 4 above shall only be made if iodine is detected in the secondary coolant by sampling required by the ODCM. 

In addition, there are process radiation monitors for the plant vent, condenser air ejectors, and steam generator blowdown as described in Section 11.2 and listed in Table 11.2-7. The alarm set points are set low to alert the operator before a significant release could occur.   

The gaseous waste processing facilities have been evaluated and the as-built arrangement and potential radioactive releases to the environment are demonstrated to be in compliance with 10CFR50, Appendix I requirements. This is addressed in supplementary licensing documents.* 

Solid Wastes Solid wastes can consist of spent resins, spent filters and miscellaneous materials. All solid wastes are packaged in containers for removal to a disposal facility. Low level waste packaged for shipment may be stored on-site in the Low Level Waste Storage Facility while awaiting shipment off-site to a disposal facility. 

11.1-22 Revised 12/05/2014 *Letter L-76-212, "Appendix I Evaluation", dated June 4, 1976 from R.E. Urhig of Florida Power and Light to D.R. Muller of the USNRC.
C27 TABLE 11.1-1   WASTE DISPOSAL SYSTEM PERFORMANCE DATA  (Two Units)
Plant Design Life  60 years Normal process capacity, liquids  Table 11.1-3 Evaporator load factor  Table 11.1-4 Annual liquid discharge 

Volume  Table 11.1-4 Activity Tritium  Table 11.1-5   Other  Table 11.1-5 Annual gaseous discharge 

Activity  Table 11.1-6 

06/26/2002 TABLE 11.1-2   WASTE DISPOSAL COMPONENTS CODE REQUIREMENTS 

Component  Code Reactor Coolant Drain Tanks ASME III, (1) Class C Spent Resin Storage Tank ASME III, (1) Class C Gas Decay Tanks ASME III, (1) Class C Waste Holdup Tank, Auxiliary Building No Code Waste Holdup Tank, Radwaste Building ASME III, (1) Class 3 Laundry and Hot Shower Tanks No Code Piping and Valves USAS-B31.1 (2) Section I Waste Gas Compressor No Code Waste Monitor Tanks ASME VIII Monitor Tanks See Table 9.2-3 Molybdate Holding Tank No Code Low Level Waste Storage Facility EPRI Guidelines & 2010 Florida Building Codes 

NOTES:  1. ASME III-American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Nuclear Vessels 
: 2. USAS-B31.1-Code for pressure piping and special nuclear cases where applicable 

Revised 12/05/2014 C27 TABLE 11.1-3  SHEET 1 of 2 COMPONENT 

==SUMMARY==
DATA TANKS                  Quantity   Type   Volume     Design     Design      Material Each     Pressure  Temperature 

Tank o F Reactor Coolant Drain 1 per Horiz 350 gal. 25 psig 267 ss 

Unit 

Laundry & Hot Shower 3 (3) Vert 600 gal. Atm 180 ss 

Waste Holdup, Aux.Bldg. 1 (3) Horiz 3242 ft 3 Atm 150 ss 

Waste Holdup, Rad. Fac. 1 (3)  10000 gal Atm 200 ss 

Spent Resin Storage 1 (3) Vert 300 ft 3 100 psig 150 ss 

Gas Decay 6 (3) Vert 525 ft 3 150 psig 150 cs 

Waste Monitor Tank 3 (3) Vert 5000 gal Atm 200 ss 

Molybdate Holding Tank  1 (3) Horiz 3000 gal Atm 150 cs  

Monitor Tank   See TABLE 9.2-3 

PUMPS               Quantity   Type      Flow     Head   Design     Design   Material (1)                                       each unit    ft. Pressure Temperature 

gpm             psig o F  

Reactor Coolant 2 per Horiz. 150 175 100 267 ss Drain Unit 3 unit cent. 

Reactor Coolant 2 per Horiz. 75/125 150/120 100 267 ss Drain Unit 4 unit cent. 

NOTES: 
: 1. Material contacting fluid. 
: 3. Shared by Units 3 and 4. 

Revised 5/10/04 TABLE 11.1-3  SHEET 2 of 2 COMPONENT 

==SUMMARY==
DATA Pumps               Quantity  Type      Flow      Head   Design   Design   Material (1)                                       Each Unit    Ft. Press. Temp. 

gpm             psig     &#xba;F 

Laundry 2 Horiz 100 250 150 180 ss cent(2) 

Waste Evaporator Feed 1* Horiz 20 100 100 150 ss  (Aux. Building)  cent(2) 

Auxiliary Building  Vert.
Sump 14 Duplex 75 70 45 220 cs 

Containment  Vert.
Sump 2 Duplex 75 70 45 220 cs 

Radwaste Facility Sump 2 Vert. 35 70   ss 

Waste Evaporator Feed 2* Horiz  35/100    250/200 150 200 ss cent 

Waste Monitor Tank 2* Horiz  35/100    250/200 150 200 ss cent 

Miscellaneous

Waste Gas Compressors 2*    Horiz (2) 22(CFM) - - - -
cent 

(2) Mechanical Seal Provided
* Shared by Unit 3 and Unit 4 

Rev. 16  10/99 TABLE 11.1-4   ESTIMATED LIQUID DISCHARGE TO WASTE DISPOSAL
* Weekly Discharges 

Source                       Peak, During   During Refuel-   Total Annual 

power, gal. ing, gal.        Discharge, gal. 

Two units      One unit at      Two units 

at power       power 

One unit 

refueling 

Laundry, shower,            12,200        112,850           1,593,590 

handwashes 

Laboratories                  600             600              31,200 

Equipment Drains, leaks      3040            2490             154,780 

Decontamination              1000             700              50,200 

Totals                      16,840        116,640           1,829,770 

Evaporator load Factor, %     <6             <39                 <12
* This table was developed as part of the original plant licensing process and is not updated. 

Rev. 16 10/99 TABLE 11.1-5 ESTIMATED LIQUID RELEASE BY ISOTOPE*  (TWO UNITS)  

Annual Yearly  Annual Yearly Release Average  Release Average 

Isotope uc      uc/cc   Isotope uc      uc/cc H 3** 2.90 x 10 9 1.28 x 10
-6 I 131 1.57 x 10 4 0.691 x 10-11 Mn 542 .16 x 10 0 0.95 x 10
-15 Te 132 1.66 x 10 3 0.731 x 10
-12 Mn 56 5.88 x 10 1 2.59 x 10
-14 I 132 4.86 x 10 2 2.14 x 10
-13 Co 58 6.58 x 10 1 2.9 x 10-14 I 133 1.99 x 10 4 0.876 x 10
-11 Co 60 7.76 x 10 0 3.42 x 10
-15 I 134 7.52 x 10
-2 3.31 x 10
-17 Sr 89 2.67 x 10 1 1.18 x 10
-14 I 135 5.80 x 10
-3 2.55 x 10
-12 Sr 90 8.04 x 10
-1 3.54 x 10
-16 Cs 134 1.73 x 10 3 0.762 x 10
-12 Y 90 9.24 x 10
-1 4.07 x 10
-16 Cs 136 2.50 x 10 2 1.10 x 10
-13 Sr 91 6.86 x 10 0 3.02 x 10
-15 Cs 137 9.40 x 10 3 4.14 x 10
-12 Y 91 4.72 x 10 1 2.08 x 10
-14 Ba 140 6.34 x 10 0 2.79 x 10
-15 Y 92 1.08 x 10 0 0.476 x 10
-15 La 140 5.82 x 10 0 2.56 x 10
-15 Mo 99 1.96 x 10 4 0.863 x 10
-11 Ce 144 2.26 x 10 1 1.00 x 10
-14  

Totals Tritium   2.90 x 10 9 uc/yr 1.28 x 10
-6 uc/cc Other Waste Disposal 7.5 x 10 4 uc/yr         3.30 x 10
-11 uc/cc Chemical and Volume Control System 2.0 x 10 3 uc/yr         0.881 x 10
-12 uc/cc
* These values are based on original system design and operating 

characteristics. While changes have been made to the original 

system, actual releases continue to meet the requirements of 

10CFR20. 

**    Initial cycle. 

TABLE 11.1-6   

ESTIMATED ANNUAL GASEOUS RELEASE BY ISOTOPE 

 (TWO UNITS)  

Activity 

Environment 

Isotope                                              Curies/yr

H 3                                                  Negligible 

Kr 85                                                7714 

Kr 85m, 87, 88                                       Negligible 

Xe 133                                               7044 

Xe 133m, 135, 135m, 138                              Negligible 

Total            14,758 

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-1 REFER TO ENGINEERING DRAWING 5613-M-3061 , SHEET 1

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 LIQUID WASTE DISPOSAL SYSTEM REACTOR COOLANT DRAIN TANK AND PUMPS FIGURE 11.1-1

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-2 

REFER TO ENGINEERING DRAWING 5613-M-3061 , SHEET 2

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 LIQUID WASTE DISPOSAL SYSTEM CONTAINMENT DRAINS FIGURE 11.1-2

FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 3 & 4 DISPOSAL OF RADIOACTIVE LIQUIDS FIGURE 11.1-3

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-4 

REFER TO ENGINEERING DRAWING 5614-M-3061 , SHEET 1

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 LIQUID WASTE DISPOSAL SYSTEM REACTOR COOLANT DRAIN TANK AND PUMPS FIGURE 11.1-4

FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT PRIMARY - SECONDARY ACTIVITY RELATIONS STUDY FIGURE 11.1-5

 [Figure 11.1 DELETED] 

Rev. 12  5/95     

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-7 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEETS  9 & 11

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 RADWASTE SOLIDIFICATION SYSTEM CEMENT HANDLING AND CONTAINER FILLING FIGURE 11.1-7

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-8 

REFER TO ENGINEERING DRAWING 5614-M-3061 , SHEET 2

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNIT 4 LIQUID WASTE DISPOSAL SYSTEM CONTAINMENT DRAINS FIGURE 11.1-8

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-9 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 1

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM WASTE HOLDUP & TRANSFER FIGURE 11.1-9

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-10 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 2

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM LAUNDRY WASTE FIGURE 11.1-10

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-11 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 3

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM DRAIN HEADERS AND SUMPS FIGURE 11.1-11

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-12 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 4

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM POLISHING DEMINERALIZER FIGURE 11.1-12

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-13 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 5

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM WASTE EVAPORATOR FEED FIGURE 11.1-13

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-14 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 6

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM WASTE EVAPORATOR PACKAGE FIGURE 11.1-14

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-15 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 7

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM LIQUID SAMPLING, MONITORING,  AND CHEMICAL ADDITION FIGURE 11.1-15

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-16 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 8

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LIQUID WASTE DISPOSAL SYSTEM WASTE MONITOR TANKS FIGURE 11.1-16

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-17 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 9

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SOLID WASTE DISPOSAL SYSTEM SPENT RESIN STORAGE FIGURE 11.1-17

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-18 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 10

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SOLID WASTE DISPOSAL SYSTEM HOLDUP AND MIXING FIGURE 11.1-18

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-19 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 11

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SOLID WASTE DISPOSAL SYSTEM CONTAINER FILL FIGURE 11.1-19

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-20 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 12

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GASEOUS WASTE DISPOSAL SYSTEM WASTE GAS COMPRESSORS FIGURE 11.1-20

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-21 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 13

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GASEOUS WASTE DISPOSAL SYSTEM WASTE GAS DECAY TANKS FIGURE 11.1-21

FINAL SAFETY ANALYSIS REPORT FIGURE 11.1-22 

REFER TO ENGINEERING DRAWING 5610-M-3061 , SHEET 14

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GASEOUS WASTE DISPOSAL SYSTEM GAS WASTE ANALYZERS FIGURE 11.1-22

11.2   RADIATION PROTECTION 11.2.1  DESIGN BASES   

Monitoring Radioactivity Releases

Criterion: Means shall be provided for monitoring the containment atmosphere and the facility effluent discharge paths for radioactivity 

released from normal operations, from anticipated transients, and 

from accident conditions. An environmental monitoring program 

shall be maintained to confirm that radioactivity released to the 

environs of the plant have not been excessive.  (1967 Proposed 

GDC 17)
The containment atmosphere, the plant vent, Unit 3 spent fuel pit exhaust, the condenser air ejector exhaust, the steam generator blowdown effluent, the 

main steam lines and the Waste Disposal System liquid effluent are monitored 

for radioactivity concentration during normal operations, anticipated 

transients, and postulated accident conditions. High radiation activity from 

any of these sources is indicated, recorded and alarmed in the control room. 

Waste disposal system liquid effluent released to the circulating water 

system canal is monitored. For the case of leakage from the containment 

under MHA   

conditions, the area radiation monitoring system, supplemented by portable   

survey equipment provides adequate monitoring of releases. An outline of the 

procedures and equipment to be used in the event of a postulated accident are 

discussed in Section 11.2.2 and 12.3. The environmental monitoring program 

is described in Section 2.   

Monitoring Fuel and Waste Storage

Criterion: Monitoring and alarm instrumentation shall be provided for fuel and waste storage and associated handling areas for conditions 

that might result in loss of capability to remove decay heat and 

to detect excessive radiation levels (1967 Proposed GDC 18).   

11.2-1 Revised 04/17/2013 Monitoring and alarm instrumentation are provided for fuel and waste storage and handling areas to detect inadequate cooling and excessive radiation 

levels. Radiation monitors are provided to maintain surveillance over the 

release of gases and liquids.   

The spent fuel pit cooling loop flow is monitored to assure proper operation, as shown in Section 9.3.   

Ventilation systems exhaust air from the auxiliary building and radwaste 

facility and discharge to the atmosphere via plant vent through roughing and 

HEPA filters. Exhaust air from Units 3 and 4 Containments discharge to the 

atmosphere via the plant vent through roughing filters. Radiation monitors 

are in continuous service in this area and actuate a high-activity alarm on 

the control board annunciator as described in Section 11.2.3.   

Fuel and Waste Storage Radiation Shielding

Criterion: Adequate shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities. (1967 Proposed GDC 68)  

Auxiliary shielding for the Waste Disposal System and its storage components are designed to limit the dose rate to levels not exceeding 0.5 mr/hr in 

normally occupied areas, to levels not exceeding 2.5 mr/hr in periodically 

occupied areas and to levels not exceeding 15 mr/hr in limited occupancy 

areas. Actual doses in these areas varies with plant conditions and may 

exceed the design values. Dose to plant personnel is controlled 

administratively to maintain doses ALARA.  

Gamma radiation is continuously monitored at various locations in the 

Auxiliary Building and fuel storage areas. A high level is alarmed locally 

and annunciated in the control room.  

Protection Against Radioactivity Release from Spent Fuel and Waste Storage Criterion: Provisions shall be made in the design of fuel and waste storage facilities such that no undue risk to the health and safety of the public could result from an accidental release of 

radioactivity. (1967 Proposed GDC 69)   

11.2-2 Revised 04/17/2013 All waste handling and storage facilities are contained and equipment is designed so that accidental releases directly to the atmosphere will not 

exceed the 10 CFR 50.67 guidelines; refer also to Sections 11.1.2, 14.2.2 and 

14.2.3. The components of the Waste Disposal System are not subjected to any 

high pressures (see Table 11.1-3) or stresses. In addition, the tanks, which 

have a design pressure greater than atmospheric pressure, piping and valves 

of the system are designed to the codes given in Table 11.1.2. Hence, the 

probability of a rupture or failure of the system is exceedingly low.   

11.2.2  PRIMARY AND SECONDARY SHIELDING   

Design Basis

Radiation shielding is designed for operation at maximum calculated thermal 

power and to limit the normal operation levels at the site boundary to below 

those levels allowed for continuous non-occupational exposure.  

Original design of the plant shielding was performed assuming a core power level of 2296 MWt and a 12-month fuel cycle length. The plant shielding was re-evaluated for the power uprate assuming a core thermal power of 2652 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% fuel defects, it is concluded that the increase in the core power level and in the fuel cycle length will have no significant impact on plant shielding adequacy and safe plant operation. 

In addition, the shielding and containment measures provided ensure that in 

the event of the maximum hypothetical accident, the evaluated off site and 

control room operator dose results will remain below the applicable limits in 

10CFR 50.67 and Regulatory Guide 1.183. 

Operating personnel are protected by adequate shielding, monitoring, and 

operating procedures. Each area is classified according to the dose rate 

allowable in the area. The allowable dose rate is based on the expected 

frequency and duration of occupancy. All areas capable of personnel 

occupancy are classified as one of five zones of radiation level as shown in 

Fig. 11.2-1 and 11.2-2. The classification of occupancy of the zones is 

listed in Table  11.2-1. Typical Zone I areas are the offices, control room, the turbine area and turbine service areas. Zone II areas include the 

passageways and local  control spaces in the Auxiliary Building and the 

operating floor of the containment during reactor shutdown. Areas designated 

Zone III include the sample rooms, valve room, fuel handling areas, and 

intermittently occupied work areas. 

11.2-3  Revised 04/17/2013 C26C26C26C26 Typical Zone IV areas are the shielded equipment compartments in the Auxiliary Building and the reactor coolant loop compartments after shutdown. 

Zone V areas are high radiation controlled areas. These areas are those 

containing high radiation components such as gas decay tank, mixed bed 

demineralizers, spent resin tank, waste container storage area, and volume 

control tank.   

All radiation areas are appropriately marked and isolated in accordance with 

10 CFR 20 and other applicable regulations.   

The unit shielding is divided into five categories according to function.  

These functions include the primary shielding, the secondary shielding, the 

accident shielding, the fuel transfer shielding, and the auxiliary shielding.  

Primary Shield

The primary shield is designed to:   
: 1. Reduce the neutron flux incident on the reactor vessel to limit the radiation induced increase in transition temperature.   
: 2. Attenuate the neutron flux sufficiently to prevent excessive activation of components.   
: 3. Limit the gamma flux in the reactor vessel and the primary concrete shield to avoid excessive temperature gradients or dehydration of the 

primary concrete shield.   
: 4. Reduce the residual radiation from the core, reactor internals and reactor vessel to levels which will permit access to the region between 

the primary and secondary shields after shutdown.   
: 5. Reduce the induced secondary radiation leakage to obtain optimum division of the shielding between the primary and secondary shields.   

11.2-4  Revised 04/17/2013 Secondary Shield The main function of the secondary shielding is to attenuate the radiation 

originating in the reactor and the reactor coolant. The major source in the 

reactor coolant is the Nitrogen - 16 activity, which is produced by neutron 

activation of oxygen during passage of the coolant through the core. 

Accident Shield

The accident shield ensures safe radiation levels for desired component 

access outside the containment following a maximum hypothetical accident. 

Fuel Handling Shield

The fuel handling shield permits the safe removal and transfer of spent fuel 

assemblies and Rod Control Cluster Assemblies (RCCAs) from the reactor vessel 

to the spent fuel pit. It is designed to attenuate radiation from spent 

fuel, RCCAs and reactor vessel internals to less than 2.5 mr/hr at the 

refueling cavity water surface and to less than 15 mr/hr within the spent 

fuel area. 

Auxiliary Shielding

The function of the shielding is to protect personnel working near various 

system components in the Chemical and Volume Control System, the Residual 

Heat Removal System, the Waste Disposal System and the Sampling System. The 

shielding provided for the auxiliary building is designed to limit the dose 

rate to less than 0.5 mr/hr in normally occupied areas, and below 2.5 mr/hr 

in periodically occupied areas. 

11.2-5 Revised 04/17/2013 Shielding Design Primary Shield The primary shield consists of the core baffle, water annuli, barrel-thermal 

shield, all of which are within the reactor vessel, the reactor vessel wall, and a concrete structure surrounding the reactor vessel. 

The primary shield immediately surrounding the reactor vessel consists of an 

annular reinforced concrete structure extending from the base of the 

containment to an elevation of 58'-0". The lower portion of the shield has a 

minimum thickness of 7.0 feet of regular concrete (P=2.3 g/cc) and is an integral part of the main structural concrete support of the reactor vessel; 

it extends upward to the refueling floor, with vertical walls 4 and 5 feet 

thick, to form an integral portion of the refueling cavity. 

The primary shield neutron flux are listed in Table 11.2-2. The flux listed 

are those occurring on the horizontal mid-plane of the core. 

At locations other than the horizontal midplane of the core the intensity of 

both the neutron and the gamma flux begins to decrease. 

Secondary Shield

The secondary shield surrounds the reactor coolant loops and the primary 

shield. It consists of interior walls in the containment, the operating floor 

at elevation 58'-0" and the floor at elevation 30'-6".   

Certain interior walls within the containment also serve as the accident 

shield. 

11.2-6 Revised 04/17/2013 The main portion of the secondary shield consists of 2' to 3'6" walls, which 

surround the coolant loops and steam generators at the 30'-6" and 58'-0" elevations. The secondary shield will attenuate the radiation levels in the 

reactor coolant loop compartment from a value of 25 rem/hr. to a level of 

less than 1 mrem/hr outside the containment. 

The original secondary shield design parameters are listed in Table 11.2-3.
With the 1995 thermal uprate and the 2012 Extended Power Uprate (EPU), core power has increased beyond the original design basis for the secondary shield. However, survey history over the plant's operational lifetime shows that the original design remains adequate to limit the dose rate outside the containment building to well within the 1 mrem/hr limit established above, when uprate scaling factors are applied. 

Accident Shield

The accident shield consists mainly of the containment structure. The 

containment structure is a reinforced post-tensioned concrete cylinder 3 ft. 

9 in. thick capped by a reinforced and post-tensioned concrete dome 3 ft. 3 

in. thick. 

Shielding has been provided within the containment in excess of that required 

for operational reasons to limit the post accident dose which otherwise might 

be present within the containment at penetration areas. 

Additional shielding has also been provided within the Auxiliary Building to 

permit post accident access to the RHR system area. 

The control room is shielded so that the post accident integrated dose from 

direct radiological shine to personnel in that room will be less than 2 Rem. 

The impacts of radiological shine on post-accident whole body gamma dose to 

control room personnel were accounted for in the analyses to address NUREG-

0578, Item 2.1.6.b and NUREG-0737, Items II.B.2.2 and III.D.3.4. Based on 

the results of these analyses, a 1-1/2 inch steel shadow shield was installed 

between the Unit 3 54-inch purge valve and north control room wall via PC/M 

80-63. The NRC commitments related to this modification are contained in 

References 4 and 5. The total control room operator shine dose consequence 

results for EPU conditions in Table 11.2-12 and Section 14.3.5.1 include the 

effects of this shadow shield. 

11.2-7 Revised 04/17/2013 C26C26 Fuel Handling Shield The refueling cavity is irregularly shaped, formed by the upper portions of 

the primary shield concrete, and other sidewalls of varying thicknesses. A 

portion of the cavity is used for storing the upper and lower internals 

packages. The walls vary in thickness, from 4 to 5 ft. 

The refueling cavity, flooded with borated water to elevation 56'-10" during 

refueling operations, provides a temporary water shield above the components 

being withdrawn from the reactor vessel. The water height during refueling 

is approximately 23 ft. above the reactor vessel flange. This height ensures 

that a minimum of 9 ft. of water will be above the active fuel of a withdrawn 

fuel assembly. Under these conditions, the dose rate from only the active 

fuel is less than 12 mrem/hr at the water surface. 

The spent fuel assemblies and RCC assemblies are remotely removed from the 

containment through the horizontal spent fuel transfer tube to be placed in 

the spent fuel pit. Concrete, 3' to 4'6" thick, shields the spent fuel 

transfer tube. This shielding is designed to protect personnel from 

radiation during a time a spent fuel assembly is passing through the main 

concrete support of the containment and the transfer tube. 

Radial shielding as the spent fuel is raised for transfer to the spent fuel 

storage pit is provided by the water and concrete walls of the fuel transfer 

canal. Actual dose rates in the area adjacent to the spent fuel storage pit 

may exceed design values during spent fuel transfer. Administrative 

procedures ensure that dose to personnel is maintained ALARA. 

Fuel is stored in the spent fuel pit of the Auxiliary Building which is 

located adjacent to the containment. Shielding for the spent fuel storage 

pit is provided by 5' 6" thick concrete walls to elevation 32' 10"; above 

this elevation the walls are tapered in places to a thickness of 3 ft. The 

pit is flooded to a level such that the water height is 23 feet above the 

stored spent assemblies. During spent fuel handling a minimum of 7 feet 11 

inches is maintained above the top of a fuel assembly. 

11.2-8 Revised 04/17/2013 Radiation from spent fuel has increased since the original shielding design due to the 1995 and 2012 power uprates and the transition to an 18 month fuel cycle. However, survey history over plant's lifetime demonstrates that the original shielding design for fuel handling continues to provide adequate protection for operators in these areas when uprate scaling factors are applied, within the limits established above.
Auxiliary Shielding

The auxiliary shield consists of concrete walls around certain components and 

piping which contain reactor coolant. In some cases, the concrete block 

walls are removable to allow personnel access to equipment during maintenance 

periods. Access to the Auxiliary Building is allowed during reactor 

operation. Equipment is shielded so that compartments may be entered without 

having to shut down,or to decontaminate equipment in an adjacent room. 

The shield material provided throughout the Auxiliary Building is regular 

concrete (p=2.3 g/cc). The principal auxiliary shielding provided is 

tabulated in Table 11.2-6. 

In addition, in some cases the installation of temporary or permanent lead 

shielding may be necessary to reduce area dose due to localized hot spots. 

11.2-9 Revised 04/17/2013 C26 11.2.3  RADIATION MONITORING SYSTEM   

The Radiation Monitoring System is designed to perform two basic functions: 
: a. Warn of any radiation health violation which might develop. 
: b. Give early warnings of a malfunction which might lead to an unsafe health condition or unit damage.   

Instruments are located at selected points in and around the unit to detect, and record the radiation levels. In the event the radiation level should 

rise above a desired setpoint, an alarm is initiated in the control room.  

The Radiation Monitoring System operates in conjunction with regular and 

special radiation surveys and with chemical and radiochemical analyses 

performed by the plant staff to provide adequate information and warning for 

the continued safe operation of the units and assurance that personnel 

exposure does not exceed 10 CFR 20 guidelines.   

The components of the Radiation Monitoring System are designed according to 

the following environmental conditions:   
: a. Temperature - 40 F to 125 F.   
: b. Humidity 95%   
: c. Pressure containment monitors will withstand containment leak test pressure   
: d. Radiation - up to 100 times the maximum scale reading without damage to instrument.   

The Radiation Monitoring System is divided into the following sub systems:   
: a. Process Radiation Monitoring System Monitors various fluid streams in operating systems.   
: b. Area Radiation Monitoring System Monitors radiation levels at various locations within the operating area of the two units.  
: c. Environmental Radiation Monitoring System Monitors radiation exposure in the area surrounding the units.  

11.2-10 Revised 04/17/2013 C26 Process Radiation Monitoring System This system consists of channels which monitor radiation levels in various 

operating systems. The output from most channel detectors is transmitted to 

the Radiation Monitoring System cabinets located in the control room area 

where the radiation level is indicated by a numerical display and recorded by 

a multipoint recorder. High radiation level alarms are annunciated in the 

control room and indicated on the Radiation Monitoring System cabinets. 

Each channel (except the R-*-15 Steam Jet Air Ejection Monitor Channels) contains a completely integrated modular assembly, which may include the following: 

a) Level Amplifier 

Amplifies the energy of the radiation pulse to provide a discriminated output to the log level amplifier. 

b) Log Level Amplifier 

Accepts the shaped pulse of the level amplifier output, performs a log integration, (converts total pulse rate to a logarithmic analog signal) 

and amplifies the resulting output for suitable indication and 

recording, c) Power Supplies 

Power supplies are contained in each drawer and/or monitoring skid for furnishing the positive and negative voltages for the transistor 

circuits, relays and alarm lights and for providing the high voltage 

for the detector. 

11.2-11 Revised 04/17/2013 C26C26 d) Test-Calibration Circuitry   

These circuits provide a precalibrated signal to perform channel test, and a solenoid operated radiation check source to verify the channel's 

operations. An annunciator light on the control board indicates when 

the channel is in the test calibrate mode. In lieu of a check source, RD-3-20 and RM-3-20 utilize an internally generated test function to 

ensure the rate meter is functioning properly. 

e) Radiation Level Numerical Display 

This display, mounted on the drawer, indicates in counts per minute on a digital display (R-14,R-*-17A/B, R-18, and R-*-19) or logarithmically 

in mR/hr on an analog display (R-4-20). The display is in Ci/cc from 10-11 to 10-5 for R-*-11 and 10
-6 to 10-1 for R-*-12. The level signal is also recorded. The display for RD-3-20 and RM-3-20 indicate in mR/hr 

on a digital display. 

f) Indicating Lights   

These lights indicate high-radiation alarm levels and circuit failure.
An annunciator on the control board is actuated on high radiation.  

R-3/4-20 also annunciate on the control board when a channel failure 

occurs.
g) Bistable Circuits   

Two bistable circuits are provided, one to alarm on high radiation  (actuation point may be set at any level over the range of the 

instruments), and one to alarm on loss of signal (circuit failure). 

h) A remotely operated long half-life radiation check source is furnished in each channel except R-3-20. The energy emission ranges are similar 

to the radiation energy spectra being monitored. The source strength 

is sufficient to cause a definite display increase above background. 

R-3-20 utilizes an internally generated test function which is used to compare to a baseline generated for the monitor. This comparison ensures that no degradation of the signal occurs without notification to the operator. The operator utilizes the test function in a manner similar to the check source and it performs a similar function. 

11.2-12 Revised 04/17/2013 C26C26C26C26C26C26C26 The R-3-15 and R-4-15 channels are modular assemblies with the following elements:
a) Radiation Detector A plastic Beta scintillation detector generates current pulses when exposed to radiation in the Steam Jet Air Ejector effluent.
b) Local Processing & Display Unit (LPDU)
The LPDU is located in the Load Center Room, Turbine Building Elev.
31'. It provides high voltage supply to the detector and processes the current pulses generated by the detector. The following elements are integral to the LPDU:  Preamplifier and amplifier circuits necessary to process the detector pulses  1024 multi-channel analyzer (MCA) necessary to determine counting in a specific range of energies  Software algorithms for using the detector counting to compute other measurements (e.g., volume activity, leakrate, etc.)  AC/DC power supply for the LPDU electronics and the detector  Display of process measurements as well as indication of alarms and faults via lights and buzzer.
c) Remote Display Unit (RDU)
The RDU is located in the Control Room. It provides a numerical and graphical display of the primary channel measurements computed by the LPDU, including count rate in CPM, volume activity in &#xb5;Ci/cc and/or leak rate in GPD. The following elements are integral to the RDU:
Status LEDs and associated relays to indicate the following operating conditions:
Operate/fault Test   Alert alarm   High alarm   High/High alarm   Programmable logic circuit which drives the automated check source sequence (see below)  Internal power supply for the electronics 

11.2-13 Revised 04/17/2013 C26 d) Data Logger The data logger records the RDU analog output during a check source test. The analog output is associated with detector counting in CPM.
The data logger is configured to convert the analog signal to count a rate value and store the measurements to non-volatile memory which can be downloaded and reviewed to track detector performance.
e) Check Source Circuit Provides an automated test of channel operability using a solenoid activated radioactive source. The source sequence is activated using the RDU keypad. During the sequence, radiation alarms are disabled, and the "Test" LED on the RDU is illuminated to indicate the channel is under testing. The radiation measurements during the test are archived by a data logger adjacent to the RDU so that the detector performance can be tracked over time.
f) Booster Relay Plate The booster relay plate contains five DPBT relays corresponding to the status relays of the RDU:
Operate  Test  Alert alarm  High alarm  High/High alarm The relays are used to drive annunciators corresponding to alarm states. The "operate" relay can be used to drive an annunciator if the channel experiences a fault. The relay plate contains an integral DC power supply which drives the coils of the booster relays and the data logger. g) Field Junction Box A field junction box houses a booster relay which drives the solenoid operated check source along with a key-switch for local or remote actuation of the check source. A terminal strip in the junction box permits user interface with the input/output signals of the LPDU:
Relay contacts  Analog input  Analog output  Serial links 

11.2-14 Revised 04/17/2013 C26 The Process Radiation Monitoring system consists of the following radiation monitoring channels:
Containment High Range Radiation Monitors (RaD-3-6311A & B, RaD-4-6311A & B)
Two high range radiation monitors and associated instrument channels are 

provided for in-containment post-accident monitoring in compliance with 

NUREG-0737, Item II.F.1. These monitors are shown on Figures 11.2-1 and 

11.2-2. 

Each channel monitors the containment radiation levels from 10 0 to 10 8 R/hr inside the containment. Each detector is a gamma ionization chamber 

installed inside the containment. The signal processor, which supplies 

indication and recording in the control room, the high range radiation module 

and recorder are located outside the containment. Two alarms are provided to 

alert the control room operator upon high radiation within the containment. 

Also a failure trip is provided to activate upon loss of power, high voltage, or signal from the detector. A sustaining signal is generated within the 

detector corresponding to 1 R/hr. A failure alarm will occur if the signal 

from the detector falls below this value. This feature assures knowledge of 

the monitor's integrity at all times. 

The safety-related redundant monitoring instrumentation channels are 

energized from independent Class 1E power sources, and are physically 

separated in accordance with Regulatory Guide 1.75. 

Containment Air Particulate Monitors (R3-11 & R4-11)

R3-11 and R4-11 are provided to measure air particulate beta radioactivity in 

each containment and to ensure that the release rate through each containment 

vent during purging is maintained below specified limits. Each monitor has a 

measuring range of at least 10
-9 to 10-6 &#xb5;Ci/cc. High radiation level for the channel initiates closure of the containment purge supply and exhaust duct 

valves and containment instrument air bleed valves, and initiates control 

room ventilation isolation. The alarm setpoints for these monitors are 

determined from Technical Specifications (Table 3.3-3) and set in accordance 

with the methodology and parameters of the Turkey Point ODCM. ODCM 

implementation is required by Technical Specification 6.8. 

11.2-15 Revised 04/17/2013 C26 The sample is drawn from the containment ductwork through a closed, sealed system monitored by a beta scintillation counter - filter paper detector assembly. The filter paper collects all particulate matter greater than 1 micron in size on its constantly moving surface, and is viewed by a photomultiplier-scintillation crystal combination. The samples are returned to the containment after it passes through a series connected (CH R3-12) or (CH R4-12) gas monitor. 

Each detector assembly is in a completely enclosed housing. The detector is a hermetically-sealed photomultiplier tube - scintillation crystal combination. Lead shielding is provided to reduce the background level to where it does not interfere with the detector's sensitivity. 

A backup containment air sampling system consisting of tubing, two isolation valves, a flow indicator, quick connects, conduit and a 120 volt receptacle provides easily accessible connections and a secure mounting location for a portable sampling pump to take "grab" samples. This system provides an alternate means to sample the containment atmosphere in the event that either RD-11 or RD-12 malfunctions. 

Containment air sampling to support personnel entry at power can be performed with R-11 and R-12, or via the backup containment air sampling system described above. 

The filter paper mechanism, an electro-mechanical assembly which controls the filter paper movement, is provided as an integral part of the detector unit. 

To reduce moisture in the Containment Air Particulate supply line and Containment Radioactive Gas Monitor return line, R-11 and R-12 lines are heat traced. Heat tracing on the R-11 and R-12 lines is not required to be operable and does not adversely affect the function of the Containment Air Particulate Monitors (R-3-11 and R-4-11) or the Containment Radioactive Gas Monitors (R-3-12 and R-4-12). 

11.2-16 Revised 01/22/2015 C28 Containment Radioactive Gas Monitors (R3-12) & (R4-12)

Each monitor is provided to measure gaseous beta radioactivity in the 

respective containment and, to ensure that the radiation release rate during 

purging is maintained below specified limits. High gas radiation level 

initiates closure of the containment purge supply and exhaust duct valves and 

containment instrument air bleed valves, and initiates control room 

ventilation isolation. 

Each monitor has a measuring range of at least 10
-6 to 10-3 &#xb5;Ci/cc. The alarm setpoints for these monitors are determined from Technical Specifications (Table 3.3-3) and set in accordance with the methodology and parameters of 

the Turkey Point ODCM. ODCM implementation is required by Technical 

Specification 6.8. 

The detector skid draws a continuous air sample from the containment 

atmosphere. After it passes through the air particulate monitor (R-*-11), it 

is drawn into the gaseous beta detector through a closed, sealed system. The 

sample is constantly mixed in the fixed, shielded volume, where it is viewed 

by the beta scintillation photomultiplier detector. The sample is then 

returned to the containment. 

The detector assembly is in a completely enclosed housing containing a beta 

scintillating detector mounted in a constant gas volume container. Lead 

shielding is provided to reduce the background level to a point where it does 

not interfere with the detector's sensitivity. A locally mounted electronic 

assembly transmits the signals to the remote indication, alarm, and control 

circuits. 

The containment air particulate and radioactive gas monitors have assemblies 

that are common to both channels. They are described as follows: 

a) The flow control assembly includes a pump unit and selector valves that 

provide a representative sample (or a "clean" sample) to the detector. 

b) The pump consists of:   
: 1. A pump to obtain the air sample.  
: 2. A flowmeter to indicate the flow rate. 
: 3. A flowmeter to indicate the flow adjustment. 
: 4. A flow alarm assembly to provide low and high flow alarm signals. 

11.2-17 Revised 04/17/2013 C26 c) Selector valves are used to direct the sample to the detector for monitoring and to block normal flow when the channel is in maintenance 

or "purging" condition. 

d) A temperature sensor and pressure sensor are used to protect the system from high sample stream temperatures and pressures. This unit 

automatically closes the sample inlet and outlet valves upon a high  

temperature and/or pressure condition. 

e) Purging is accomplished with a valve control arrangement whereby the normal sample flow is blocked and the detector purged with a "clean" sample. This facilitates detector calibration by establishing the 

background level and aids in verifying sample activity level. 

f) The control and indicating assembly in the control room provides remote 

access to radiation monitoring functions. This assembly provides  

monitoring functions, readout display of monitored data, and status 

alarm indication for each channel. 

g) The electronic mass flow measurement system is calibrated from 0 to 5 standard cubic feet per minute. A local sight flow gauge is provided 

for reference only, and must be compensated for actual flow rate.  

Alarm lights are actuated by the following: 
: a. Flow transmitter (low and high flow). 
: b. The pressure/temperature sensors (high pressure/high temperature). 
: c. The filter paper sensor (paper drive malfunction). 
: d. Failure of any microprocessor controlled self test 

On both units, one common alarm light is turned off, an annunciator is 

actuated, and supplemental information is available to the control room 

operator. 

Plant Vent Gas Monitors (R-14 and RaD 6304) 

The plant vent gas monitors detect radiation passing through the plant vent 

to the atmosphere. Each detector consists of a thin-walled, self-quenching 

type Geiger-Mueller tube (high sensitivity beta-gamma detector) operated in 

parallel with an impedance matching network. 

11.2-18 Revised 04/17/2013 Monitor R-14 has a maximum sensitivity of 5 x 10
-7 &#xb5;Ci/cc. The alarm setpoint for this monitor is determined by and set in accordance with the methodology and parameters of the Turkey Point ODCM. ODCM implementation is 

required by Technical Specification 6.8. 

Remote indication and annunciation of R-14 is provided on the Waste Disposal 

System control board in the Control Room. On high radiation level alarm the 

gas release valve in the Waste Disposal System is automatically closed. 

Monitor RaD 6304 covers a range from 10
-7 to 10 5 &#xb5;Ci/cc for Xe-133. It transmits a pulse signal to the control console in the computer room. High 

radiation, intermediate radiation and rate of rise alarms are provided. RaD-

6304 also functions to collect halogens and particulates on filter elements 

for later analysis in compliance with NUREG-0737, Item II.F.1.2, "Sampling 

and Analysis of Plant Effluents", and Regulatory Guide 1.97. 

Condenser Air Ejector Monitors (R3-15, R4-15, RaD-3-6417 & RaD-4-6417)

Each channel monitors the discharge from the air ejector exhaust header of 

the condenser for gaseous radiation which is indicative of a primary to 

secondary system leak. 

R-*-15 use a single inline beta scintillator while RAD-*-6417 SPINGs use a beta scintillation counter for low range noble gas and emergency-compensated G-M detectors for medium and high range noble gas. All detectors monitor a fixed volume sufficiently shielded to prevent background radiation from reducing maximum sensitivity. R-*-15 has a range of 1.0E-07 to 1.0E-01 
&#xb5;Ci/cc for Kr-85. RAD-*-6417 has a range from 1.0E-07 to 1.0E+05 &#xb5;Ci/cc for Xe-133. The alarm setpoints for these monitors are determined by and set in accordance with the methodology and parameters of the Turkey Point ODCM.
ODCM implementation is required by Technical Specification 6.8 

Gaseous radioactive effluent releases via the steam jet air ejectors on the main condensers are monitored for iodine, particulate, and noble gas activity by RaD-3-6417 and RaD-4-6417 steam jet air ejector vent monitors. The ODCM requires the gaseous effluent from the steam jet air ejector vents to be continuously sampled and analyzed weekly for radioactive iodine and particulates during plant operating Modes 1-4, when primary-to-secondary leakage is detected. The Technical Specifications require the steam jet air ejector vents to be continuously monitored for high-range noble gas activity during plant operating Modes 1-3, while the ODCM requires continuous monitoring for noble gas activity releases during Modes 1-4. As described in Section 3.0 of the ODCM, iodine and particulate sampling of the steam jet air ejector vents is permitted to be performed using developed compensation factors, which estimate the iodine and particulate activity release concentrations as a function of the noble gas concentrations. 

11.2-19 Revised 04/17/2013 C26C26 Component Cooling Liquid Monitors (R3-17A, R3-17B, R4-17A & R4-17B)
Each channel continuously monitors the component cooling loop of the 

Auxiliary Coolant System for radiation indicative of a leak of reactor 

coolant from the Reactor Coolant System and/or the residual heat removal loop 

in the Auxiliary Coolant System. A scintillation counter is located in an 

inline well. A high-radiation level alarm signal initiates closure of the 

valve located in the component cooling head tank vent line to prevent 

radioactive gas release. 

The measuring range of each monitor is 10
-5 to 10-2 &#xb5;Ci/cc. The alarm setpoints for these monitors are determined by and set in accordance with the 

methodology and parameters of the Turkey Point ODCM. ODCM implementation is 

required by Technical Specification 6.8. 

Waste Disposal System Liquid Effluent Monitor (R-18) 

This channel continuously monitors all Waste Disposal System liquid releases 

from the plant. Automatic valve closure action is initiated by this monitor 

to prevent further release after a high-radiation level is indicated and 

alarmed. A scintillation counted and holdup tank assembly monitors these 

effluent discharges. Remote indication and annunciation are provided on the 

Waste Disposal System control board. 

The measuring range of this monitor is 10
-5 to 10-2 &#xb5;Ci/cc. The alarm setpoint for this monitor is determined by and set in accordance with the 

methodology and parameters of the Turkey Point ODCM. ODCM implementation is 

required by Technical Specification 6.8. 

Steam Generator Liquid Sample Monitors (R3-19 & R4-19)

Each channel monitors the liquid phase of the secondary side of the steam 

generators for radiation, which would indicate a primary-to-secondary system 

leak, providing backup information to that of the condenser air removal gas 

monitor. Samples from the bottom of each of the steam generators are mixed 

in a common header and the common sample is monitored by a scintillation 

counter and holdup tank assembly. Upon indication of a high-radiation level, blowdown is automatically isolated. Each steam generator is sampled in order 

to determine the source of the activity. This sampling sequence is achieved 

by manually obtaining steam generator liquid samples at the primary sample 

sink for laboratory analysis after allotting sufficient time for sample 

equilibrium to be established. 

A high-radiation level signal will close the isolation valves in the sample 

lines, the discharge from the blowdown tank to the circulating water 

discharge (environment), and the blowdown recovery flow control valves. 

11.2-20 Revised 04/17/2013 The measuring range of each monitor is 10
-5 to 10-2 &#xb5;Ci/cc. The set point is selected to transfer the blowdown as noted above, at an activity concentration equivalent to no more than 6.1 x 10
-8 &#xb5;Ci/cc in the circulating water. The alarm setpoints for these monitors are determined by and set in 

accordance with the methodology and parameters of the Turkey Point ODCM.  

ODCM implementation is required by Technical Specification 6.8. 

In channels R-18, and R-19, a photomultiplier tube-scintillation crystal (NaI) combination, mounted in a hermetically sealed unit, is used for liquid 

effluent radiation actuation. Lead shielding is provided to reduce the 

background level so it does not interfere with detector's sensitivity. The 

in-line, fixed volume container is an integral part of the detector unit. 

Main Steam Line Monitor (RAD-6426)

The Main Steam Line High-Range Noble Gas Effluent Monitor (RAD-6426) was 

installed at Turkey point as a result of actions required following the 

accident at TMI. RAD-6426 is used in post-accident monitoring as required to 

meet the requirements of Regulatory Guide 1.97, Revision 3. Monitor RAD-6426 

is identified a Type E (Effluent Release Monitoring), Category 2 Variable (instrumentation designated for indicating system operating status). 

The function of RAD-6426 is to detect and measure concentrations of noble gas 

fission products in plant gaseous effluents during and following an accident, and to provide the plant operator and emergency planning agencies with 

information on plant releases of noble gases. RAD-6426 is not included in 

the current Probabilistic Risk Assessment (PRA) and is not a Maintenance Rule 

risk-significant component. 

The Main Steam Line Monitor design uses two Geiger-Muller detectors within 

one assembly with overlapping ranges placed adjacent to each steam line, upstream of the Atmospheric Dump Valves and Main Steam Safety Valves, to 

detect high-energy gammas that penetrate the pipe wall. Each detector 

assembly is shielded in order to protect the detectors from background 

radiation. The detector assembly response to the high-energy gammas is then 

analytically correlated to the total noble gas volumetric activity in the 

steam line. Each detector assembly has a range from 10
-1 to 10 3 &#xb5;Ci/cc to meet R.G 1.97 requirements. The output from the Main Steam Line Monitors 

does not go to the Radiation Monitoring Cabinets in the Control Room, but is 

an input to the Distributed Control System (DCS), which provides the monitor 

information to displays (ERDADS) in the Control Room, Technical Support 

Center, and Emergency Offsite Facility. 

As a Category 2, Type E instrument, RAD-6426 does not meet any of the 10CFR 

50.36(c)(2)(ii) screening criteria for inclusion in the Technical 

Specifications Post Accident Monitoring Table. 

11.2-21 Revised 01/08/2014 C27 As a result, a License Amendment (Reference 6) was approved to relocate the Main Steam Line Monitor Limiting Conditions for Operation and Surveillance Requirements from the Technical Specifications to the UFSAR and related procedures.
The functionality of the monitor is determined by performance of procedures for channel checks, functional testing and channel calibration on a frequency equivalent to the previous Technical Specification Surveillance Requirements.
Specifically, a channel check is required on a monthly basis, and a channel calibration is required on a refueling basis. Performance of these surveillances is governed by plant procedures, in conjunction with the preventative maintenance program. The related procedures contain instructions for notifications and compensatory actions during the times that the monitor is not functional. The monitor is required to be functional in Modes 1, 2 and 3.
Reactor Coolant Letdown Line Activity Monitors (R3-20 & R4-20)
One channel for each unit is provided for detection of fuel clad failure which consists of a fixed position gamma sensitive GM detector for RD-4-20, local indication and signal transmission to a radiation monitoring rack in 

the control room, where it is indicated and alarmed on high activity level. 

RD-3-20 utilizes a gamma sensitive ion chamber detector. A remotely operated 

check source is included for R-4-20. The detector is located on the CVCS 

reactor coolant letdown outside the Containment Building where background 

radiation is relatively low and the flow transit time from the core is 

greater than 40 second to permit 7.2 second N-16 activity to decay to an 

acceptable level. A channel alarm induced by a rapid rise in coolant 

activity signals the requirement to take and count a coolant sample. The 

alarm setpoints for these monitors are determined by and set in accordance 

with the methodology and parameters of the Turkey Point ODCM. ODCM 

implementation is required by Technical Specification 6.8. For R-3-20 an 

internally generated test function is utilized as described previously.
Spent Fuel Pool Vent Monitor - Unit 3 (RaD-3-6418)
The Spent Fuel Pool Vent Monitor detects radiation passing through the Unit 

3 spent fuel pool vent to atmosphere. A beta-gamma sensitive Geiger-Mueller 

tube is used to monitor the gaseous radiation level. Monitor RaD-3-6418 

covers a range from 10
-7 to 10 5 microcuries per cc for Xe-133. Indication and alarms are provided on the console in the cable spreading room.  

RaD-3-6418 also functions to collect halogens and particulates on filter 

elements for later analysis in compliance with NUREG-0737, Item II.F.1.2, "Sampling and Analysis of Plant Effluents", and Regulatory Guide 1.97. 

11.2-22 Revised 04/17/2013 C26C26C26C26 Area Radiation Monitoring System

This system consists of channels which monitor radiation levels in various 

areas. These areas are as follows: 

Detector Tag No. Channel No.             Area Monitor 

RD-3-1401   1 Unit 3 Cntmt Personnel Access Hatch  RD-3-1402   2 Unit 3 Cntmt Refueling Floor El. 58' RD-3-1403   3 Unit 3 Cntmt Incore Instr. Equip.
RD-4-1404   4 Unit 4 Cntmt Personnel Access Hatch RD-4-1405   5 Unit 4 Cntmt Refueling Floor El. 58' RD-4-1406   6 Unit 4 Cntmt Incore Instr. Equip.
RD-3-1407   7 Unit 3 Spent Fuel Pit Transfer Canal RD-4-1408   8 Unit 4 Spent Fuel Pit Transfer Canal RD-1409   9 Aux. Bldg. Laundry Tank and Pump Room RD-1410  10 Aux. Bldg. Chemical Storage Area RD-4-1411  11 Unit 4 Cask Handling Facility  RD-3-1412  12 Unit 3 Cask Handling Facility  RD-3-1413  13 Aux. Bldg. Outside Unit 3 Sample Room RD-4-1414  14 Aux. Bldg. Outside Unit 4 Sample Room RD-3-1415  15 Aux. Bldg. North End of N/S Corridor RD-4-1416  16 Aux. Bldg. South End of N/S Corridor RD-1417  17 Aux. Bldg. East End of E/W Corridor RD-1418  18 Aux. Bldg. West End of E/W Corridor RD-3-1419  19 Unit 3 Spent Fuel Pit Exhaust RD-1420  20 Control Room RD-3-1421  21 Unit 3 Spent Fuel Building North Wall RD-4-1422  22 Unit 4 Spent Fuel Building South Wall RD-3-1423  23 Unit 3 New Fuel Building RD-4-1424  24 Unit 4 New Fuel Building 

===System Description===

Each of the channels is identical, and each channel is comprised of a 

detector, preamplifier, local indicator and a remote cabinet mounted 

indicator in the Control Room. 

11.2-23 Revised 04/17/2013 C26 Channels 21 - 24 provide accidental criticality monitoring in accordance with 10 CFR 50.68(b) (Reference 1). Upon implementation of the license amendment 

for the installation of the spent fuel pool cask racks (References 2 and 3),
the spent fuel pool licensing basis was changed to 10 CFR 50.68(b) from 

compliance with 10 CFR 70.24. 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." 

The Detector

This is composed of a matched ion chamber and preamplifier pair. Calibration 

constants determined by the manufacturer are used by the preamplifier 

assembly to optimize the combined detector and preamplifier response curve.  

The calibration constants for a channel are entered at the remote indicator 

in the Control Room. The ion chamber and the preamplifier are mounted 

separately. The preamplifier converts current from the ion chamber into an 

analog logarithmic DC output to drive the local meter. The preamplifier also 

converts the ion chamber current into a digital signal which is transmitted 

to the remote indicator. Any failure of the preamplifier will activate an 

alarm at the channel indicator in the Control Room. High voltage for the ion 

chamber is developed and controlled in the preamplifier assembly. 

The Local Indicator

The DC output from the detector passes through this indicator which is 

located in a separate box from the preamplifier. Radiation levels are 

indicated on a logarithmic scale which is calibrated from 10
-1 to 10 7 mR/hr. High radiation levels actuate a horn and a red flashing light locally. The 

lowest decade of the meter scale is corrected for "live zero". 

11.2-24 Revised 04/17/2013 The Remote Indicator This is a cabinet mounted module which accepts the signal from the 

preamplifier via a digital highway. The signal is processed to provide two 

visual displays, one current output, and an alarm relay output. Other 

outputs are available for future modifications. Visual displays are a five 

digit display of the radiation value and a multi-color bargraph indicator 

which covers the range of 10
-1 to 10 7 mR/hr. The bargraph has three LED segments per decade. The bargraph will change color in the event of an alarm 

condition. Front panel alarm indicators and rear panel output relays for 

alarm annunciation are also included. Front panel pushbuttons are provided 

to turn power on/off, display alarm limit set points, to acknowledge alarms, and to activate a check source function. Analog outputs of 4-20 ma is 

connected for recorder and computer monitoring. Analog output of 0 to 10 VDC 

is available but not connected. A communication loop transfers data between 

the remote indicator and the preamplifier. 

Five LEDs are used to provide visual indication of status on the front face 

of the remote indicator. They are as follows: 
: 1. HIGH Alarm Red LED 
: 2. WARN Alarm Amber LED (Not Used) 
: 3. Fail Alarm Red LED 
: 4. Range Alarm Red LED 
: 5. Check Source Green LED 

Only the high alarm is connected to an annunciator window on the "Common X" panel. The high alarm will flash until acknowledged. The WARN set point is 

not used. The fail and range alarms are not connected to annunciation. The 

check source LED is lit while the check source function is activated. 

Radiation Monitoring System Cabinet

All of the remote indicators are centralized in one cabinet which is located 

conveniently in the Control Room. The cabinet houses 24 remote indicators 

and a 30 point recorder. Each remote indicator is provided 120 VAC power 

within the cabinet. Each remote indicator provides all power to its channel 

including the local preamplifier, local alarm light and local horn. The 

recorder sequentially records the outputs from the remote indicators at a 

chart speed of 2 inches per hour and a print rate of 10 seconds per point. 

11.2-25 Revised 04/17/2013 Health Physics Program Facilities and Access Provisions The facility has been divided into four basic areas: 
: 1. Controlled Area - As defined in 10CFR20, means an area, outside of a restricted area but inside the site boundary, access to which can be limited by the licensee for any reason. This area, which includes the cooling canals, comprises the site within the boundary shown in Figure 2.2-4. 2. Restricted Area - As defined in 10CFR20, means an area to which is limited by the licensee for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials. Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area. Restricted areas are located within the security fence shown in Figure 11.2-3. 
: 3. Generating Station Area - This area, also referred to as the protected area (PA), is that within the security fence, shown in Figure 11.2-3, and is occupied by the two nuclear units and their associated structures. Access to the Generating Station Area is through a guarded gate. 4. Radiation Controlled Area - The Radiation Controlled Area (RCA) is shown in Figure 11.2-3. This area includes that in which radioactive materials and radiation above 0.5 mrem/hr may be present. The Radiation Controlled Area includes the auxiliary building, Units No. 3 & 4 containment, fuel handling buildings, waste handling facility building, dry storage warehouse, the steam generator storage building, and the Low Level Waste Storage Facility. It does not include the rod control switchgear rooms. Access to the Radiation Controlled Area is limited to those individuals authorized for entry. Entry into the Radiation Controlled Area is through a clearly marked Radiation Control Point.
Restricted Areas may be established outside the RCA (but within the Controlled Area) in accordance with 10CFR Parts 19 and 20.
Any area inside the Radiation Controlled Area in which radioactive materials and radiation may be present shall be surveyed, classified and conspicuously posted with the appropriate radiation caution sign. The Radiation Controlled Area dress facility is employed as a protective clothing change area and storage area. 

11.2-26 Revised 12/05/2014 C27 Personnel decontamination showers are located in the Decontamination Shower Facility, located in the north end of the dress facility. 

All personnel monitor themselves on leaving the radiation controlled areas. 

Administrative and physical security measures are employed to prevent 

unauthorized entry of personnel to any designated high radiation area or 

contaminated area. These measures include the following: 
: 1. Areas accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 0.100 rem in 

one (1) hour at 30 centimeters from the radiation source or from any 

surface that the radiation penetrates, are barricaded and conspicuously 

posted as "high radiation areas."  Administrative controls require the 

issuance of a Radiation Work Permit (RWP) prior to entry to any high 

radiation areas or contaminated area. 
: 2. Locked doors are provided to prevent unauthorized entry into those areas in which the radiation levels could result in an individual receiving a 

dose equivalent in excess of 1.0 rem in one (1) hour at 30 centimeters 

from the radiation source or from any surface that the radiation 

penetrates. Doors shall remain locked, except during periods of access 

by personnel under an approved RWP. For individual high radiation areas 

that are located within large areas, such as, the pressurized water 

reactor (PWR) containment, where no enclosure exists for purposes of 

locking, and where no enclosure can be reasonably constructed around the 

individual area, that individual area shall be barricaded, conspicuously 

posted, and a flashing light shall be activated as a warning device. 

11.2-27 Revised 04/17/2013  
: 3. Any individual or group of individuals permitted to enter a high radiation area is provided with or accompanied by one or more of the following:  
: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area. 
: b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated 

dose is received. Entry into such areas with this monitoring 

device may be made after the dose levels in the area have been 

established and personnel have been made knowledgeable of them. 
: c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for 

providing positive control over the activities within the area and 

shall perform periodic radiation surveillance at the frequency 

specified by the health physics supervisor on the Radiation Work 

Permit.  
: 4. All personnel are required to wear protective clothing for entry into designated contamination areas. The areas involved are decontaminated as 

necessary to prevent the spread of contamination. Decontamination is 

performed under the direction of health physics personnel. 

Personnel Monitoring

The official and permanent record of accumulated external radiation exposure 

received by individuals is obtained principally from the interpretation of 

thermoluminescent dosimeters (TLD). Direct reading dosimeters (which include 

both self-reading pocket ionization chambers and digital alarming dosimeters) 

provide day-by-day indication of external radiation exposure. 

All plant assigned personnel subject to occupational radiation exposure are 

issued beta-gamma thermoluminescent dosimeters (TLDs) and are required to 

wear them at all times practical while within the Radiation Controlled Area. 

Neutron sensitive TLDs are issued to personnel whenever a significant neutron 

exposure is possible. 

11.2-28 Revised 04/17/2013 Plant assigned personnel are issued TLDs at the entrance to the Radiation Controlled Area and return them prior to leaving at the end of the day. The 

TLDs are processed on a routine basis. Personnel TLDs may also be processed 

for administrative exposure control purposes or when it appears that an 

overexposure may have occurred. 

Direct reading dosimeters are issued, in addition to the TLD badge, to 

personnel working in the Radiation Controlled Area. Direct reading 

dosimeters are read, recorded and re-zeroed regularly. Dosimeter records 

furnish the exposure data for the administrative control of radiation 

exposure. 

Special or additional personnel monitoring devices are issued as may be 

required under unusual conditions. For example, finger rings may be 

prescribed for monitoring exposure to the hands. 

Non-qualified personnel entering the Radiation Controlled Area are escorted 

by qualified personnel and are issued personnel monitoring devices as 

appropriate prior to entering the Radiation Controlled Area. An escort may 

not be required for those who have received the necessary radiation 

protection training when this arrangement is approved by the Health Physics 

Supervisor and authorized by the Plant Manager - Nuclear. 

Personnel Protective Equipment

The nature of the work to be done is the governing factor in the selection of 

protective clothing to be worn in the Radiation Controlled Area. The 

protective apparel available include shoe covers, head covers, gloves, and 

coveralls or lab coats. Additional items of specialized apparel such as 

plastic or rubber suits, face shields, and respirators are also available.  

health physics-trained personnel shall evaluate the radiological  conditions 

and specify the required items of protective clothing to be worn. 

11.2-29 Revised 04/17/2013 Process or other engineering controls (e.g., containment or ventilation) are used, to the extent practical, to control the concentrations of radioactive 

material in the air. When it is not practical to apply process or other 

engineering controls to control the concentrations of radioactive material in 

the air to values below those that define an airborne radioactivity area, the 

following are used, consistent with maintaining the total effective dose 

equivalent (TEDE) as low as reasonably achievable: 
: a. Control of access; 
: b. Limitation of exposure times; 
: c. Use of respiratory protection equipment; or 
: d. Other controls. 

Respiratory protection equipment selected provides a protection factor 

greater than the multiple by which peak concentrations of airborne 

radioactive materials in the working area are expected to exceed the values 

specified in 10 CFR 20. If the selection of a respiratory protection device 

with a protection factor greater than the multiple defined in the preceding 

sentence is inconsistent with the goal of keeping the TEDE as low as 

reasonably achievable, respiratory protection equipment with a lower 

protection factor may be selected only if such a selection would result in 

keeping TEDE as low as reasonably achievable. 

Respirator devices available for use include: 
: 1. Full-face respirator (filter, filter/charcoal canister, or supplied air) 
: 2. Air-fed hoods (supplied air) 
: 3. Self-contained breathing apparatus 

Self-contained or supplied air breathing apparatus are available for use in a 

situation involving exposure to gaseous activity or oxygen deficient 

atmospheres. 

Respirators are maintained by checking for mechanical defects, contamination, and cleanliness by health physics trained personnel. 

11.2-30 Revised 04/17/2013 Monitoring Instrumentation A Health Physics Room and Radiochemistry laboratory are provided for the 

health physics and chemistry personnel. These facilities include both 

laboratory and counting rooms. These are equipped to analyze routine air 

samples and contamination swipe surveys. Areas are available for the storage 

of portable radiation survey instruments, respiratory protection equipment 

and contamination control supplies. 

A portal monitor is located at the personnel exits from the Protected Area 

and provides a final radiation survey of all personnel leaving the Protected 

Area. 

The types of portable radiation survey instruments available for routine 

monitoring functions are listed in Table 11.2-9. 

Survey instruments are calibrated periodically, and maintenance records are 

provided for each instrument according to plant operational procedures. 

11.2.4 EVALUATION 

Evaluation of LOCA Control Room Dose This section describes the shielded dose determined during the reanalysis of events performed under the Regulatory Guide 1.183 (Reference 7) methodology with Alternative Source Term (AST).
The total control room dose requires the calculation of direct shine dose contributions from:
* the radioactive material on the control room filters,
* the radioactive plume in the environment, and
* the activity in the primary containment atmosphere  through the containment walls, and  through the purge line penetration (that has line of sight to the control room).
The limiting contribution to the total dose to the operators from direct radiation sources such as the control room filters, the containment atmosphere, and the released radioactive plume were calculated for a LOCA/MHA. The 30-day direct shine dose to a person in the control room, considering occupancy, is provided in Table 11.2-12. 

11.2-31 Revised 04/17/2013 C26 Direct shine dose is determined from three different sources to the control room operator after a LOCA/MHA. These sources are the containment walls, the purge duct penetration area (different shielding than containment walls, but same source term), the control room make-up and recirculating air filter and the external cloud that envelops the control room. All other sources of direct shine dose are considered negligible. The MicroShield 5 code is used to determine direct shine exposure to a dose point located in the control room. The exposure results from the series of cases for each source location were then corrected for occupancy using the occupancy factors specified in Regulatory Guide 1.183. The cumulative exposure and dose are subsequently calculated to yield the total 30-day direct shine dose from each source.
Operator dose during a design basis LOCA for actions outside the control room are evaluated in detail in UFSAR Section 14.3.5.
Evaluation of Vital Area Access Outside the Control Room The Turkey Point shielding design ensures that radiation to personnel performing vital accident mitigating steps outside the Control Room is within the 5 rem dose limit of 10 CFR 50, Appendix A, GDC 19, in compliance with Item II.B.2 of NUREG-0737.
Operating procedures have been evaluated to identify the subset of actions required for accident mitigation that occur outside the Control Room envelope. For each of these actions, a conservative post-accident dose rate has been developed based on the following: 
 - Mission location - Time after accident - Contributing sources of radioactivity - Available shielding Table 11.2-13 identifies each of the credited actions occurring outside the Control Room envelope and the maximum dose an operator could receive performing each mission. The doses presented in Table 11.2-13 include the transit path from the Control Room to each vital access area. In all cases, the resulting dose is within the 5 rem limit of GDC 19.
The principal contributors to mission dose outside containment are radionuclides from three distinct fluid volumes present during accident conditions: the containment atmosphere, pressurized RCS sample or letdown fluid, and depressurized sump fluid.   

11.2-32 Revised 04/17/2013 C26 The total isotope inventory for the Turkey Point whole-core source term, presented in Table 14.3.5-7, is multiplied by appropriate core release fractions to obtain the radionuclide inventory specific to each of these three sources:
Source A - Containment atmosphere - 100% noble gases and 25% halogens Source B - Sample or letdown streams - 100% noble gases, 50% halogens, and 1%
remainders Source C - Sump fluid - 50% halogens and 1% remainders To varying degrees, these sources contribute to mission dose due to each mission area's proximity to the Containment Building wall, containment penetrations, or sample or recirculating sump fluid piping. In addition, dose from containment atmosphere leakage to the environs surrounding the mission areas is also considered.
The dose contribution from immersion in the post-accident cloud of containment atmosphere leakage (Source A) credits the operator with wearing a self contained breathing apparatus (SCBA) that eliminates at least 95% of inhaled radioactive contamination.
As shown in Figures 9.9-1, 9.9-4 and 9.4-5, the control room has its own 

independent ventilating  system. In the event of a MHA the control room 

ventilating system is  automatically placed in the recirculating mode as 

discussed in Section 9.9. The control room ventilation system is supplied by 

emergency power.   

The radiation sources used with the original auxiliary shielding design criteria resulted from a loss of coolant accident caused by a double-ended rupture of a reactor coolant loop where the engineered safety features function to prevent melting of fuel cladding and to limit the metal-water reaction to a negligible amount. This would result in only the fission products which are in the fuel rod gaps being released to the containment. It was assumed that all gap activity, except that of the noble gases, would be absorbed in the sump water which flows in the residual heat removal loop and associated equipment. 

11.2-33 Revised 04/17/2013 C26C26 Mission dose evaluations prepared for EPU conditions are based on recirculating sump water containing 1% failed fuel and 50% of the core halogen inventory ("Source C" described above). Gamma energy release rates for Source C are presented in Table 11.2-11. 

The radioactivity in the containment could be an additional source of 

radiation to the Auxiliary Building following a loss-of-coolant accident.  

However, the radiological exposure rate in the Auxiliary Building from this 

source would be less than one percent of that from heat removal system 

piping. Operator dose from streaming radiation through containment 

penetrations is considered on a case-by-case basis. 

An evaluation was made of direct radiation levels surrounding recirculation piping of varying size. The evaluation was based on the radiation sources and evaluation parameters tabulated on Table 11.2-11. The results of the evaluation are presented in Figure 11.2-4, showing dose rates and 31-day integrated dose as a function of distance from a 20-ft length of recirculation piping. 

If maintenance of equipment near the recirculation loop is absolutely 

essential to the continued operation of the engineered safety features during 

the recirculation phase, local shielding would permit some operations in the 

vicinity of the loop. 

If maintenance directly on the loop proper is required, such operations would 

be limited in duration as radiation levels adjacent to equipment containing 

the sump water and fission products might be as high as 200 to 300 rem per 

hour shortly after the initiation of recirculation. Any such emergency 

maintenance operations described above could be carried out behind portable 

shielding and using portable breathing equipment to limit the inhalation 

hazard from possibly leaking components.  

11.2-34 Revised 04/17/2013 C26C26C26C26 11.2.5  TEST AND INSPECTION CAPABILITY   

Complete radiation surveys are made throughout the containment and Auxiliary 

Building during initial phases of start-up for comparison with future 

periodic surveys. Survey data are compared to design levels up to rated full 

power. Survey data are evaluated and reviewed to ensure that operating 

personnel will not be exposed in excess of applicable limits.   

Checks of the waste liquid effluent monitors response to a test source are 

made periodically.   

The plant vent gas monitor is calibrated during shutdown, and normal response 

of each monitor can be tested at any time using a, remotely operated test 

source to verify the instruments response and alarm functions.   

11.2-35 Revised 04/17/2013 11.

==2.6  REFERENCES==
: 1. 10 CFR 50.68, "Criticality Accident Requirements". 
: 2. NRC Letter to FPL, "Turkey Point Units 3 and 4 - Issuance of Amendments Regarding Temporary Spent Fuel Pool Cask Racks" (TAC Nos.MB 6909 and MB 

6910), License Amendments 226/222, November 24, 2004. 
: 3. FPL Letter L-2003-213 to the NRC, "Turkey Point Units 3 and 4 - RAI Response for Addition of Spent Fuel Pool Cask Area Rack Amendment",
September 8, 2003. 
: 4. FPL Letter L-80-16 to the NRC, "NUREG-578 Short Term Requirements,"
January 11, 1980. 
: 5. FPL Letter L-81-285 to the NRC, "Post TMI Requirements - Control Room Habitability", July 9, 1981. 
: 6. NRC Letter (Accession No. ML12024A104) Jason C. Paige (NRC) to Mano Nazar (FPL), " Turkey Point, Units 3 and 4 - Issuance of Amendments Regarding High Range - Noble Gas Effluent Monitors, Main Steam Lines Accident Monitoring Instrumentation (TAC Nos. ME6891 and ME 6892)", dated June 15, 2012. 
: 7. USNRC, Regulatory Guide 1.183, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants", July 2000. 

11.2-36 Revised 04/17/2013 C26C26 TABLE 11.2-1   RADIATION ZONE CLASSIFICATIONS 

Zone Condition of Occupancy Max. Dose Rate (1% Fuel Defects)          mrem/hr 

I  Normal Occupancy  <
0.5  II  Periodic Occupancy  <
2.5  III  Short Specific Occupancy  <
: 15. IV  Minimal Occupancy  <
100  V  Controlled Access  > 100 

TABLE 11.2-2   PRIMARY SHIELD NEUTRON FLUX AND DESIGN PARAMETERS 

Original Calculated Neutron Flux / Shield Horizontal Mid-plane (1)      Energy Group Incident Flux Leakage Flux n/cm 2-sec     n/cm 2-sec         E >1 Mev. 1.2 x 10 9  1.15 x 10 2  5.3 Kev. <
E <1 Mev. 1.7 x 10 10  2.75 x 10 2  .625 < E < 5.3 Kev. 9.8 x 10 9  4.75 x 10 2  E < .625 ev. 4.1 x 10 9  2.4 x 10 5 Original Core Design Parameters (1)  A. Thermal Power Rating (100% power)                   2300 MWt (1)  B. Effective Dimensions   1. Height 12.0 ft   2. Diameter 9.98 ft  C. Volume Fractions 
: 1. UO 2 0.3022   2. Zircaloy 0.0933   3. Water 0.5980   4. 304 Stainless Steel 0.0061   5. 718 Inconel 0.0004  D. Operating Times (Equivalent Full Power Hours) 
: 1. Initial cycle 11,700   2. Equilibrium cycle 8,200  E. Mode of Operation Base Load  F. Fraction of Fuel Rods with Cladding Defects 0.01                       

NOTE : 
: 1. Plant shielding was designed at the plant's initial power rating.
Radiological impacts due to the 1995 thermal power uprate were analyzed and found not to be significant. The design was also shown to be adequate for the 2012 extended power uprate, due to the low-leakage fuel management methods that minimize flux near the core periphery. 

Revised 04/17/2013 C26C26 TABLE 11.2-3   ORIGINAL SECONDARY SHIELD DESIGN PARAMETERS 

Core power density                                              86.4 w/cc (1) Reactor coolant liquid volume                                   9400 ft 3  Maximum purification letdown rates                              120 gpm 

Average water temperature in core                               580 o F System operating pressure                                       2250 psia 

Reactor coolant transit times: 

Core                                                       0.9 sec. 

Core exit to steam generator inlet                         2.0 sec. 

Steam generator inlet channel                              0.6 sec. 

Steam generator tubes                                      3.2 sec. 

Steam generator tubes to vessel inlet                      2.7 sec. 

Vessel inlet to core                                       2.1 sec. 

Total Out of Core                                               10.6 sec. 

Total power dose rate outside secondary shield                                      < 1 mr/hr (1)   

NOTE :  1. Plant shielding was designed at the plant's initial power rating.
Radiological impacts outside the secondary shield due to thermal power uprate and extended power uprate were analyzed and found not to be significant. 

Revised 04/17/2013 C26C26C26 TABLE 11.2-4   ACCIDENT SHIELD DESIGN PARAMETERS DELETED                          

Revised 04/17/2013 C26 TABLE 11.2-5   REFUELING SHIELD DESIGN PARAMETERS DELETED                          

Revised 04/17/2013 C26 TABLE 11.2-6   PRINCIPAL AUXILIARY SHIELDING 

Component Concrete Shield Thickness, Ft. - In.
Demineralizers 3-6 

Charging pumps 1-6 Holdup tank 2-6 Volume control tank 3-0 Reactor coolant filter 2-9 Gas stripper 2-6 Gas decay tanks 4-0 Waste gas compressor 2-8 Waste evaporators 2-0 Waste holdup tank, Aux. Bldg.
1-0 to 1-6 Waste holdup tank, Rad. Fac.
2-0 Distillate demineralizers 1-0 Waste monitor tanks 1-0 Waste holdup/mixing tanks 3-0 Cement mixers 3-0 Design parameters for the auxiliary shielding include: 

Core thermal power 2652 MWt RCS Activity NOTE 1  Dose rate outside auxiliary 

building and radwaste facility 
<1 mr/hr Dose rate in the building 

outside shield walls 
<2.5 mr/hr 

NOTE 1: The auxiliary shielding design was re-evaluated and found acceptable for EPU using scaling factors to compare original to EPU dose results. RCS activity at uprate conditions is assumed to be consistent with full power operation at the Technical Specification limit for RCS Dose Equivalent (DE) I-131. 

Revised 04/17/2013 C26C26 TABLE 11.2-7   RADIATION MONITORING SYSTEM CHANNEL SENSITIVITIES 

Channel             Sensitivity Range Detected Isotopes Process R3-11 & R4-11      1.0 x 10
-9 to 1.0 x 10
-6*        I 131 ,I 133 ,Cs 134 ,Cs 137 R3-12 & R4-12      1.0 x 10
-6 to 1.0 x 10
-3*        Kr 85 ,Ar 41 ,Xe 135 ,Xe 133 R-14               5.0 x 10
-7 to 1.0 x 10
-4*        Kr 85 ,Ar 41 ,Xe 135 ,Xe 133 R3-15 & R4-15      1.0 x 10
-7 to 1.0 x 10
-1*        Kr 85 ,Ar 41 ,Xe 135 ,Xe 133 R3-17A, R3-17B,             

R4-17A, R4-17B     1.0 x 10
-6 to 1.0 x 10
-2*        Co 60 ,Mixed Fission Products R-18               1.0 x 10
-5 to 1.0 x 10
-2*        Co 60 ,Mixed Fission Products R3-19, R4-19       1.0 x 10
-5 to 1.0 x 10
-2*        Co 60 ,Mixed Fission Products R3-20, R4-20       1.0 x 10 o to 1.0 x 10
+5**        Kr 85 ,Ar 41 ,Xe 133 ,Xe 135 Area Rl thru R24   1.0 x 10
-1 to 1.0 x 10
+7**         

Notes:
* is given in  Ci/cc            ** is given in mr/hr 

Prefixes R3 or R4 designate Unit #3 or Unit #4. Channels without 

prefix number monitor both units. 

Revised 04/17/2013 C26 TABLE 11.2-7a   RADIATION MONITORING, SYSTEM CHANNEL ALARM SET POINTS 

 [TABLE INTENTIONALLY LEFT BLANK] 

Rev. 13  10/96 TABLE 11.2-8 DETECTING MEDIUM CONDITIONS 

Channel                            Medium             Temperature Range, C 

Area:
R-1                                Air                10-50 

through R-24 

Process: 

R3-11                              Air                10-50 

R4-11                              Air                10-50 

R3-12                              Air                10-50 

R4-12                              Air                10-50 

R-14                               Air                4-50 

R3-15                              Air                10-50 

R4-15                              Air                10-50 

R3-17  A&B                         Water              4-71 

R4-17  A&B                         Water              4-71 

R-18                               Water              15-71 

R3-19                              Water              15-71 

R4-19                              Water              15-71 

R3-20**                            Water              10-80 R4-20*                             Water              10-80
* Detector mounted on outside of pipe carrying medium. 
** Detector mounted external to pipe carrying medium. 

Revised 09/21/2012 C26C26 TABLE 11.2-9 PORTABLE RADIATION SURVEY INSTRUMENTS 

Type Low Range beta-gamma Survey Meter 

Intermediate Range beta-gamma Survey Meter 

High Range beta-gamma Survey Meter 

Personnel Monitoring beta-gamma Survey Instruments 

Neutron Survey Meter 

High Volume Air Particulate Sampler 

Low Volume Air Particulate Samples 

Beta-gamma and gamma Portal Monitors 

Direct Reading Dosimeters (includes both self-reading pocket ion chambers and digital alarming dosimeters) 

Low Level gamma Scintillation Survey Meters 

Alpha Scintillation Survey Meters 

Rev. 13  10/96 TABLE 11.2-10 INSTANTANEOUS RADIATION SOURCES RELEASED TO THE CONTAINMENT FOLLOWING TID-14844 ACCIDENT RELEASE - Mev/sec DELETED                                Revised 04/17/2013 C26 TABLE 11.2-11 LOCA ACTIVITY SOURCES IN CIRCULATING IN RESIDUAL HEAT REMOVAL LOOP AND ASSOCIATED EQUIPMENT - Mev/sec Time After Release Energy MeV 0 hr 1 hr 2 hr 8 hr 24 hr 168 hr 31 day 0.01 1.26E+17 3.37E+16 2.80E+16 1.93E+16 1.28E+16 3.14E+15 8.74E+14 0.025 7.61E+16 1.69E+16 1.38E+16 8.94E+15 5.71E+15 1.95E+15 5.79E+14 0.038 9.69E+16 2.29E+16 1.91E+16 1.45E+16 1.18E+16 5.76E+15 1.15E+15 0.058 1.17E+17 2.51E+16 2.02E+16 1.27E+16 7.86E+15 2.21E+15 7.76E+14 0.085 1.81E+17 3.99E+16 3.08E+16 2.34E+16 2.09E+16 1.11E+16 1.47E+15 0.125 2.70E+17 9.65E+16 8.42E+16 6.66E+16 5.21E+16 1.27E+16 3.11E+15 0.225 8.75E+17 2.50E+17 2.52E+17 2.95E+17 2.10E+17 2.72E+16 3.93E+15 0.375 1.72E+18 6.32E+17 5.63E+17 4.70E+17 4.09E+17 2.31E+17 3.09E+16 0.575 6.90E+18 3.61E+18 2.96E+18 1.56E+18 8.07E+17 9.89E+16 3.50E+16 0.85 1.08E+19 4.17E+18 2.58E+18 6.26E+17 2.81E+17 1.09E+17 6.80E+16 1.25 9.46E+18 3.35E+18 2.69E+18 1.23E+18 2.94E+17 1.29E+16 3.94E+15 1.75 3.17E+18 1.47E+18 1.14E+18 5.33E+17 1.58E+17 5.63E+16 1.56E+16 2.25 1.37E+18 2.64E+17 1.93E+17 7.56E+16 1.77E+16 4.20E+15 1.71E+15 2.75 8.21E+17 4.08E+16 2.28E+16 5.42E+15 4.31E+15 3.27E+15 9.01E+14 3.5 7.21E+17 4.78E+16 1.57E+16 3.79E+14 4.69E+13 3.60E+13 1.13E+13 5 5.82E+17 8.99E+14 2.43E+14 1.24E+12 2.32E+10 3.35E+08 3.27E+08 7 1.16E+15 5.46E+07 5.43E+07 5.43E+07 5.43E+07 5.40E+07 5.26E+07 9.5 3.07E+12 8.48E+06 8.47E+06 8.47E+06 8.47E+06 8.42E+06 8.21E+06 

Revised 04/17/2013 C26 TABLE 11.2-12 CONTROL ROOM DIRECT SHINE SHIELDED DOSE RESULTS USING AST  (EPU CONDITIONS) 

SOURCE  DIRECT SHINE DOSE (rem) Containment  

Walls 0.060 Purge Duct 0.337 External Cloud 0.277 CR Recirculation Filters 0.054 Total 0.728 

Revised 04/17/2013 C26 TABLE 11.2-13  VITAL AREA ACCESS MISSION DOSES MISSION UNIT/ TRAIN/ EQPT DOSE (rem)
Unit 3, Train A, MCC 3C 1.92 Unit 3, Train B, MCC 3B 0.96 Unit 4, Train A, MCC 4C 1.68 Verify Cold Leg Recirculation Capability Unit 4, Train B, MCC 4B 0.96    Unit 3 1.04 Close Radiation Shield 

Doors Unit 4 1.12    Unit 3 3.06 Reset Pressurizer Heater 

Lockout Relay (shift to 

"Emergency Mode")
Unit 4 3.17    Unit 3, MCC3A 3.06 SI Accumulator Isolation 

MOV Breaker Operation Unit 3, MCC3B 0.46  Unit 3, MCC3C 1.59  Unit 4, MCC4A 3.17  Unit 4, MCC4B 1.37  Unit 4, MCC4C 0.46    Unit 3 1.27 Recovery from Failed Open 

Cold Leg Recirculation 

Valve Unit 4 4.80    Unit 3 3.76 Recovery from Failed Open 

Cold Leg Direct RHR 

Injection Valve Unit 4 3.76    Unit 3 1.27 Recovery from Failed  

closed MOV-3/4-869 Unit 4 4.80    Unit 3 2.10 AFE Control Valve Backup 

Nitrogen Bottle Change-out Unit 4 2.10    EDG Fuel Oil Replacement Unit 3 2.19  Unit 4 2.19    EDG Lube Oil Replacement Unit 3 2.02  Unit 4 2.02  

Revised 04/17/2013 C26 

FINAL SAFETY ANALYSIS REPORT FIGURE 11.2-1 

REFER TO ENGINEERING DRAWING 5610-M-50   

Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 RADIATION ZONE DIAGRAM PLAN FULL POWER OPERATION WITH 1%
FAILED FUEL FIGURE 11.2-1 C26C26 

FINAL SAFETY ANALYSIS REPORT FIGURE 11.2-2 

REFER TO ENGINEERING DRAWING 5610-M-51   

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 AREA RADIATION ZONE PLAN FULL POWER OPERATION WITH 1% FAILED FUEL FIGURE 11.2-2

FINAL SAFETY ANALYSIS REPORT FIGURE 11.2-3 

REFER TO ENGINEERING DRAWING 5610-C-2   

REV. 13  (10/96)
FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 GENERAL STATION AREA FIGURE 11.2-3

31- day Integrated EPU Dose from Post - Accident Recirculating 20' Lines 

T=0 hr Post LOCA Dose Rates (R/hr) from 20' Pipe Distance (ft) Nominal Pipe Diameter (inches) 3 6 10 14 Contact 1.11E+05 2.57E+05 4.70E+05 5.39E+05 1 1.87E+04 7.18E+04 1.98E+05 2.73E+05 2 9.22E+03 3.54E+04 9.64E+04 1.31E+05 3 5.95E+03 2.30E+04 6.26E+04 8.55E+04 4 4.26E+03 1.66E+04 4.52E+04 6.23E+04 5 3.23E+03 1.27E+04 3.46E+04 4.80E+04 6 2.54E+03 1.00E+04 2.74E+04 3.82E+04 7 2.05E+03 8.11E+03 2.22E+04 3.12E+04 8 1.68E+03 6.69E+03 1.84E+04 2.59E+04 9 1.41E+03 5.61E+03 1.54E+04 2.18E+04 10 1.19E+03 4.76E+03 1.31E+04 1.86E+04 

Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 LOCA RECIRCULATION PIPING T=0 HR DOSE RATE and 31 - DAY INTEGRATED DOSE  FIGURE 11.2-4 C26 

FINAL SAFETY ANALYSIS REPORT FIGURE 11.2-5 

DELETED    

Revised 04/17/2013 FLORIDA POWER & LIGHT COMPANY TURKEY POINT PLANT UNITS 3 & 4 SENSITIVITY OF DOSE TO ACTIVITY IN THE RESIDUAL HEAT REMOVAL WATER FIGURE 11.2-5 C26 11.3  RADIOACTIVE MATERIALS SAFETY 11.3.1  MATERIALS SAFETY PROGRAM 

Procedures, facilities and equipment for handling and processing of 

radioactive liquid, gaseous and solid wastes are described in Section 11.  

Procedures, facilities and equipment for the safe handling and storage of new 

fuel assemblies and spent fuel assemblies are described in Section 9.5. 

Various radioactive sources are employed to calibrate and/or check the 

process and effluent radiation monitors and the area radiation monitors 

described in Section 11.2.3 and the portable radiation survey instruments 

listed in Table 11.2-9. Check sources that are integral to the area, process 

and effluent monitors are handled and stored by employing the normal Health 

Physics Operating procedures. The same consideration applies to radionuclide 

sources of exempt quantities which are used to periodically check the 

radiation monitoring equipment. 

Radioactive sources purchased or prepared by the Chemistry Department or 

under the direction of the Radiochemist for the calibration, testing or 

standardization of laboratory counting equipment will be stored under 

administrative control in the radiochemistry laboratory or in a designated 

storage area. 

Radioactive sources purchased by the Health Physics Department for the 

calibration, testing or standardization of laboratory counting equipment 

shall be stored under administrative control in the health physics counting 

room, the health physics calibration facility, the radiochemistry laboratory 

or in a designated storage area. 

If a sealed source containing greater than 100 microcuries of beta and/or 

gamma emitting material or 5 microcuries of alpha emitting material is found 

to be leaking greater than or equal to 0.005 microcuries, it shall be 

immediately removed from service. The source will be disposed of in 

accordance with plant waste disposal procedures or repaired. A report 

containing a brief description of the event and remedial action taken, shall 

be made to the Nuclear Regulatory Commission on an annual basis. Records 

shall be maintained current which will include, but are not necessarily 

limited to date received, supplier, isotope, quantity, and date of ultimate 

11.3-1 Rev. 16  10/99 disposal or consumption of the source at which time it will be removed from the inventory list. All documentation accompanying the purchase of any 

sealed source shall become a permanent part of the Health Physics Department 

records. 

Radioactive sources and materials are subject to controls for the purpose of 

radiation protection. These controls include: 

a) Monitoring for external dose rate and removable contamination upon receipt  at the plant and prior to shipment away from the plant. Both 

the packaging surface and the transport vehicle are monitored prior to 

shipment away from the plant. 

b) Each sealed source obtained by license is labeled as to the quantity of activity, isotope and source identification number. The radiation 

symbol is affixed to all of the above sources except those which are 

contained in area and process monitoring components. Radioactive 

sources are stored under administrative control when not in use. 

c) Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below.  
: i. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:  
: 1) With a half-life greater than 30 days (excluding Hydrogen 3), and   2) In any form other than gas.
ii. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and 

11.3-2 Rev. 13  10/96 iii. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and 

following repair or maintenance to the source. 

d) Records on the results of inventories, leak tests and the receipt and final disposition dates shall be maintained for accountable sealed 

sources. The Health Physics Supervisor is responsible for the 

accountability and documentation of accountable sources. 

e) Radiation work permits which provide detailed instructions for all work in radiation, high radiation, and airborne radioactivity areas. 

Radiation work permits are described in the Turkey Point Plant 

Radiation Protection Manual. 

11.3-2a Rev. 16  10/99 In the event of an inventory discrepancy of sealed sources, the Health Physics Supervisor will investigate and determine if the loss may result in a 

substantial hazard to persons in unrestricted areas. If required the loss 

will be reported in accordance with the requirements of 10 CFR 20. 

The sealed sources will be handled and used in accordance with the Turkey 

Point Plant Radiation Protection Manual. Recognized methods for the safe 

handling of radioactive materials are implemented to maintain potential 

external and internal doses at levels that are as low as reasonably 

achievable (ALARA). The radioactive materials safety program is described in 

the Turkey Point Plant Radiation Protection Manual. 

11.3.2  FACILITIES AND EQUIPMENT 

The radiochemistry laboratory consists of a 31'- 6" x 18' room containing a 

fume hood, cabinets, counter-tops, and a counting room, along with necessary 

chemistry hardware. The radiochemistry counting room is located in the 

Health Physics Control Building. 

The fume hood is a five foot wide radioactive model of the Fisher 

Conserv-Air. Air is drawn by a blower-motor which provides a 125 FPM face 

velocity and a 2300 FPM duct velocity. Filtering is provided by a 24" x 24" x 2" prefilter followed by a 24" x 11-1/2" CWS type filter with a 99.95% 

efficiency rating for 0.3 to 0.5 micron-sized particles. This exhaust is fed 

into the plant vent exhaust plenum. Two additional filters (a 1" prefilter 

and a 10" DOP tested absolute filter) filter this exhaust before it is 

released to the plant vent. This exhaust is continuously monitored by 

particulate, iodine, and gaseous detectors. 

Equipment and facilities for the sampling of radioactive liquids and gases 

are described in Section 9.4. The area radiation monitoring and the process 

and effluent monitoring systems are detailed in Subsection 11.2.3. Health 

physics instrumentation is listed in Table 11.2-9. 

11.3-3 Rev. 16  10/99 The health physics facilities include: (a) office area for health physics supervisors and support personnel; (b) records area; (c) computer room; (d) 

areas for controlling Rca access; (e) material release building; (f) 

instrument calibration and radioactive source storage area; and (g) counting 

room.  

11.3.3 PERSONNEL AND PROCEDURES 

The key person responsible for the supervision of the handling and monitoring 

of the materials is the Health Physics Supervisor whose experience and 

qualifications are listed in the Plant Technical Specifications. 

The radiation safety instructions to working personnel appropriate to the 

handling and use of radioactive materials are listed in the Turkey Point 

Plant Radiation Protection Manual. 

Radioactive sources that are subject to the material controls described in 

the Radiation Protection Manual will only be used or handled by or under the 

direction of chemistry and radiation protection personnel. Each individual 

using these sources are familiar with the radiological restrictions and 

limitations placed on their use. These limitations protect both the user and 

the source. 

A comprehensive basic Health Physics Training Program is given to all 

personnel assigned to Turkey Point Units 3 and 4 with unescorted access to 

the RCA. Supervisors are responsible for ensuring that their employees 

receive adequate on the job radiation protection training. The amount and 

type of training depends on the kind of work they perform and where they 

work. Orientation lectures on radiation and radiation protection are given 

to all new employees. In the course of their work, employees will receive 

additional training in radiation protection practices from supervisors, senior co-workers and chemistry and radiation protection personnel. 

11.3-4 Rev. 16  10/99 All personnel must pass a Health Physics examination before they are allowed access to the radiation control area unescorted. Those persons who have not 

successfully completed the Health Physics Training program and examination 

are escorted. 

11.3.4 REQUIRED MATERIALS 

A listing of isotopes, maximum quantities, forms and uses for all purchased 

byproduct, source and special nuclear materials is given in Table 11.3-1. 

Instrumentation check and calibration sources between atomic numbers 3 and 83 

having less than 100 mCi beta/gamma activity or 100 milligrams of source or 

special nuclear material have been excluded from this listing. 

11.3-5 TABLE 11.3-1   BYPRODUCT, SOURCE AND SPECIAL NUCLEAR MATERIALS: RADIOACTIVE SOURCES LISTING

Isotope Quantity Form Use           

U-235              See Section 3.2      See Section 3.2           Reactor fuel 

U-238              See Section 3.2      See Section 3.2           Reactor fuel 

Pu-Be              4 @ = 100 Ci ea      Each source contains      These neutron                two capsules inserted     sources are                     between Sb-Be pellets, no longer                  all sealed in a  used in the                  stainless steel tube core. They             are stored in                                                   the spent fuel pool. 

Any byproduct      Not to exceed        Gas or Liquid             Calibration of 

material with      150 millicuries                                Analytical atomic numbers  of each radio-      Instrumentation between 3 and 83  nuclide       used for radio chemical analysis 

Tritium            Up to 1 Curie        Liquid                    Calibration of 

Analytical instrumentation 

used for radio-chemical 

analysis. 

Americium          Up to 1 mci          Liquid    Calibration of Analytical 

instru-                                                                   mentation used for radio- 

chemical 

analysis. 

Pu-Be              Up to 10 ci          Solid Capsule             Instrument check for 

ex-core reactor 

instrumentation 

Calibrate Health Physics neutron survey instrument-ation 

Americium 241      Up to 10 ci          Solid Capsule             Calibration of Health Physics Beryllium          Equipment 

Rev. 13  10/96 11.4-1 Rev. 17 11.4 Radiological Administrative Controls The following programs shall be established, implemented, and maintained:
11.4.1 In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: 
  (1) Training of personnel,   (2) Procedures for monitoring, and (3) Provisions for maintenance of sampling and analysis equipment.
11.4.2 Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following: 
: 1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM. 
: 2. A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and 
: 3. Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
11.4.3 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.}}