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| number = ML16330A235
| number = ML16330A235
| issue date = 10/29/2016
| issue date = 10/29/2016
| title = Turkey Point, Units 3 & 4, Updated Final Safety Analysis Report, Chapter 11, Waste Disposal and Radiation Protection System
| title = Updated Final Safety Analysis Report, Chapter 11, Waste Disposal and Radiation Protection System
| author name =  
| author name =  
| author affiliation = Florida Power & Light Co
| author affiliation = Florida Power & Light Co
Line 14: Line 14:
| document type = Updated Final Safety Analysis Report (UFSAR)
| document type = Updated Final Safety Analysis Report (UFSAR)
| page count = 121
| page count = 121
| revision = 0
}}
}}
=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 µCi/ml and 7 x 10
-7 µCi/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    ºF
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.}}

Latest revision as of 11:24, 16 March 2019

Updated Final Safety Analysis Report, Chapter 11, Waste Disposal and Radiation Protection System
ML16330A235
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 10/29/2016
From:
Florida Power & Light Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML16330A191 List:
References
L-2016-198
Download: ML16330A235 (121)
v • d • e


Text

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 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 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 µCi/ml and 7 x 10

-7 µCi/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 ºF

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 µCi/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 µCi/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 µCi/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 µCi/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 µCi/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

µCi/cc for Kr-85. RAD-*-6417 has a range from 1.0E-07 to 1.0E+05 µCi/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 µCi/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 µCi/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 µCi/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 µCi/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 µCi/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.

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