ML091400697

Final Safety Analysis Report (Fsar), Amendment 93, Section 12, Radiation Protection Through Page 12.5-8

ML091400697
Person / Time
Site:	Watts Bar
Issue date:	04/30/2009
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
Download: ML091400697 (128)

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WATTS BAR WBNP-76 TABLE OF CONTENTS Section Title Page 12.0 RADIATION PROTECTION 12.1 Assuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA)12.1-1 12.1.1 Policy Considerations 12.1-1 12.1.2 Design Considerations 12.1-1 12.1.3 ALARA Operational Considerations 12.1-1 12.2 RADIATION SOURCES 12.2-3 12.2.1 Contained Sources 12.2-3 12.2.1.1 Primary System Sources 12.2-3 12.2.1.2 Auxiliary Systems Sources 12.2-4 12.2.1.3 Sources During Refueling 12.2-10 12.2.1.4 Maximum Hypothetical Accident (MHA) Sources 12.2-10 12.2.1.5 Condensate Demineralizer Waste Evaporator 12.2-11 12.2.2 Airborne Radioactive Material Sources 12.2-11 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3-1 12.3.1 Facility Design Features 12.3-1 12.3.2 Shielding 12.3-3 12.3.2.1 Design Objectives 12.3-3 12.3.2.2 Design Description 12.3-3 12.3.3 Ventilation 12.3-15 12.3.3.1 Airflow Control 12.3-15 12.3.3.2 Typical System 12.3-15 12.3.3.3 Additional Radiation Controls 12.3-16 12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation 12.3-17 12.3.4.1 Area Radiation Monitoring Instrumentation 12.3-17 12.3.4.2 Airborne Particulate Radioactivity Monitoring 12.3-19 12.3.4.3 Deleted by Amendment 84. 12.3-23 12.3.4.4 Special Radiation Monitors 12.3-23 12.4 DOSE ASSESSMENT 12.4-1 12.5 RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5-1 12.5.1 Organization 12.5-1 12.5.2 Equipment, Instrumentation, and Facilities 12.5-2 12.5.3 Procedures 12.5-4 Table of Contents 12-i

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WATTS BAR WBNP-76 LIST OF TABLES Section Title Table 12.2-1 Chemical And Volume Control System Mixed Bed Demineralizer Table 12.2-2 Chemical And Volume Control System Cation Bed Demineralizer Table 12.2-3 Chemical and Volume Control System Seal Water Return Filter Table 12.2-4 Chemical And Volume Control System-Hcldup Tank Liquid Space Table 12.2-5 Chemical And Volume Control System-Hcidup Tank Vapor Space Table 12.2-6 Chemical And Volume Control System - Evaporator Feed Mixed Bed Ion Exchanger Table 12.2-7 Chemical And Volume Control System-Evaporator Feed Cation Bed Ion Exchanger Table 12.2-8 Gas Stripper And Boric Acid Evaporator Package-Vent Condenser Table 12.2-9 Gas Stripper And Boric Acid Evaporator Package-Evaporator Bottoms Table 12.2-10 Spent Fuel Pool Cooling System-Spent Fuel Pool Demineralizer Table 12.2-11 Spent Fuel Pool Cooling System-refueling Water Purification Filter Table 12.2-12 Core Inventory Table 12.2-13 Irradiated In-Core Detector Drive Wire Sources (MEV/CM-SEC)

Table 12.2-14 Irradiated Ag-In-Cd Control Rod Sources Table 12.2-15 Deleted By Amendment 84 Table 12.2-16 Deleted By Amendment 84 Table 12.2-17 Deleted By Amendment 84 Table 12.2-18 Deleted By Amendment 84 Table 12.2-19 Estimated Average Airborne Radioactivity Concentrations In The Containment Building Table 12.2-20 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Turbine Building Table 12.2-21 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Auxiliary Building Table 12.2-22 Estimated Average Airborne Radioactivity Concentrations In The Instrument Room Table 12.3-1 Deleted by Amendment 84 Table 12.3-2 Deleted by Amendment 84 Table 12.3-3 Ventilation Air Exhaust Points Table 12.3-4 Location of Plant Area Radiation Monitors Table 12.3-5 Airborne Particulate Activity Monitoring Channels Table 12.3-6 Shield Wall Thickness Table 12.3-7 Special Radiation Monitors Table 12.4-1 Restricted Area Boundary Dose Rates For 2000 Hr/Year Exposure Table 12.4-2 Onsite Dose Rates (Mrem/Yr)

List of Tables 12-iii

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WATTS BAR WBNP-91 LIST OF FIGURES Section Title Figure 12.2-1 Neutron Flux Distribution Figure 12.2-2 Gamma Flux Distribution Figure 12.2-3 Reactor Coolant N-16 Activity Figure 12.3-1 Plans - Roof - Radiation Zone Map Figure 12.3-2 Plan El. 772.0 and Above - Radiation Zone Map Figure 12.3-3 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 772.0, 782.0 and 786.0 - Radiation Zone Map Figure 12.3-4 Powerhouse, Auxiliary, Reactor and Control Building - Plan El. 755.0 and 757.0 - Radiation Zone Map Figure 12.3-5 Powerhouse, Units 1 and 2 - Plan El. 757.0 and 755.0 - Radiation Zone Map Figure 12.3-6 Powerhouse, Units 1 and 2 - Plan El. 737.0 and 729.0 - Radiation Zone Map Figure 12.3-7 Auxiliary Building, Units 1 and 2 - Radiation Zone Map, Condensate Demineralizer Figure 12.3-8 Powerhouse, Auxiliary, Reactor and Control Buildings - Radiation Zone Map - Plan El. 729.0 and 737.0 Figure 12.3-9 Powerhouse, Units 1 and 2 - Plan El. 713.0 and 708.0 Figure 12.3-10 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 708.0 and 713.0 - Radiation Zone Map Figure 12.3-11 Powerhouse, Units 1 and 2 - Plan El. 685.5 Figure 12.3-12 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 676.0 and 692.0 - Radiation Zone Map Figure 12.3-13 Powerhouse, Units 1 and 2 - Turbine Building - Plan El. 670.0 Radiation Zone Map Figure 12.3-14 Powerhouse, Units 1 and 2 - Reactor Building - Radiation Zone Map Figure 12.3-15 Powerhouse, Units 1 and 2 - Reactor Building - Radiation Zone Map Figure 12.3-16 Powerhouse, Auxiliary Building, Mechanical, Heating, Ventilating and Air Conditioning Figure 12.3-17 Fuel Transfer Tube Shielding Details Figure 12.3-18 Powerhouse Service Building Figure 12.3-19 Service Building - Miscellaneous Details, Hot Shop Facilities Figure 12.4-1 Dose Points Inside Restricted Area List of Figures 12-v

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WATTS BAR WBNP-92 12.0 RADIATION PROTECTION 12.1 Assuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA) 12.1.1 Policy Considerations TVA has established a formal program to ensure that occupational radiation exposures to employees are kept as low as reasonably achievable (ALARA). The program consists of: (1) full management commitment to the overall objectives of ALARA; (2) issuance of specific administrative documents and procedures to the TVA design and operating groups that emphasize the importance of ALARA through the design, testing, startup, operation, maintenance and decommissioning phases of TVA nuclear plants; and (3) continued appraisal of inplant radiation and contamination conditions by the onsite radiation protection staff.

12.1.2 Design Considerations The facility and equipment design features for control of occupational radiation exposures are described in detail in Section 12.3. Although the original design of Watts Bar Nuclear Plant predated Regulatory Guide 8.8, the concept of keeping occupational exposures ALARA is an important consideration throughout new designs and modifications of the plant. In addition, the plant design effort routinely considers radiation protection experience at other nuclear plants.

New designs and modifications of the plant are performed and reviewed by engineers and health physicists with several years of experience in radiation protection design.

In addition, the design of the plant is continually reviewed and modified as necessary when new ALARA concerns become known. Close communication among the design staff, equipment vendors, operating and maintenance personnel, and Radiological Control Personnel is maintained in order to design Watts Bar Nuclear Plant and its equipment with ALARA considerations as a primary concern.

Dose assessment based on operating experience is discussed in Section 12.4.

In general, piping which may contain significant concentrations of radioactive materials is not field-run. Some sample and radiation monitoring lines are field-run. While the exact location is set in the field, the general location is determined by the designer to minimize radiation exposure.

12.1.3 ALARA Operational Considerations Consistent with TVA's overall commitment to keep occupational radiation exposures as low as reasonably achievable, specific plans and procedures are followed by operating and maintenance staff to assure that ALARA goals are achieved in the operation of the plant. Operational ALARA policy and procedures are formulated at the Corporate level in Nuclear Power and are implemented at each nuclear plant through the issuance of division procedures and plant instructions for the purpose of maintaining Total Effective Dose Equivalent (TEDE) ALARA. These procedures and instructions are consistent with the intent of Section C.1 of Regulatory Guide 8.8 and Assuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA) 12.1-1

WATTS BAR WBNP-92 Regulatory Guide 8.10. Included in these operating procedures and plant instructions are the provision that employee radiation exposure trends are reviewed periodically by management staff at the plant and in the central office. Summary reports are prepared that describe: (a) major problem areas where high radiation exposures are encountered; (b) which worker group is accumulating the highest exposures; and (c) recommendations for changes in operating, maintenance, and inspection procedures or modifications to the plant as appropriate to reduce exposures.

Maintenance activities that could involve significant radiation exposure of employees are carefully planned and carried out using well-trained personnel and proper equipment. Where applicable, specific radiation exposure reduction techniques, such as those set out in Section C.3 of Regulatory Guide 8.8, are used. Careful personnel radiation and contamination monitoring are integral parts of such maintenance activities. Upon completion of major maintenance jobs, personnel radiation exposures are evaluated and assessed relative to predicted man-rem exposures so that appropriate changes can be made in techniques or procedures for future jobs.

Additionally at the plant level, the Plant Operations Review Committee reviews operating and maintenance activities involving the major systems of the plant (i.e.,

radwaste, NSSS, etc.) to further assure that occupational exposures are kept as low as reasonably achievable.

An ALARA committee composed primarily of supervisory personnel is established to review periodically the effectiveness of implementation of the ALARA Program.

Reviews include the site performance against ALARA goals, employee ALARA suggestions, ALARA planning docuemnts, and trends. The Plant Manager or Assistant Plant Manager will normally serve as chairman of the site ALARA committee.

REFERENCES None 12.1-2 Assuring that Occupational Radiation Exposures Are as Low as Reasonably Achievable (ALARA)

WATTS BAR WBNP-84 12.2 RADIATION SOURCES 12.2.1 Contained Sources The source terms employed in the design of plant shielding are presented in this section. Source terms are presented in accordance with the following organization:

(1) Primary system sources (2) Auxiliary systems sources (3) Sources during refueling (4) Maximum hypothetical accident (MHA) sources Systems for which source terms are presented in this section are essentially the same in the Watts Bar plant as they are in the Sequoyah plant. Therefore, the shielding design source terms developed for the Sequoyah FSAR are applicable and are incorporated into the Watts Bar FSAR. For completeness, the bases for development of these source terms are given in the following sections.

12.2.1.1 Primary System Sources 12.2.1.1.1 Sources Shielded by Primary Shield Concrete The major sources shielded by the primary concrete shield are the neutron and gamma sources inside the reactor pressure vessel, the gamma sources in the pressure vessel wall itself, and the coolant activities inside those parts of the coolant pipes from the reactor vessel to approximately the outer edges of the weld inspection openings.

Core center plane neutron fluxes across the primary shield concrete are shown on Figure 12.2-1. Fluxes across the primary concrete for distances (H) above and below the core center plane are determined by using the following axial peaking factors (Fp):

H (ft.) 0 1 2 3 4 5 6 Fp 1.0 1.0 0.99 0.93 0.76 0.51 0.20 The neutron fluxes along the vessel z axis (vertically) are essentially zero at the top and bottom outer surfaces of the reactor vessel.

Principal gamma sources during power operation are the prompt fission gammas, the delayed fission product gammas, prompt capture gammas, and the activation (Co-60, Fe-59, etc.) gammas. Core center plane gamma fluxes across the primary shield concrete during full power operation are shown on Figure 12.2-2. Fluxes across the primary concrete for other axial distances are found by using the same axial peaking factors as given above for neutron fluxes. The principal gamma sources during shutdown are the fission product delayed gammas and the activation gammas.

RADIATION SOURCES 12.2-1

WATTS BAR WBNP-84 The principal gamma sources in the reactor coolant are the fission product, reactor coolant activation product, and noncoolant activation (corrosion) product activities.

The fission product and noncoolant activation product activities are given in Table 11.1-2. The only coolant activation product significant for shielding purposes is N-16.

Values of the N-16 activity are shown in Figure 12.2-3.

12.2.1.1.2 Sources Shielded by Secondary Shield Concrete The secondary concrete shielding provides shielding for that portion of the primary coolant system which lies outside the primary concrete shield. Major elements are the coolant recirculation pumps, the steam generators, the pressurizer and all connecting piping. In addition, the secondary shielding attenuates radiation that penetrates the primary shield concrete. Activities in the primary coolant system have been identified above. Specified activities in the vapor and liquid spaces of the pressurizer are given in Table 11.1-4.

12.2.1.2 Auxiliary Systems Sources The auxiliary systems equipment for which a determination of shielding sources is required consists of pumps, heat exchangers, demineralizers, filters, units of evaporator packages, liquid tanks, gas tanks, and pipes. Inventories (or specific activity and volume or mass from which inventory can be obtained) used in the determination of required shielding for auxiliary systems components are given in Tables 12.2-1 through 12.2-11. Development of these source terms is described in the following subsections.

12.2.1.2.1 Chemical and Volume Control System - Mixed-Bed Demineralizers The reactor coolant with the specific activities (Ci/gm) given in Table 11.1-2 is assumed to flow at 75 gal/min (284,000 gm/min) through a mixed-bed demineralizer for 365 days. Consistent with the basis for generating the specific activities in Table 11.1-1, a mixed-bed removal efficiency of 0.09 for cesium, molybdenum, and yttrium and 0.90 for all other nongaseous elements is assumed.

In the case of the isotopes with half-lives that are not long in comparison with the rate at which reactor coolant is purified in the chemical and volume control system, demineralizer inventories can be higher than those obtained using these assumptions.

This effect is taken into account by increasing the inventories of the following isotopes by the factor 1.11: Mn-56, Br-84, Rb-89, Sr-91, Y-91m, Y-91, Te-132, Te-134, I-132, I-133, I-134, and I-135. Similarly, the inventories of Mo-99, Tc-99m, and Cs-138 are increased by the factor 11.1. Source term inventories are given in Table 12.2-1.

12.2.1.2.2 Chemical and Volume Control System - Cation Bed Demineralizer Reactor coolant is assumed to flow at an average 7.5 gal/min (28,400 gm/min) through the cation bed demineralizer for 365 days. A cation removal efficiency of 90% for cesium, molybdenum and yttrium is assumed. (For this calculation a removal efficiency of 0.0 for these three elements is assumed for the upstream mixed-bed demineralizer.) In the case of isotopes with half-lives that are not long in comparison with the rate at which reactor coolant is processed through the cation demineralizer, 12.2-2 RADIATION SOURCES

WATTS BAR WBNP-84 demineralizer inventories can be higher than those obtained using these assumptions.

This effect is taken into account by increasing, by the factor 1.11, the inventories of the following isotopes: Mo-99, Tc-99m, Cs-136, Cs-138, and Y-90.

Inventories of other cations on the cation bed demineralizer are not included in Table 12.2-2. In general, the inventories of these other cation isotopes will be a maximum of about 1.0% of the inventories shown in Table 12.2-2, for the mixed-bed demineralizer.

These maximum values, as translated into gamma source strengths, are small in comparison with the cation bed shielding source term represented by the gamma radiation from the molybdenum, yttrium, and cesium isotopes.

12.2.1.2.3 Chemical and Volume Control System - Volume Control Tanks Isotopic inventories in the vapor and liquid spaces of the volume control tank are given in Table 11.1-3.

12.2.1.2.4 Chemical and Volume Control System - Reactor Coolant Filter The source for which reactor coolant filter shielding is designed consists of activity from the upstream demineralizers. For shielding calculations, 1.0% of the mixed-bed demineralizer inventory as given in Table 12.2-1 is distributed uniformly throughout the reactor coolant filter cartridge. In the generation of the mixed-bed demineralizer inventory, it was assumed that the mixed-bed demineralizer removes all nongaseous activity including cesium, yttrium, and molybdenum. Thus, the mixed-bed demineralizer inventory as calculated contains the maximum inventory that could be collected on the mixed-bed demineralizers and cation demineralizer combined. It follows that the reactor coolant filter inventory determined with the stated prescription includes 1.0% of this maximum combined inventory. The filter cartridge volume over which the isotopic inventories are uniformly distributed is 1.17 cubic feet (diameter equals 8-15/16 inches, length equals 32-5/16 inches).

12.2.1.2.5 Chemical and Volume Control System - Seal Water Return Filter The only significant radioactive source on the seal water return filter is accumulated during operation with the excess letdown heat exchanger. The excess letdown heat exchanger, can be employed when normal letdown is temporarily out of service or it can be used to supplement maximum letdown during final states of heatup. The seal water return filter will thus collect activity during relatively short and infrequent time periods. A maximum inventory on the filter is generated by allowing undemineralized and unfiltered reactor coolant to flow continuously for 30 days through the filter at the design flow rate of the excess letdown heat exchanger. The design flow rate is 24.7 gal/min or 12,380 lb/hr. The filter is assumed to remove all corrosion product activities and 0.4% of the cesium activity in the reactor coolant. The maximum inventory on a filter is given in Table 12.2-3.

12.2.1.2.6 Chemical and Volume Control System - Seal Water Injection Filters The only significant radioactive source on the seal water injection filters will consist of activity of particulates that are passed by the seal water return filter (98% retention of 25 micron size) but are collected by the online injection filter (98% retention of 5 micron RADIATION SOURCES 12.2-3

WATTS BAR WBNP-84 size). The maximum inventory on each filter assumed for shielding calculations is taken to be the same as that given in Table 12.2-3 for the seal water return filter. The inventory is distributed over a filter volume of 1.25 cubic foot (diameter equals 8.625 inches, length equals 37.0 inches).

12.2.1.2.7 Chemical and Volume Control System - Holdup Tanks Holdup tank inventory development begins with the assumption that reactor coolant is discharged from both units simultaneously into a single holdup tank. Flow into the tank is 120 gal/min/unit. Flow from one unit is assumed to be undemineralized and unfiltered. For the flow from the other unit a decontamination factor (D.F.) of 10 across a mixed-bed demineralizer is taken for all nongaseous isotopes except those of cesium, yttrium and molybdenum. Filling continues until liquid occupies one-fourth of the tank volume. The liquid volume is then 32,000 gallons and the vapor space volume is 96,000 gallons. Under equilibrium conditions, 68% of the gases will be retained in the liquid. It is assumed that the other 32% is retained in the tank vapor space although physically, some of the vaporized gas would have been forced out of the tank to the vent header. Particulate daughters of noble gases entering the tank are retained in the tank liquid space. Inventories in the tank liquid and vapor spaces are given in Table 12.2-4 and 12.2-5, respectively.

12.2.1.2.8 Chemical and Volume Control System - Evaporator Feed Mixed-Bed Ion Exchangers A constant flow at 30 gpm (boric acid-evaporator capacity) through an evaporator feed mixed-bed ion exchanger for 60 days is assumed. This procedure is a conservative mock-up of actual ion exchanger operation for shorter time periods during a year of plant operation. It is further assumed that all flow has been processed through one of the chemical and volume control system mixed-bed demineralizers and that 10% of the volume has been processed through a chemical and volume control system cation demineralizer. Thus, a minimum of 9.0% of the cesium, yttrium, and molybdenum isotopic activities and 90% of the other nongaseous activities have been removed from the coolant prior to its processing through the evaporator feed mixed-bed ion exchanger. The evaporator feed ion exchanger is assumed to remove all of the remaining nongaseous activity. Since coolant activities, as given in Table 11.1-2, do not take into account isotopic removal by the evaporator feed ion exchangers, the ion exchange inventories developed with this method are too high. Corrections to the inventories of isotopes with long half-lives are readily obtained. In the case of the Cs-134, Cs-137, and Y-91 (excluding that part of the Y-91 inventory that results from decay of Sr-91 on the ion exchanger), the inventory is reduced by the factor, (0.09 x 75 gpm)/(0.09 x 75 gpm + 0.91 x 30 gpm) = 0.198. The similarly obtained correction factor for other long-lived activity, (0.90 x 75 gpm)/(0.90 x 75 gpm + 0.10 x 30 gpm) =

0.96 is not applied. The derived shielding source team inventories are given in Table 12.2-6.

The CVCS boric acid evaporator package is not required for the operation of Unit 1; however, the package is installed and connected to the waste disposal system. Liquid waste will be processed through the mobile demineralizer until such time that the boric acid evaporator becomes an available option for processing liquid radioactive waste.

12.2-4 RADIATION SOURCES

WATTS BAR WBNP-84 12.2.1.2.9 Chemical and Volume Control System - Evaporator Feed Cation Bed Ion Exchanger The shielding source term inventory on the evaporator feed cation bed ion exchanger is taken to be the same as the cesium, molybdenum, and yttrium inventories on the evaporator feed mixed bed ion exchanger except that yttrium activities on the mixed-bed that result from decay of strontium on the mixed-bed are excluded. The inventories of other cations on the cation demineralizer are not included since they would be very small in comparison with the inventories obtained from cesium, yttrium and molybdenum. (See the note in Section 12.2.1.2.8 relative to the Unit 1 operation without the boric acid evaporator package.)

The derived shielding source term inventories are given in Table 12.2.7.

12.2.1.2.10 Chemical and Volume Control System - Ion Exchange Filters The maximum source on either of the ion exchange filters is assumed to be 1.0% of the maximum inventory that could exist on an evaporator cation bed ion exchanger.

The cation exchanger is downstream of the mixed-bed ion exchanger and thus any detached resin fines from the mixed-bed ion exchanger would become lost in the cation bed resin. For shielding calculations the filter inventory is uniformly distributed over a filter cartridge volume of 0.44 cubic feet (diameter equals 7 inches, length equals 19-7/8 inches). (See the note in Section 12.2.1.2.8 relative to the Unit 1 operation without the boric acid evaporator package.)

12.2.1.2.11 Gas Stripper and Boric Acid Evaporator Package The sources associated with the recycle evaporator are specified in this section. The gaseous activity is concentrated in the vent condenser portion, while particulate activity is concentrated in the evaporator section.

The maximum gaseous activity concentrations occur when the unit is processing reactor coolant at the maximum rate. No credit is taken for radioactive decay before processing in the gas stripper. Gas concentration in the reactor coolant is assumed to be 35 standard cubic centimeters per kilogram. The activity concentrations in the vent condenser in microcuries per cubic centimeter are thus obtained by increasing reactor coolant concentrations, in microcuries per gram, by the factor, 1000 gm/35 cc = 28.6 gm/cc. Concentrations are given in Table 12.2.8 (these concentrations are not affected by considerations of hydrogen removal via venting from the holdup tank if the fractional release of nitrogen and hydrogen from the coolant is assumed to be the same as that of the radioactive gases). (See the note in Section 12.2.1.2.8 relative to the Unit 1 operation without the boric acid evaporator package.)

For shielding calculations, the vent condenser is considered to be a cylindrical source with a diameter of 8 inches and a length of 20 inches. Although the composition is approximately 30% stainless steel, 22% water and 48% vapor (volume percents) the shielding calculations do not take credit for any attenuation in the steel.

To generate the shielding source inventories in the evaporator bottoms, it is assumed that reactor coolant is processed through the evaporator at 30 gpm (1.14 x 105 RADIATION SOURCES 12.2-5

WATTS BAR WBNP-84 gm/min). It is assumed that only 0.10 of all nongaseous isotopes are removed in the evaporator feed mixed-bed and cation bed ion exchangers. Processing continues until isotopic concentrations in the evaporator bottoms is about 40 Ci/gm of liquid. The inventory is given in Table 12.2-9. This inventory is obtained in 417 minutes of processing. At this point the activity in the liquid has been concentrated by about a factor of 23. (See the note in Section 12.2.1.2.8 relative to the Unit 1 operation without the boric acid evaporator package.)

For shielding calculations, the source geometry is considered to be a cylinder with a diameter of 3.5 feet, and a length of 9.9 feet. The composition in volume percents is 10% stainless steel, 77% water, and 13% air. The source inventory is uniformly distributed throughout the 712 gallons homogenized volume.

12.2.1.2.12 Spent Fuel Pool Cooling System Demineralizer and Filters The principal isotopes that will be collected on the demineralizer and filters of the spent fuel pool cooling system are I-131, Cs-134, Cs-137, and the corrosion product activities, Mn-54, Co-58, Co-60, Fe-59 and Cr-51. In the development of source terms for these components, fission product release from the fuel rods subsequent to the initiation of shutdown is neglected. Minimum time between beginning of cooldown and filling of the refueling canal is at least three days. Therefore, continuous purification at 75 gal/min (284,000 gm/min) through the chemical and volume control system mixed-bed and cation bed demineralizers is assumed to continue for three days before the reactor vessel head is removed and primary coolant mixes with water in the refueling canal (flow through the mixed-bed demineralizer could be 120 gpm but the design flow rate through the cation bed demineralizer is only 75 gpm). A D.F. of 10 between a point upstream of the mixed-bed demineralizer and a point downstream of cation bed demineralizer is taken for the iodine and cesium isotopes. Credit for radioactive decay during the three days of purification is taken for I-131.

Although the demineralizer D.F. of 10 is also applicable to the corrosion product activities, removal of corrosion product activities in the demineralizers could be balanced by further releases into the coolant. In fact, the evidence suggests that corrosion product activity levels in the coolant could actually increase during such temperature changes. Also, large increases in Co-58 activity have been observed after removal of the reactor vessel head and dilution of the coolant of the refueling canal[3]. This effect has been taken into account by increasing the corrosion product inventory in the reactor coolant at beginning of shutdown by a factor of 5. The radioactive inventories available for demineralization with the spent fuel pool demineralizer and for filtration with the refueling water purification filter are thus determined as follows:

(1) I-131 inventory is decreased exponentially with the removal constant, 1.60 day-1, (2) Cs-134 and Cs-137 inventories are decreased exponentially with the removal constant, 1.51 day-1, and 12.2-6 RADIATION SOURCES

WATTS BAR WBNP-84 (3) Corrosion product inventories are increased by a factor of 5. No credit is taken for radioactive decay during accumulation on the demineralizer.

Demineralizer inventories are given in Table 12.2-10.

The inventory on a refueling water purification filter will be accumulated in one of two ways. If water entering the filter has first passed through the spent fuel pool demineralizer, the filter source would consist of activity on resin dislodged from the demineralizer. One percent of the demineralizer inventories would be used as a filter source. If water is purified through the refueling water purification filter only, the filter inventory would be identical to the spent fuel pool demineralizer inventory in the case of the corrosion product activities.

Iodine would not be collected in significant quantity. As in the case of the seal water return and seal water injection filters, a filtration efficiency of 0.4% would be assumed for cesium. This latter mode of operation results in the worst case inventory for shielding calculations and thus is the basis for the refueling water purification filter inventories given in Table 12.2-11.

No large quantities of radioactivity are expected to be collected on the spent fuel pool skimmer filter and the spent fuel pool filter. These filters filter the water in the transfer canal and in the spent fuel pool only. Negligible activity from the refueling canal inside the containment will enter the transfer canal through the fuel transfer tube. Identifiable sources of spent fuel pool and skimmer filter activity are the following:

(1) Particulate corrosion and fission product activities that become dislodged from fuel assembly surfaces primarily during movement of a fuel assembly.

(2) Activity from fuel assemblies with defective cladding.

(3) In the case of the spent fuel pool filter only, activity on resin fines dislodged from the upstream spent fuel pool demineralizer. An upper limit for this source is considered to be 1.0% of the demineralizer activity inventory.

(4) Activity introduced from the refueling water storage tanks when the spent fuel pool is periodically emptied and refilled.

Of these, the largest potential contribution to filter activity is expected to be the corrosion and filter product activities dislodged from the fuel assembly. An extreme upper limit for this source should be the inventory assigned to a refueling water purification filter. Use of this inventory as the source term for the spent fuel pool filter and the spent fuel pool skimmer filter more than adequately takes into account activity contributions from the other three activity origins identified above. Therefore, the inventory on a refueling water purification filter, given in Table 12.2-11, is also used as the source term in shielding calculations for the spent fuel pool filter and the spent fuel pool skimmer filter.

RADIATION SOURCES 12.2-7

WATTS BAR WBNP-84 12.2.1.3 Sources During Refueling The principal radioactive source in the proximity of the fuel assembly transfer path is the spent fuel assembly fission product inventory discussed below. Activity in the refueling canal water will normally be reduced to 0.01 microcurie per cubic centimeter before fuel transfer begins.

The maximum fuel assembly fission product concentrations are calculated by TID-14844 methodology[5], and they are included in Table 12.2-12. The total power distribution peaking factor is 2.40. The axial peaking factor is 1.37, and the radial peaking factor is 1.75. The average inventory in a fuel assembly is 1/2.40 of the inventory in Table 12.2-12. Since there are 193 fuel assemblies in the core, the total core inventory is 193 times the average inventory in a fuel assembly.

Other sources encountered during refueling operations are the irradiated in-core instrumentation detectors and the control rods. When removed from the vessel, the irradiated in-core instrumentation detectors and the irradiated Ag-In-Cd control rods are also important sources. During refueling, the irradiated in-core instrumentation detectors are stored in the crane wall. The irradiated in-core detector drive wire sources are given in Table 12.2-13. The shielding provided for spent fuel assemblies also shields spent irradiated control rod sources either in place in an assembly or when being transferred from one assembly to another.

The absorber materials used in the hybrid control rods are boron carbide (B4C) and silver-indium-cadmium (Ag-In-Cd). The gamma ray source strengths associated with the Ag-In-Cd absorber are listed in Table 12.2-14 for various times after shutdown.

The values are per cubic centimeter of absorber for an irradiation period of 4 years.

There are no significant gamma ray sources associated with the B4C absorber.

The material used for the control rod cladding, secondary source rod cladding, and burnable absorber rod cladding and inner sheath is type-304 stainless steel with a maximum cobalt content of 0.12 weight percent. The gamma ray source strengths associated with the stainless steel are also listed in Table 12.2-14 for various times after shutdown. The values are per cubic centimeter of stainless steel for an irradiation time of 15 years.

12.2.1.4 Maximum Hypothetical Accident (MHA) Sources For MHA (maximum hypothetical accident) exposure calculations, fractions of the equilibrium core fission product inventory (see Table 12.2-12) that are assumed to be released from the fuel rods are the following:

(1) 100% of the noble gases released to containment (2) 50% of the core iodines released to the sump and 50% of pure iodines released to containment (3) 1% of core particulates released to the sump Refer to Chapter 15 for details of accident analysis.

12.2-8 RADIATION SOURCES

WATTS BAR WBNP-84 12.2.1.5 Condensate Demineralizer Waste Evaporator The Condensate Demineralizer Waste Evaporator (CDWE) package is not required for the operation of Unit 1, however the package is installed and connected to the waste disposal system. System description, function, and interfaces are contained in Section 11.2.2. Liquid waste will be processed through the mobile demineralizer until such time that the CDWE becomes an available option for processing liquid radioactive wastes.

12.2.2 Airborne Radioactive Material Sources Expected airborne activity levels and the corresponding derived air concentration (DAC) fractions, for the Containment, Turbine Building, and Auxiliary Building and the instrument room are presented in Tables 12.2-19 through 12.2-22, respectively. The DAC fractions are determined by dividing the estimated radioactive air concentrations by the corresponding DAC values. For the determination of these concentrations, the assumptions and models are those employed in Section 11.3.7 for the determination of gaseous effluent releases except for the following assumptions: (1) lower containment, upper containment, and the instrument room are assumed to be purged after each 60 days of operation, (2) purge flows are 14,958 cfm (upper containment),

7,500 cfm (lower containment), 540 cfm (instrument room), and (3) purge lasts for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

REFERENCES (1) Deleted by Amendment 84.

(2) Deleted by Amendment 84.

(3) "Coolant Activity Experiences at Connecticut Yankee," by R. H. Graves, Nuclear News, November 1970, page 66. (See 12.2.1.2.12)

(4) ANSI/ANS-18.1-1984, "Radioactive Source Term for Normal Operation of Light Water Reactors," December 31, 1984.

(5) TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.

RADIATION SOURCES 12.2-9

WATTS BAR WBNP-84 Table 12.2-1 Chemical And Volume Control System Mixed Bed Demineralizer Inventory Inventory Isotope (Microcuries) Isotope (Microcuries)

Cr 51 .137(+8) I 131 .107(+11)

Mn 54 .700(+8) I 132 .543(+9)

Mn 56 .184(+7) I 133 .204(+10)

Co 58 .917(+9) I 134 .126(+8)

Co 60 .932(+8) I 135 .365(+9)

Fe 59 .239(+8) Ba 137m .124(+11)

Br 84 .548(+6) Ba 140 .286(+8)

Br 88 .270(+8) La 140 .299(+8)

Rb 89 .631(+6) Ce 144 .241(+8)

Sr 89 .102(+9) Pr 144 .241(+8)

Sr 90 .146(+8) Mo 99 .865(+10)

Sr 91 .454(+6) Tc 99m .758(+10)

Y 90 .145(+8) Cs 134 .240(+10)

Y 91m .277(+6) Cs 136 .968(+8)

Y 91 .172(+8) Cs 137 .133(+11)

Zr 95 .228(+8) Cs 138 .125(+8)

Nb 95m .455(+6)

Nb 95 .341(+8)

Te 132 .498(+9)

Te 134 .522(+6)

a. Inventory is distributed over a resin volume of 30 cubic feet.

b. Numbers in parentheses denote exponents to the base 10.

12.2-10 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-2 Chemical And Volume Control System Cation Bed Demineralizer Inventory Isotope (Microcuries)

Mo 99 .865(+9)

Tc 99m .758(+9)

Cs 134 .240(+10)

Cs 136 .107(+9)

Cs 137 .133(+11)

Cs 138 .125(+7)

Ba 137m .124(+11)

Y 90 .204(+5)

Y 91 .168(+8)

a. Inventory is distributed over a resin volume of 20 cubic feet.

b. Numbers in parentheses denote exponents to the base 10.

RADIATION SOURCES 12.2-11

WATTS BAR WBNP-84 Table 12.2-3 Chemical and Volume Control System Seal Water Return Filter Inventory Isotope (Microcuries)

Cr 51 .265(+7)

Mn 54 .301(+7)

Mn 56 .607(+6)

Co 58 .877(+8)

Co 60 .298(+7)

Fe 59 .324(+7)

Cs 134 .335(+7)

Cs 136 .113(+7)

Cs 137 .162(+8)

Cs 138 .165(+5)

Ba 137m .151(+8)

a. Inventory is distributed over a filter cartridge volume of 3.72 ft3, (Diameter = 13.75", Length 43.25)

b. Numbers in parentheses denote exponents to the base 10.

12.2-12 RADIATION SOURCES

WATTS BAR WBNP-0 Table 12.2-4 Chemical And Volume Control System-Hcldup Tank Liquid Space Inventory Inventory Isotope (Microcuries) Isotope (Microcuries)

Cr 51 .620(+5) Te 132 .172(+8)

Mn 54 .514(+5) Te 134 .808(+6)

Mn 56 .146(+7) Cs 134 .255(+8)

Co 58 .167(+7) Cs 136 .170(+8)

Co 60 .494(+5) Cs 137 .121(+9)

Fe 59 .667(+5) Cs 138 .492(+8)

Br 84 .912(+6) Ba 137m .110(+9)

Br 88 .337(+9) Ba 140 .280(+6)

Rb 89 .111(+7) La 140 .104(+6)

Sr 89 .255(+6) Ce 144 .180(+5)

Sr 90 .734(+u) Pr 144 .180(+5)

Sr 91 .117(+6) Kr 85m .153(+9)

Zr 95 .447(+5) Kr 85 .338(+9)

Nb 95 .427(+5) Kr 87 .573(+8)

I 131 .166(+9) Kr 88 .235(+9)

I 132 .484(+9) Xe 131m .156(+9)

I 133 .257(+9) Xe 133m .260(+9)

I 134 .181(+8) Xe 133 .236(+11)

I 135 .131(+9) Xe 135m .921(+7)

Xe 135 .486(+9)

Xe 138 .849(+7)

a. Inventory is distributed over a volume, of 32,000 gallons (the bottom 1/4 of the tank volume).

b. Numbers in parentheses denote exponents to the base 1.0.

RADIATION SOURCES 12.2-13

WATTS BAR WBNP-0 Table 12.2-5 Chemical And Volume Control System-Hcidup Tank Vapor Space Inventory Isotope (Microcuries)

Kr 85m .721(+8)

Kr 85 .183(+9)

Kr 87 .270(+8)

Kr 88 .111(+9)

Xe131m .736(+8)

Xe 133m .123(+9)

Xe 133 .112(+11)

Xe 135m .433(+7)

Xe 135 .229(+9)

Xe 138 .400(+7) a.Inventory is distributed over a volume of 96,000 gallons (the top 3/4 of the tank volume).

b.Numbers in parentheses denote exponents to the base 10.

12.2-14 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-6 Chemical And Volume Control System - Evaporator Feed Mixed Bed Ion Exchanger Inventory Inventory Isotope (Microcuries) Isotope (Microcuries)

Cr 51 .475(+6) Te 132 .200(+8)

Mn 54 .707(+6) Te 134 .209(+5)

Mn 56 .738(+5) I 131 .474(+9)

Co 58 .186(+8) I 132 .220(+8)

Co 60 .720(+6) I 133 .821(+8)

Fe 59 .643(+6) I 134 .505(+6)

Br 84 .219(+5) I 135 .147(+8)

Br 88 .108(+7) Ba 137m .165(+10)

Rb 89 .253(+5) Ba 140 .122(+7)

Rb 89 .256(+7) La 140 .127(+7)

Sr 90 .108(+6) Ce 144 .247(+6)

Sr 91 .182(+5) Pr 144 .247(+6)

Y 90 .180(+6) Mo 99 .316(+10)

Y 91M .111(+5) Tc 99m .277(+10)

Y 91 .697(+7) Cs 134 .362(+9)

Zr 95 .488(+6) Cs 136 .376(+9)

Nb 95m .908(+4) Cs 137 .177(+l0)

Nb 95 .586(+6) Cs 138 .457(+7)

a. Inventory is distributed over a resin volume of 27 ft3.

b. Numbers in parentheses denote exponents to the base 10.

RADIATION SOURCES 12.2-15

WATTS BAR WBNP-84 Table 12.2-7 Chemical And Volume Control System-Evaporator Feed Cation Bed Ion Exchanger Inventory Isotope (Microcuries)

Mo 99 .316(+10)

Tc 99m .362(+9)

Cs 134 .376(+9)

Cs 136 .177(+10)

Cs 138 .457(+7)

Ba 137m .165(+10)

Y 90 .746(+5)

Y 91 .697(+7) a.Inventory is distributed over a resin volume of 20 ft3.

b.Numbers in parentheses denote exponents to the base 10.

12.2-16 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-8 Gas Stripper And Boric Acid Evaporator Package-Vent Condenser Concentration Isotope (Microcuries/Cubic Centimeter)

Kr 85m 63.0 Kr 85 134 Kr 87 34.3 Kr 88 106 Xe 131m 54.3 Xe 133m 91.5 Xe 133 8230 Xe 135m 5.43 Xe 135 180 Xe 138 19.0

a. Activity concentration exists in 48% of the total condenser volume. Shielding calculations are performed by assuring concentrations in the entire condenser volume, 1.648 x 104 cm3, are 48%

of the concentrations in this table.

RADIATION SOURCES 12.2-17

WATTS BAR WBNP-0 Table 12.2-9 Gas Stripper And Boric Acid Evaporator Package-Evaporator Bottoms Inventory Inventory Isotope (Microcuries) Isotope (Microcuries)

Cr 51 .439(+4) Zr 95 .317(+4)

Mn 54 .365(+4) Nb 95m .167(+1)

Mn 56 .623(+5) Nb 95 .303(+4)

Co 58 .118(+6) Mo 99 .242(+8)

Co 60 .351(+4) Tc 99m .671(+7)

Fe 59 .473(+4) I 131 .117(+8)

Br 84 .219(+5) I 132 .248(+7)

Br 88 .108(+7) I 133 .169(+8)

Rb 89 .253(+5) I 134 .501(+6)

Sr 89 .181(+5) I 135 .746(+7)

Sr 90 .521(+3) Te 132 .120(+7)

Sr 91 .710(+4) Te 134 .209(+5)

Y 90 .613(+3) Ba 137m .439(+7)

Y 91m .369(+4) Ba 140 .198(+5)

Y 91 .260(+5) La 140 .784(+4)

Y 92 .189(+4) Ce 144 .128(+4)

Pr 144 .128(+4)

Cs 134 .995(+6)

Cs 136 .659(+6)

Cs 137 .474(+7)

Cs 138 .502(+6) a.Inventory is distributed over a homogenized volume of 712 gallons.

b.Numbers in parentheses denote exponents to the base 10.

12.2-18 RADIATION SOURCES

WATTS BAR WBNP-0 Table 12.2-10 Spent Fuel Pool Cooling System-Spent Fuel Pool Demineralizer Inventory Isotope (Microcuries)

I 131 .500(+7)

Cs 137 .261(+7)

Cs 134 .547(+6)

Mn 54 .940(+6)

Co 58 .305(+8)

Co 60 .905(+6)

Fe 59 .122(+7)

Cr 51 .114(+7) a.Inventory is distributed uniformly over a resin volume of 30 cubic feet.

b.Numbers in parentheses denote exponents to the base 10.

RADIATION SOURCES 12.2-19

WATTS BAR WBNP-0 Table 12.2-11 Spent Fuel Pool Cooling System-refueling Water Purification Filter Inventory Isotope (Microcuries)

Cs 137 .104(+5)

Cs 134 .219(+4)

Mn 54 .940(+6)

Co 58 .305(+8)

Co 60 .905(+6)

Fe 59 .122(+7)

Cr 51 .114(+7) a.Numbers in parentheses denote exponents to the base 10.

12.2-20 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-12 Core Inventory (Page 1 of 3)

Half-Life Fisson Inventory Isotope (Sec.) Yield (ci) 1 Kr 83m 6.6960E+03 5.3069E-03 1.64E+07 2 Kr 85m 1.6128E+04 1.3017E-02 3.95E+07 3 Kr 85 3.3862E+08 2.8825E-03 9.99E+05 4 Kr 87 4.5780E+03 2.5421E-02 7.59E+07 5 Kr 88 1.0080E+04 3.5840E-02 1.08E+08 6 Kr 89 1.9080E+02 4.6812E-02 1.40E-08 7 Kr 90 3.2320E+01 4.6891E-02 1.45E-08 8 Xe 131m 1.0282E+06 3.9694E-04 6.68E+05 9 Xe 133m 1.9440E+05 1.9140E-03 5.16E+06 10 Xe 133 4.5706E+05 6.7705E-02 2.03E+08 11 Xe 135m 9.3900E+02 1.0564E-02 5.46E+07 12 Xe 135 3.2940E+04 6.6334E-02 5.55E+07 13 Xe 137 2.2980E+02 6.1325E-02 1.89E+08 14 Xe 138 8.5020E+02 6.2836E-02 1.79E+08 15 Xe 139 4.0000E+01 5.1578E-02 1.59E+08 16 Xe 140 1.3600E+01 3.7182E-02 1.15E+08 17 I 130 4.4496E+04 2.4100E-06 7.43E+03 18 I 131 6.9638E+05 2.8352E-02 8.80E+07 19 I 132 8.2080E+03 4.2083E-02 1.34E+08 20 I 133 7.4880E+04 6.7653E-02 1.97E+08 21 I 134 3.1560E+03 7.6117E-02 2.31E+08 22 I 135 2.3756E+04 6.4065E-02 1.79E+08 23 I 136m 4.6000E+01 2.1095E-02 6.51E+07 24 Br 83 8.6400E+03 5.3069E-03 1.64E+07 25 Br 84m 3.6000E+02 1.9217E-04 5.93E+05 26 Br 84 1.9080E+03 9.6650E-03 2.98E+07 27 Br 85 1.8000E+02 1.2953E-02 3.99E+07 28 Br 87 5.5700E+01 2.2016E-02 6.79E+07 29 Cs 134 6.5070E+07 4.5000E-07 6.25E+02 30 Cs 135 7.2580E+13 6.6348E-02 1.10E+02 31 Cs 136 1.1215E+06 5.2710E-05 1.63E+05 32 Cs 137 9.4671E+08 6.2626E-02 7.78E+06 33 Cs 138 1.9320E+03 6.7178E-02 2.07E+08 34 Cs 139 5.6400E+02 6.4137E-02 1.98E+08 35 Cs 140 6.3700E+01 5.9022E-02 1.82E+08 36 Cs 141 2.4900E+01 4.4186E-02 1.36E+08 37 Rb 88 1.0680E+03 3.6243E-02 1.12E+08 38 Rb 89 9.1200E+02 4.8470E-02 1.49E+08 39 Rb 90m 2.5600E+02 1.0034E-02 3.09E+07 40 Rb 90 1.5400E+02 4.9249E-02 1.52E+08 41 Rb 91 5.8000E+01 5.7450E-02 1.77E+08 RADIATION SOURCES 12.2-21

WATTS BAR WBNP-84 Table 12.2-12 Core Inventory (Page 2 of 3)

Half-Life Fisson Inventory Isotope (Sec.) Yield (ci) 42 Se 84 1.9800E+02 9.4511E-0 32.91E+07 43 Sr 89 4.3632E+06 4.8501E-02 1.50E+08 44 Sr 90 8.9937E+08 5.9155E-02 7.73E+06 45 Sr 91 3.4200E+04 5.9163E-02 1.82E+08 46 Sr 92 9.7560E+03 5.9482E-02 1.83E+08 47 Sr 93 4.5600E+02 6.2663E-02 1.93E+08 48 Sr 94 7.8000E+01 6.0148E-02 1.85E+08 49 Y 90 2.3040E+05 5.9157E-02 1.82E+08 50 Y 91m 2.9826E+03 3.6685E-02 1.13E+08 51 Y 91 5.0553E+06 5.9171E-02 1.82E+08 52 Y 92 1.2744E+04 5.9560E-02 1.84E+08 53 Y 93 3.6360E+04 6.3667E-02 1.96E+08 54 Y 94 1.1460E+03 6.4079E-02 1.98E+08 55 Y 95 6.4200E+02 6.4304E-02 1.98E+08 56 Y 96 1.3800E+02 5.9745E-02 1.84E+08 57 Zr 95 5.5279E+06 6.4593E-02 l.99E+08 58 Zr 97 6.1200E+04 5.9446E-02 1.83E+08 59 Nb 95 3.0370E+06 6.4594E-02 l.99E+08 60 Nb 97m 6.0000E+01 5.5931E-02 1.72E+08 61 Nb 97 4.3260E+03 5.9603E-02 1.84E+08 62 Mo 99 2.3832E+05 6.1327E-02 1.89E+08 63 Tc 99m 2.1672E+04 5.3968E-02 1.66E+08 64 Tc 99 6.7210E+12 6.1327E-02 1.10E+03 65 Tc 101 7.5200E+02 5.0440E-02 1.56E+08 66 Ru 103 3.4214E+06 3.1351E-02 9.67E+07 67 Ru 105 1.5984E+04 9.8670E-03 3.04E+07 68 Ru 106 3.1882E+07 3.9171E-03 8.52E+06 69 Ru 107 2.5200E+02 1.7339E-03 5.35E+06 70 Rh 103m 3.4200E+03 3.1038E-02 9.57E+07 71 Rh 105m 4.5000E+01 2.6641E-03 8.22E+06 72 Rh 105 1.2730E+05 9.8670E-03 3.04E+07 73 Rh 106 3.0400E+01 3.9171E-03 1.21E+07 74 Rh 107 1.3020E+03 1.7340E-03 5.35E+06 75 Sn 130 2.2320E+02 9.0010E-03 2.78E+07 76 Sb 127 3.0931E+05 1.2844E-03 3.96E+06 77 Sb 129 1.5525E+04 6.3718E-03 1.96E+07 78 Sb 130m 3.7800E+02 1.1157E-02 3.44E+07 79 Sb 130 2.4000E+03 2.9877E-03 9.21E+06 80 Sb 133 1.6200E+02 2.2346E-02 6.89E+07 12.2-22 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-12 Core Inventory (Page 3 of 3)

Half-Life Fisson Inventory Isotope (Sec.) Yield (ci) 81 Te 125m 5.0110E+06 6.7480E-05 2.08E+05 82 Te 127m 9.4176E+06 2.2418E-04 6.80E+05 83 Te 127 3.3660E+04 1.2799E-03 3.95E+06 84 Te 129m 2.9030E+06 1.9080E-03 5.88E+06 85 Te 129 4.1760E+03 6.2156E-03 1.92E+07 86 Te 131m 1.0800E+05 3.5440E-03 1.09E+07 87 Te 131 1.5000E+03 2.5405E-02 7.83E+07 88 Te 132 2.8080E+05 4.1877E-02 1.29E+08 89 Te 133m 3.3240E+03 3.9298E-02 1.21E+08 90 Te 133 7.4700E+02 3.0283E-02 9.34E+07 91 Te 134 2.5080E+03 6.7648E-02 2.09E+08 92 Ba 137m 1.5312E+02 5.9248E-02 1.83E+08 93 Ba 139 5.0940E+03 6.4816E-02 2.00E+08 94 Ba 140 1.1059E+06 6.3164E-02 1.95E+08 95 Ba 141 1.0962E+03 5.8670E-02 1.81E+08 96 Ba 142 6.4200E+02 5.8292E-02 1.80E+08 97 La 140 1.4497E+05 6.3221E-02 1.95E+08 98 La 141 1.4148E+04 5.8868E-02 1.82E+08 99 La 142 5.5620E+03 5.9304E-02 1.83E+08 100 La 143 8.4000E+02 5.9369E-02 1.83E+08 101 Ce 141 2.8080E+06 5.8868E-02 1.82E+08 102 Ce 143 1.1880E+05 5.9687E-02 1.84E+08 103 Ce 144 2.4538E+07 5.4554E-02 1.34E+08 104 Ce 145 1.8000E+02 3.9396E-02 1.21E+08 105 Pr 143 1.1733E+06 5.9687E-02 1.84E+08 106 Pr 144 1.0368E+03 5.4555E-02 1.68E+08 107 Pr 145 2.1528E+04 3.9408E-02 1.22E+08 108 Np 239 2.0300E+05 0.0000E+00 0.00E+00 RADIATION SOURCES 12.2-23

WATTS BAR WBNP-0 Table 12.2-13 Irradiated In-Core Detector Drive Wire Sources (MEV/CM-SEC)

Photon Energy Time After Shutdown (Mev/Y 0 4 Hours 12 Hours 1 Day 1 Week 1 Mo 0.8 4.1 x 1010 1.1 x 1010 2.6 x 109 1.1 x 109 1.0 x 109 1.0 x 1.3 1.2 x 109 1.2 x 109 1.2 x 109 1.2 x l09 1.1 x 109 8.4 x 2.2 4.0 x 1010 1.3 x 1010 1.5 x 109 5.8 X 107 --- ---

12.2-24 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-14 Irradiated Ag-In-Cd Control Rod Sources Control Rod, Ag-In-Cd Tip Energy Group Source Strength at Time After Shutdown (Mev/cm3-sec)

(Mev/gamma) 1 Day 1 Week 1 Month 6 Months 1 Year 5 Years 8 8 8 8 7 0.20-0.40 2.3 x 10 2.3 x 10 2.2 x 10 1.4 x 10 8.5 x 10 1.5 x 106 0.40 - 0.90 1.1 x 1012 1.1 x 1012 1.0 x 1012 6.6 x 1011 4.0 x 1011 7.1 x 109 0.90 - 1.35 2.0 x 1011 1.9 x 1011 1.8 x 1011 1.2 x 1011 7.2 x 1010 1.3 x 109 1.35 - 1.80 3.7 x 1011 3.7 x 1011 3.4 x 1011 2.3 x 1011 1.4 x 1011 2.5 x 109 The absorber cross-sectional area is 0.589 square centimeters per rod.

The absorber material density is 10.17 grams per cubic centimeter.

Control Rod Cladding (Type 304 Stainless Steel)

Energy Group Source Strength at Time After Shutdown (Mev/cm3-sec)

(Mev/gamma) 1 Day 1 Week 1 Month 6 Months 1 Year 5 Years 9 9 9 7 5 0.20-0.40 7.1 x 10 6.1 x 10 3.4 x 10 8.3 x 10 9.9 x 10 0 0.40 - 0.90 3.1 x 1010 2.9 x 1010 2.6 x 1010 1.2 x 1010 6.4 x 109 2.3 x 108 0.90 - 1.35 1.4 x 1011 2.3 x 1011 2.3 x 1011 2.1 x 1011 2.0 x 1011 1.2 x 1011 1.35 - 1.80 1.9 x 108 1.8 x 108 1.4 x 108 3.3 x 107 5.4 x 106 0 The cladding cross-sectional area per rod is 0.136 square centimeter.

Maximum cobalt is 0.12 weight percent.

RADIATION SOURCES 12.2-25

WATTS BAR WBNP-84 12.2-26 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-15 Deleted By Amendment 84 RADIATION SOURCES 12.2-27

WATTS BAR WBNP-84 Table 12.2-16 Deleted By Amendment 84 12.2-28 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-17 Deleted By Amendment 84 RADIATION SOURCES 12.2-29

WATTS BAR WBNP-84 Table 12.2-18 Deleted By Amendment 84 12.2-30 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-19 Estimated Average Airborne Radioactivity Concentrations In The Containment Building (Page 1 of 3)

LOWER COMPARTMENT UPPER COMPARTMENT Concentration Concentration 6 Hour Purge 10CFR20 DAC 6 Hour Purge 10CFR20 DAC Isotope (Ci/cc) Fraction (Ci/cc) Fraction Kr-85m 2.98E-06 1.49E-01 2.01E-07 1.01E-02 Kr-85 7.43E-06 7.43E-02 1.62E-06 1.62E-02 Kr-87 2.23E-06 4.45E-01 1.21E-07 2.41E-02 Kr-88 4.93E-06 2.46E+00 3.13E-07 1.57E-01 Xe-131m 1.42E-05 3.56E-02 1.77E-06 4.43E-03 Xe-133m 1.41E-06 1.41E-02 1.18E-07 1.18E-03 Xe-133 5.17E-05 5.17E-01 5.08E-06 5.08E-02 Xe-135m 9.09E-07 1.01E-01 2.42E-08 2.69E-03 Xe-135 1.67E-05 1.67E+00 1.20E-06 1.20E-01 Xe-137 7.85E-08 7.85E-01 7.03E-10 7.03E-03 Xe-138 7.89E-07 1.97E-01 1.97E-08 4.93E-03 Br-84 4.52E-11 2.26E-06 1.78E-12 8.92E-08 I-131 2.67E-10 1.34E-02 2.94E-11 1.47E-03 I-132 9.52E-10 3.17E-04 5.87E-11 1.96E-05 I-133 7.60E-10 7.60E-03 5.77E-11 5.77E-04 I-134 1.19E-09 5.94E-05 5.75E-11 2.87E-06 I-135 1.34E-09 1.91E-03 9.36E-11 1.34E-04 Rb-88 3.13E-06 1.04E-01 2.82E-07 9.38E-03 Cs-134 2.12E-11 5.31E-04 4.57E-12 1.14E-04 Cs-136 2.07E-12 6.88E-06 2.64E-13 8.82E-07 Cs-137 2.86E-11 4.77E-04 6.23E-12 1.04E-04 H-3 2.79E-06 1.40E-01 6.08E-07 3.04E-02 Na-24 9.77E-11 4.89E-05 7.24E-12 3.62E-06 Cr-51 8.05E-12 1.01E-06 1.30E-12 1.62E-07 Mn-54 4.77E-12 1.59E-05 1.01E-12 3.37E-06 Fe-55 3.63E-12 4.54E-06 7.84E-13 9.80E-07 Fe-59 8.17E-13 8.18E-06 1.46E-13 1.46E-06 Co-58 1.30E-11 4.33E-05 2.49E-12 8.29E-06 RADIATION SOURCES 12.2-31

WATTS BAR WBNP-84 Table 12.2-19 Estimated Average Airborne Radioactivity Concentrations In The Containment Building (Page 2 of 3)

LOWER COMPARTMENT UPPER COMPARTMENT Concentration Concentration 6 Hour Purge 10CFR20 DAC 6 Hour Purge 10CFR20 DAC Isotope (Ci/cc) Fraction (Ci/cc) Fraction Co-60 1.61E-12 1.61E-04 3.50E-13 3.50E-05 Zn-65 1.52E-12 1.52E-05 3.18E-13 3.18E-06 Sr-89 3.84E-13 6.41E-06 7.01E-14 1.17E-06 Sr-90 3.65E-14 1.82E-05 7.95E-15 3.98E-06 Sr-91 1.96E-12 1.96E-06 1.40E-13 1.40E-07 Y-90 3.65E-14 1.22E-07 7.95E-15 2.65E-08 Y-91m 1.05E-12 1.50E-08 7.97E-14 1.14E-09 Y-91 1.97E-14 3.94E-07 5.27E-15 1.05E-07 Y-93 8.58E-12 8.58E-06 6.19E-13 6.19E-07 Zr-95 1.09E-12 2.18E-05 2.06E-13 4.13E-06 Nb-95 8.66E-13 1.73E-06 1.93E-13 3.85E-07 Mo-99 1.39E-11 2.31E-05 1.19E-12 1.99E-06 Tc-99m 1.06E-11 1.76E-07 9.58E-13 1.60E-08 Ru-103 2.02E-11 6.73E-05 3.52E-12 1.17E-05 Ru-106 2.70E-10 5.40E-02 5.74E-11 1.15E-02 Rh-103m 2.02E-11 4.04E-08 3.53E-12 7.05E-09 Rh-106 2.70E-10 2.70E-03 5.74E-11 5.74E-04 Te-129m 5.04E-13 5.04E-06 8.49E-14 8.49E-07 Te-129 3.60E-11 1.20E-06 1.94E-12 6.46E-08 Te-131m 3.19E-12 1.59E-05 2.49E-13 1.25E-06 Te-131 7.76E-12 3.88E-06 2.97E-13 1.49E-07 Te-132 3.70E-12 4.11E-05 3.27E-13 3.64E-06 Ba-137m 2.70E-11 2.70E-04 5.86E-12 5.87E-05 Ba-140 3.11E-11 5.19E-05 3.97E-12 6.61E-06 La-140 5.73E-11 1.15E-04 6.30E-12 1.26E-05 Ce-141 3.97E-13 1.98E-06 6.63E-14 3.32E-07 Ce-143 5.95E-12 8.50E-06 4.70E-13 6.71E-07 Ce-144 1.19E-11 1.99E-03 2.52E-12 4.20E-04 12.2-32 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-19 Estimated Average Airborne Radioactivity Concentrations In The Containment Building (Page 3 of 3)

LOWER COMPARTMENT UPPER COMPARTMENT Concentration Concentration 6 Hour Purge 10CFR20 DAC 6 Hour Purge 10CFR20 DAC Isotope (Ci/cc) Fraction (Ci/cc) Fraction Pr-143 6.85E-12 2.28E-05 9.25E-13 3.08E-06 Pr-144 1.19E-11 2.39E-07 2.52E-12 5.04E-08 Np-239 4.74E-12 5.27E-06 3.98E-13 4.43E-07 Total 24.20 6.78 2.00 0.45 RADIATION SOURCES 12.2-33

WATTS BAR WBNP-84 Table 12.2-20 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Turbine Building (Page 1 of 3)

Turb. Bldg Equilibrium Conc. 10CFR20 Isotope (Ci/cc) DAC Fraction Kr-85m 3.29E-14 1.64E-09 Kr-85 5.09E-14 5.09E-10 Kr-87 2.77E-14 5.54E-09 Kr-88 5.64E-14 2.82E-08 Xe-131m 1.24E-13 3.09E-10 Xe-133m 1.42E-14 1.42E-10 Xe-133 4.85E-13 4.85E-09 Xe-135m 2.22E-14 2.46E-09 Xe-135 1.76E-13 1.75E-08 Xe-137 2.85E-15 2.85E-08 Xe-138 1.77E-14 4.43E-09 Br-84 7.50E-16 3.75E-11 I-131 1.30E-14 6.52E-07 I-132 2.99E-14 9.98E-09 I-133 3.71E-14 3.71E-07 I-134 2.45E-14 1.22E-09 I-135 5.64E-14 8.06E-08 Rb-88 1.61E-14 5.37E-10 Cs-134 2.18E-15 5.45E-08 Cs-136 2.57E-16 8.56E-10 Cs-137 2.82E-15 4.70E-08 H-3 9.24E-10 4.62E-05 Na-24 8.54E-15 4.27E-09 Cr-51 6.99E-16 8.73E-11 Mn-54 3.66E-16 1.22E-09 Fe-55 2.77E-16 3.47E-10 Fe-59 6.77E-17 6.77E-10 Co-58 1.04E-15 3.48E-09 Co-60 1.22E-16 1.22E-08 12.2-34 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-20 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Turbine Building (Page 2 of 3)

Turb. Bldg Equilibrium Conc. 10CFR20 Isotope (Ci/cc) DAC Fraction Zn-65 1.11E-16 1.11E-09 Sr-89 3.22E-17 5.36E-10 Sr-90 2.77E-18 1.39E-09 Sr-91 1.61E-16 1.61E-10 Y-90 2.77E-15 9.23E-09 Y-91m 2.75E-17 3.93E-13 Y-91 1.23E-18 2.45E-11 Y-93 7.01E-16 7.01E-10 Zr-95 8.76E-17 1.75E-09 Nb-95 6.32E-17 1.26E-10 Mo-99 1.34E-15 2.23E-09 Tc-99m 6.80E-16 1.13E-11 Ru-103 1.78E-15 5.92E-09 Ru-106 2.00E-14 3.99E-06 Rh-103m 1.78E-15 3.55E-12 Rh-106 1.85E-14 1.85E-07 Te-129m 4.33E-17 4.33E-10 Te-129 1.27E-15 4.22E-11 Te-131m 3.04E-16 1.52E-09 Te-131 1.65E-16 8.24E-11 Te-132 3.68E-16 4.09E-09 Ba-137m 2.70E-15 2.71E-08 Ba-140 2.88E-15 4.81E-09 La-140 5.17E-15 1.03E-08 Ce-141 3.44E-17 1.72E-10 Ce-143 5.74E-16 8.20E-10 Ce-144 9.09E-16 1.51E-07 Pr-143 5.76E-16 1.92E-09 Pr-144 9.09E-16 1.82E-11 RADIATION SOURCES 12.2-35

WATTS BAR WBNP-84 Table 12.2-20 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Turbine Building (Page 3 of 3)

Turb. Bldg Equilibrium Conc. 10CFR20 Isotope (Ci/cc) DAC Fraction Np-239 4.70E-16 5.22E-10 Total 5.20E-05 12.2-36 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-21 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Auxiliary Building Aux. Bldg Equilibrium Conc.

Isotope (Ci/cc) 10CFR20 DAC Fraction Kr-85m 2.30E-09 1.15E-04 Kr-85 3.70E-09 3.70E-05 Kr-87 1.99E-09 3.98E-04 Kr-88 3.95E-09 1.97E-03 Xe-131m 9.09E-09 2.27E-05 Xe-133m 9.94E-10 9.94E-06 Xe-133 3.51E-08 3.51E-04 Xe-135m 1.21E-09 1.34E-04 Xe-135 1.24E-08 1.24E-03 Xe-137 1.46E-10 1.46E-03 Xe-138 1.07E-09 2.69E-04 Br-84 4.72E-12 2.36E-07 I-131 1.70E-11 8.49E-04 I-132 7.50E-11 2.50E-05 I-133 5.27E-11 5.27E-04 I-134 1.10E-10 5.49E-06 I-135 9.68E-11 1.38E-04 Rb-88 1.37E-09 4.58E-05 Cs-134 4.21E-14 1.05E-06 Cs-136 5.17E-15 1.72E-08 Cs-137 5.59E-14 9.32E-07 H-3 1.39E-09 6.95E-05 Na-24 2.82E-13 1.41E-07 Cr-51 1.86E-14 2.32E-09 Mn-54 9.58E-15 3.19E-08 Fe-55 7.18E-15 8.98E-09 Fe-59 1.80E-15 1.80E-08 Co-58 2.76E-14 9.20E-08 Co-60 3.18E-15 3.18E-07 RADIATION SOURCES 12.2-37

WATTS BAR WBNP-84 Table 12.2-21 Estimated Average Airborne Equilibrium Radioactivity Concentrations In The Auxiliary Building Zn-65 3.06E-15 3.06E-08 Sr-89 8.38E-16 1.40E-08 Sr-90 7.18E-17 3.59E-08 Sr-91 5.72E-15 5.72E-09 Y-90 7.18E-17 2.39E-10 Y-91m 2.89E-15 4.13E-11 Y-91 3.18E-17 6.36E-10 Y-93 2.50E-14 2.50E-08 Zr-95 2.34E-15 4.67E-08 Nb-95 1.68E-15 3.36E-09 Mo-99 3.84E-14 6.40E-08 Tc-99m 2.87E-14 4.78E-10 Ru-103 4.50E-14 1.50E-07 Ru-106 5.40E-13 1.08E-04 Rh-103m 4.50E-14 9.00E-11 Rh-106 5.40E-13 5.40E-06 Te-129m 1.14E-15 1.14E-08 Te-129 1.29E-13 4.30E-09 Te-131m 9.02E-15 4.51E-08 Te-131 3.46E-14 1.73E-08 Te-132 1.02E-14 1.13E-07 Ba-137m 5.33E-14 5.33E-07 Ba-140 7.81E-14 1.30E-07 La-140 1.50E-13 3.00E-07 Ce-141 9.01E-16 4.50E-09 Ce-143 1.68E-14 2.40E-08 Ce-144 2.40E-14 4.00E-06 Pr-143 1.69E-14 5.63E-08 Pr-144 2.40E-14 4.80E-10 Np-239 1.32E-14 1.47E-08 Total 7.79E-03 12.2-38 RADIATION SOURCES

WATTS BAR WBNP-84 Table 12.2-22 Estimated Average Airborne Radioactivity Concentrations In The Instrument Room Concentration Concentration 6 Hour 6 Hour Purge 10CFR20 DAC Purge 10CFR20 DAC Isotope (Ci/cc) Fraction Isotope (Ci/cc) Fraction Kr-85m 8.55E-08 4.27E-03 Zn-65 7.28E-14 7.28E-07 Kr-85 3.64E-07 3.64E-03 Sr-89 1.71E-14 2.85E-07 Kr-87 5.39E-08 1.08E-02 Sr-90 1.79E-15 8.94E-07 Kr-88 1.35E-07 6.76E-02 Sr-91 5.84E-14 5.84E-08 Xe-131m 5.29E-07 1.32E-03 Y-90 1.79E-15 5.96E-09 Xe-133m 4.50E-08 4.50E-04 Y-91m 3.28E-14 4.69E-10 Xe-133 1.75E-06 1.75E-02 Y-91 1.10E-15 2.19E-08 Xe-135m 1.20E-08 1.34E-03 Y-93 2.57E-13 2.57E-07 Xe-135 5.00E-07 5.00E-02 Zr-95 4.95E-14 9.90E-07 Xe-137 3.77E-10 3.77E-03 Nb-95 4.29E-14 8.59E-08 Xe-138 9.87E-09 2.47E-03 Mo-99 4.47E-13 7.45E-07 Br-84 8.43E-13 4.21E-08 Tc-99m 3.50E-13 5.83E-09 I-131 9.44E-12 4.72E-04 Ru-103 8.78E-13 2.93E-06 I-132 2.55E-11 8.50E-06 Ru-106 1.30E-11 2.61E-03 I-133 2.34E-11 2.34E-04 Rh-103m 8.79E-13 1.76E-09 I-134 2.62E-11 1.31E-06 Rh-106 1.30E-11 1.30E-04 I-135 3.93E-11 5.62E-05 Te-129m 2.15E-14 2.15E-07 Rb-88 1.17E-07 3.91E-03 Te-129 8.60E-13 2.87E-08 Cs-134 1.03E-12 2.59E-05 Te-131m 9.92E-14 4.96E-07 Cs-136 7.77E-14 2.59E-07 Te-131 1.39E-13 6.95E-08 Cs-137 1.40E-12 2.34E-05 Te-132 1.20E-13 1.34E-06 H-3 1.37E-07 6.84E-03 Ba-137m 1.32E-12 1.32E-05 Na-24 2.97E-12 1.49E-06 Ba-140 1.17E-12 1.95E-06 Cr-51 3.36E-13 4.21E-08 La-140 2.02E-12 4.04E-06 Mn-54 2.30E-13 7.67E-07 Ce-141 1.69E-14 8.45E-08 Fe-55 1.77E-13 2.21E-07 Ce-143 1.86E-13 2.66E-07 Fe-59 3.60E-14 3.60E-07 Ce-144 5.75E-13 9.58E-05 Co-58 5.93E-13 1.98E-06 Pr-143 2.64E-13 8.80E-07 Co-60 7.88E-14 7.88E-06 Pr-144 5.75E-13 1.15E-08 RADIATION SOURCES 12.2-39

WATTS BAR WBNP-84 Table 12.2-22 Estimated Average Airborne Radioactivity Concentrations In The Instrument Room Concentration Concentration 6 Hour 6 Hour Purge 10CFR20 DAC Purge 10CFR20 DAC Isotope (Ci/cc) Fraction Isotope (Ci/cc) Fraction Np-239 1.52E-13 1.68E-07 Total 0.1775 12.2-40 RADIATION SOURCES

WATTS BAR WBNP-84 Figure 12.2-1 Neutron Flux Distribution Radiation Sources 12.2-41

WATTS BAR WBNP-84 Figure 12.2-2 Gamma Flux Distribution 12.2-42 Radiation Sources

WATTS BAR WBNP-84 Figure 12.2-3 Reactor Coolant N-16 Activity Radiation Sources 12.2-43

WATTS BAR WBNP-84 THIS PAGE INTENTIONALLY BLANK 12.2-44 Radiation Sources

WATTS BAR WBNP-88 12.3 RADIATION PROTECTION DESIGN FEATURES 12.3.1 Facility Design Features Some specific design features to limit inplant radiation exposures are provided in the following sections.

Where practicable, instruments and components which require frequent maintenance or calibration are located in the lowest practicable radiation fields. This practice serves the twofold purpose of reducing exposure to operations personnel as well as lessening radiation damage to this equipment.

Penetrations of shielding and containment walls are located when practicable, and designed so as to minimize exposures. Details of design considerations in the location of shield wall penetrations are provided in Section 12.3.2.2.

Radiation sources and routinely occupied areas are separated wherever possible. In general, pipes or ducts containing potentially highly radioactive fluids do not pass through routinely occupied areas and long runs of radioactive piping are restricted to shielded pipe chases where practicable. Some general areas do have radioactive pipes/ducts passing through. This piping/ductwork was considered in the development of the radiation zone maps (Figs. 12.3-1 through 12.3-15). These areas will have limited access based on actual radiation levels. In addition, an effort is made to assure that piping entering an equipment cubicle serves only the equipment housed in that cubicle.

Design features are incorporated to minimize the spread of contamination and to facilitate decontamination in the event spillage occurs. Floor drains are provided in equipment cubicles to prevent the spread of radioactive liquids. Tight fitting doors with seals are utilized in access ways to cubicles housing radioactive gas handling equipment or equipment which processes high temperature fluids. If the doors are louvered, they are equipped with back draft dampers. Walls and floors in these areas are coated with special materials which are easily cleaned in case excessive leakage of contaminants has occurred.

The layout of ducts and pipes is designed to minimize buildup of contamination.

Lengthy runs of horizontal radioactive piping are avoided where possible. Vents and drains on piping are located so as to minimize potential crud traps.

The Ventilation System is designed to ensure control of airborne contaminants and for easy access and service to keep doses As Low As Reasonably Achievable (ALARA) during alterations, maintenance, decontamination, and filter changes. Air flow patterns are controlled throughout the plant such that cleaner areas are exhausted to areas of higher potential airborne radioactivity which are then exhausted to the atmosphere through air cleanup units. Air cleanup units are designed for ease of maintenance and to facilitate removal of filters to minimize personnel exposure from contaminated filters.

Ventilation system design in plant buildings is discussed in Section 9.4.

RADIATION PROTECTION DESIGN FEATURES 12.3-1

WATTS BAR WBNP-89 Radiation monitoring devices are located throughout the plant to assist in the control of personnel exposure. Area and airborne radioactivity monitoring equipment with local readout is provided in selected areas to which personnel normally have access.

Portable continuous air monitors, area monitors, and hand-held survey instruments are available to provide surveillance of areas not covered by fixed monitors and/or supplement fixed monitor surveillance. The locations of radiation monitors are discussed in detail in Section 11.4 and Section 12.3-4. Section 11.4 provides information on process and effluent radiation monitors.

Where practicable, shielding is provided between radiation sources and areas to which personnel may have normal or routine access, and shielding is designed for maintaining doses ALARA. Where practicable, the designs include component arrangement such that those components which may require periodic maintenance are shielded from the components most likely to be high radioactive sources. Movable shielding and convenient means for its utilization are available for use where permanent shielding is needed but impractical. Details of shielding design are discussed in Section 12.3.2.

Remote handling equipment is provided wherever it is needed and practicable. Valves with remote operators are used in plant areas when manual operation is frequently required. Valve and valve gallery locations are discussed in Section 12.3.2.2.

Provisions are made to remotely remove certain cartridge filters which are potentially highly radioactive. Remote handling equipment is used to package and transfer liquid and solid waste products. Remote handling equipment is also used to transfer new fuel and spent fuel assemblies.

Instrumentation and sampling is designed to ensure that exposures will be ALARA during such routine operations as sampling off-gas, primary coolant, liquid waste, and instrument calibration. Liquid sampling from highly radioactive components is remoted to a hot sampling room. All radiation protection features, such as shielding, are designed to allow normal plant operations to continue unimpeded when radioactivity inventories are at design levels. Normal plant design radiation levels are based on ANSI/ANS-18.1, 1984, which represents approximately 1/8% failed fuel. Radiation design levels for accident conditions are based on TID-14844 methodology.

Layouts of the containment and surrounding Shield Buildings and of the Auxiliary, Control, and Turbine Buildings are provided in Figures 12.3-1 through 12.3-19. While generally to scale, these drawings cannot be scaled to determine accurately the thickness of concrete shield walls. Shield wall thicknesses are therefore tabulated in Table 12.3-6.

The layouts provide the radiation zone designations including zone boundaries and maximum expected radiation levels during all phases of normal plant operation.

Refueling/shutdown radiation levels will be no higher and in most areas will be much lower than those shown. These layouts show elevations and room numbers for reference to Table 12.3-6, so that shield wall thickness may be readily determined for any location. The layouts also show controlled access areas, decontamination areas, the location of the onsite laboratory for analysis of chemical and radioactivity samples 12.3-2 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 and the location of the counting room. The locations of area radiation monitors are tabulated according to elevation and nearest position coordinates in Table 12.3-4.

Airborne radioactivity monitors are tabulated in Table 12.3-5 and special radiation monitors are tabulated in Table 12.3-7. These elevations and position coordinates are readily locatable on the layout drawings, Figures 12.3-1 through 12.3-19.

12.3.2 Shielding 12.3.2.1 Design Objectives The design objectives of the plant shielding are the following:

(1) During normal operation, including anticipated operational occurrences, to restrict occupational doses to the 10 CFR 20 limits and insure operational radiation dose to personnel is ALARA.

(2) To restrict off-site exposures in accordance with the As Low As Reasonably Achievable (ALARA) provisions in 10 CFR 50.

(3) To limit, under accident conditions, the off-site exposure from activity in the containment so that the total exposure from this source and from airborne radiation will not exceed the 10 CFR 100 exposure limits.

(4) To satisfy the requirements of 10 CFR 50, Appendix A, Criterion 19. Sufficient radiation protection is provided to permit access and occupancy of the main control room under accident conditions without personnel receiving excessive radiation exposure. The design also provides limited access, defined in Section 12.3.2.2, to other plant areas during accident conditions.

The dose an operator receives during any such extra-control room visits and the dose received while gaining access to and occupying the main control room will not exceed exposures of 5 rem deep dose equivalent (DDE).

12.3.2.2 Design Description Plant Shielding Expected frequency, duration of occupancy, and access controls determine what exposure rates will be allowed in all interior and other on-site areas in order to assure that shielding design objectives will be met. Each area is classified as one of seven types listed on Figure 12.3-1.

In numerous cases where access requirements are expected to range from almost continuous occupancy to a few hours per week (access Types II and III), shielding is required to achieve acceptable exposure rate levels.

Shield Walls Presented in this section are the criteria for the erection of the plant shield walls and for penetrations through these walls. The calculational methods used to determine the thickness and other dimensions of the shield walls are given at the end of this section.

RADIATION PROTECTION DESIGN FEATURES 12.3-3

WATTS BAR WBNP-88 See Section 12.2.1 for the relationship between the Sequoyah and Watts Bar shielding design source terms.

Many structural walls also serve a shielding requirement which often sets the wall thickness. Some walls serve only a shielding function. Most of these shielding walls are cast in place up to within 2 inches of the ceiling above. When necessary, this gap between wall and ceiling is filled over part of the wall thickness with grout. Those shield walls or portions of shield walls that are subject to removal for equipment repair or replacement are constructed of solid concrete blocks.

Except for two applications, which are cited in later subsections, the poured concrete shield walls throughout the plant are ordinary concrete with a minimum density of 145.0 lb/ft3.

Areas where design DDE exposure rates are greater than 5 mrem/hr incorporate permanent shielding, to the degree feasible, to prevent the need for temporary shielding.

Access to many equipment enclosures is provided through the sidewalls of the compartments. In these cases, the effectiveness of the shield walls in limiting exposure rates outside the equipment enclosures is maintained by providing labyrinth entrances. Access to some equipment enclosures, principally filter, and demineralizer is through the floor above. In these cases, the removable concrete floor slab that provides the entrance generally has the same thickness as the cubicle walls.

The design criterion for shield wall penetrations in the Auxiliary Building, such as those for piping and ventilation ducts, is to locate them whenever practical so that their effect on the DDE exposure rates in accessible areas outside the shielded enclosure is minimized. Often this criterion is satisfied by locating the penetrations as nearly as possible to the corners and to the ceiling of the shielded enclosure. In using this technique, however, consideration is given to the increased length of piping sources that may result. If direct or reflected radiation passing through the penetrations of a shield wall creates a radiation area outside the wall, the criteria given for the erection of shield walls are used to establish the necessity for a wall to shield this area.

The following general shielding considerations are employed in the arrangement of Shield Building penetrations:

(1) Where practical, most penetrations of the Shield Building except those that connect the Shield Building to a shielded enclosure in the Auxiliary Building, are opposite unpenetrated areas of the crane wall. This arrangement adequately shields outside areas and areas inside the Auxiliary Building from sources inside the containment shell during normal operation. When this arrangement is not used, shadow shields or high-density silicon elastomer fill are provided to eliminate radiation streaming from major sources inside the containment to areas outside the Shield Building.

12.3-4 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 (2) Radiation sources in the annulus between the containment and the Shield Building are located behind unpenetrated portions of the Shield Building or behind the Shield Building penetrations that connect the Shield Building to shielded enclosures in the Auxiliary Building.

(3) Penetrations of crane wall sections that provide necessary shielding for containment areas accessible during power operations are avoided when possible.

(4) Shadow shields or high-density silicon elastomer fill are provided at Shield Building penetrations that connect the Shield Building to unshielded areas where access cannot be completely controlled during accident conditions.

Valve and Valve Operation Stations The following arrangement guidelines were used in the initial plant layout for manually operated valves that control process equipment function. The radiation dose rates specified were estimated at the time the initial layout was performed. Actual radiation dose to personnel operating valves will be controlled by radiation work permit, which includes surveys to determine actual radiation dose at the work location and limitation of time spent at that location.

(1) Valves are located and operated in the enclosure with the controlled equipment. This arrangement is used only when design level activities in the equipment and piping and anticipated occupancy for valve operation are such that acceptable exposure limits will not be exceeded. This arrangement is not used if the whole body exposure rate at the valve is greater than 0.1 rem/hr. The limit imposed in the case of each valve depends on expected occupancy requirements and is generally much less than 0.1 rem/hr. Another requirement for using this arrangement is that sources and piping in the equipment enclosure can be sufficiently removed, without economic penalty, to allow valve maintenance or that design activities are low enough to keep personnel doses under acceptable levels during valve maintenance without source removal. (Source removal can involve pumping or draining a liquid, venting a gas, flushing demineralizer resin, or replacing a filter cartridge). For this purpose, the acceptable whole body exposure rate is 0.006 rem/hr. To perform maintenance for an 8-hour shift at an average dose rate above this level, an employee needs the approval of his supervisor, the health physicist or his authorized representative, and the shift supervisor.

(2) Valves are located and operated in a radiation area outside the equipment enclosure. With this arrangement, whole body exposure rates at the valve must be less than 0.1 rem/hr, and generally much lower limits are set.

(3) In the third type of arrangement, valves are located in the equipment enclosure but are operated from behind a shield wall. For this arrangement, whole body exposure rates at the valve operating station must be less than 0.015 rem/hr. The exposure limitations during valve maintenance are the same as those for the first arrangement.

RADIATION PROTECTION DESIGN FEATURES 12.3-5

WATTS BAR WBNP-88 (4) In this arrangement, valves are located in a valve gallery. Generally, a number of valves share a valve gallery. Typically, these are most of the valves that serve a few identical or similar plant components. One side of the valve gallery is formed by a shield wall which separates the valves from the process equipment. The opposite side of the gallery is a shield wall which is penetrated by either extension stem arrangements joining valves to hand-wheel operators or by flexible shaft controls.

The extension stem is solid metal and the annular space between extension stem housing and shield wall sleeve is grout filled. With this arrangement, the effectiveness of the shield wall between valves and handwheel operators is virtually undisturbed and the whole body exposure rates at the handwheel are less than 0.001 rem/hr. The flexible shaft control is used in the case of a few filters. In some cases, ducts for the shafts follow an oblique or curved path through the wall to prevent direct radiation streaming from high intensity sources. The design whole body exposure level outside the valve gallery for this arrangement is 0.0025 rem/hr.

The first design objective of the valve gallery is to allow valve maintenance without first removing the sources from the process equipment. Some of the design guides to achieve this objective are the following:

(a) Penetrations through the shield wall between the equipment enclosures and valve gallery are as near the ceiling and a close to the corner of the equipment enclosures as practical.

(b) Piping runs in the gallery, that will contain radioactive fluid when the control valve is isolated for maintenance, are kept as short as practical.

(c) Excessive annular spaces between pipe and pipe sleeve in the wall between equipment and valves are avoided.

With these precautions, it is expected that the design objective of 0.006 rem/hr whole body exposure rate will be achieved when the process equipment contains up to a significant fraction of design level activity. The design objective should be achieved in most cases even when the process equipment contains design level activity. As an outside limit, the design assures an exposure rate of less than 0.1 rem/hr in the valve gallery during valve maintenance without removal of the process equipment sources.

Even at exposure levels of 0.1 rem/hr some valve inspection and maintenance would be possible.

A second objective for locating some of the valves in valve galleries instead of in the equipment enclosures is that even after removal of the process sources, the remaining activity on he inside walls of the equipment and/or high contamination levels in the enclosure may require extensive decontamination work before valve maintenance if the valve is located in the enclosure.

A third objective for locating control valves in valve galleries is that this arrangement provides a second shield between process equipment and general access areas. This 12.3-6 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 is a worthwhile consideration when any unanticipated shielding deficiencies can result in high exposure rates.

Another advantage is that, in the unlikely event of valve operator failure, the valve gallery arrangement allows limited direct operation at the valve location until maintenance is performed.

Most of the advantages of locating hand-operated valves in valve galleries also apply to the location of remote-manual (motor operated or pneumatically-operated) valves in valve galleries.

Manually operated valves used to isolate, drain, or vent process equipment such as pumps that contain relatively small amounts of activity are generally located and operated in the enclosure with the equipment. As a rule, remoting the valve and/or its operation from the equipment is a design consideration only when one or both of the following conditions can exist: (1) the anticipated dose from the process equipment during valve operation is significant and (2) the anticipated dose received from the equipment during valve maintenance is significant and large compared with that which could be received if a remote valve station were used.

A valve is never used to isolate, drain, or vent process equipment located and operated in the enclosure with the equipment if the whole body exposure level is greater than 0.1 rem/hr. The limit selected for each valve depends on the expected occupancy time at the valve station and is generally much less than 0.1 rem/hr. If anticipated exposure rates are too high to allow location and manual operation of the valve in the enclosure with the equipment, one of the following procedures is used: (1) the operation of the valve is from behind a shield wall which limits the whole body exposure rate at the operating location to less than 0.015 rem/hr or (2) the valve is located in a valve gallery and operation of the valve is from behind the valve gallery wall which restricts the whole body exposure rate at the valve operating location to 0.0025 rem/hr. Typically, these valves share a valve gallery with the equipment control valves.

Motor-operated or pneumatic valves that isolate, drain, or vent process equipment are located in valve galleries if process equipment activity levels could be high enough to prohibit emergency access to the valves.

Primary and Secondary Shielding The primary shield consists of the following parts:

(1) Shield elements inside the reactor pressure vessel. These elements, which are the core baffle, the core barrel, neutron pads, and water annuli, provide a water shield and a steel shield, each several inches thick.

(2) The reactor pressure vessel.

(3) A concrete structure surrounding the reactor vessel from the floor at the 702.78 foot elevation to the floor at the 725.12 foot elevation. The concrete thickness is 5 feet 9 inches on the radius through each of eight out-of-core RADIATION PROTECTION DESIGN FEATURES 12.3-7

WATTS BAR WBNP-88 neutron detector slots. On all other radii, the concrete thickness opposite the active fuel is 8 feet 6 inches. There is an opening in the shield at each of the eight primary coolant pipes. Four of the openings start at the vessel flange surface elevation of 725.12 feet and go down to Elevation 712.71 feet. The other four openings extend from the vessel flange surface to Elevation 692.00 feet.

That part of the opening above each pipe (nozzle inspection ports) is filled during power operation with a removable plug. Removal of the plugs during shutdown allows inspection of the weld joints between the primary coolant pipes and the reactor vessel nozzles. Inspection time available will be very limited since exposure levels under pressure vessel equilibrium Co-60 and Fe-59 activity conditions will be on the order of 10 rem/hr at the bottom of the opening and 1 rem/hr at the top.

Except across the refueling canal, the primary concrete structure extends upward at reduced thickness (minimum is 2'6") from the 725.12 feet elevation to the operating floor (Elevation 756.63 feet). (The blowout panels in this upper structure are located just under the floor at Elevation 756.63 feet. The panels extend from Elevation 754-13 feet down to Elevation 749.63 feet. With this arrangement, radiation from the reactor vessel that penetrates the blowout panel area is attenuated by at least one reflection off concrete before it reaches accessible plant areas outside the primary concrete).

The upper part of the primary concrete shielding is completed by the walls of the refueling canal which extend upward from Elevation 709.23 feet, by the control rod drive missile shield and by a gate which spans the refueling canal from Elevation 756.63 feet down to Elevation 725.12 feet. The control rod drive missile shield and the gate are removed during refueling. The primary shielding makes possible necessary access inside the crane wall during shutdown.

The secondary shield consists principally of the crane wall, the Shield Building, the concrete operating floor at Elevation 756.63 feet, and the concrete structures which combine with the crane wall to enclose those sections of the steam generators and the portion of the pressurizer that extend above Elevation 756.63 feet.

In addition to their providing biological radiation protection, the primary and secondary shielding are arranged and structured to provide additional shielding functions such as:

(1) The primary shielding elements inside the vessel attenuate neutron flux sufficiently to prevent excessive radiation damage to the reactor vessel.

(2) The primary shielding prevents excessive radiation damage to plant components from neutron and gamma radiations, and the secondary shielding prevents excessive radiation damage to plant components from gamma radiation.

(3) The metal and water inside the pressure vessel and the pressure vessel itself serve to reduce the heat flux from neutron and gamma radiation at the vessel outer surface. Cooling necessary to avoid high temperatures and possible dehydration in the surrounding concrete is, thus, an easier task.

12.3-8 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 (4) Parts of the primary and secondary shields serve as portions of the divider, necessary for the ice condenser containment, between lower and upper containment compartments.

(5) The Shield Building, which is part of the secondary shielding, is also part of the double containment.

Personnel enter and leave the containment vessel through either of two personnel air locks. To protect (from primary coolant system radiation) personnel entering the containment through the airlock from the platform at Elevation 716.00 feet, heavy concrete (218.0 LB/ft3) is used in a section of the crane wall. With the reactor at significant power levels, personnel access to the lower compartment, which is access Type IVa, will be prohibited except under cases of extreme emergency. During full power operation, the upper compartment and the ice condenser upper plenum, access Type IVa areas, will be entered infrequently but as necessary for upper compartment inspection and ice bed and ice condenser inspection and maintenance. The seal table and instrument room, which are access Type IV areas, will be entered routinely during full power operation. The accumulator rooms, ventilation equipment rooms, and tunnel area outside the crane wall will be entered from the seal table and instrument room only as needed and as radiation and airborne contamination permit. Some of these rooms contain access Type IVa areas. Access to the annulus between the containment vessel and the Shield Building is not normally required during power operation; however, access, if necessary, is through a hatch. Most annulus areas are access Type IV areas although some areas opposite crane wall penetrations are access Type IVa areas.

Auxiliary Building Shielding Shielding in the fuel handling area of the Auxiliary Building is discussed in a following subsection. The balance of the shielding in the Auxiliary Building protects personnel, during normal operation including anticipated operational occurrences, from the components and piping of the following systems and facilities:

(1) Chemical and Volume Control System (CVCS)

(2) Waste Disposal Systems (WDS)

(3) Residual Heat Removal System (RHR)

(4) Spent Fuel Pool Cooling and Cleanup System (SFPCCS)

(5) Sampling System collection and analysis facilities The hot instrument shop and decontamination area enclosures furnish some minimal shielding, but their main function is to minimize the spread of contamination.

The Auxiliary Building shielding is designed to limit DDE exposure levels in accessible corridors and open spaces in the building to 0.001 rem/hr (access Type II); however, exceptions occur at certain shield wall penetrations. If the exposure rate at a RADIATION PROTECTION DESIGN FEATURES 12.3-9

WATTS BAR WBNP-89 penetration exceeds 0.005 rem/hr, the procedures in Section 12.5.3 for designating radiation areas apply. Auxiliary Building shielding is also designed so that equipment areas may be entered for maintenance without shutdown of adjacent operating systems or system equipment. Satisfying this requirement results in a high degree of compartmentalization in the building.

Most piping carrying fluid of high specific activity is routed through shielded pipe chases. The pipe chase walls have a minimum thickness of 27 inches of concrete, which will significantly reduce the DDE exposure rate from a 14-inch Residual Heat Removal (RHR) System pipe carrying reactor coolant water. Pipe chases run along the A-5 and A-11 coordinate lines from Elevation 676.0 ft. to Elevation 757.0 ft (see Figures 12.3-4, 12.3-8, 12.3-10 and 12.3-12). The pipe chase areas are enlarged at one end between the floors at Elevation 713.0 feet and Elevation 737.0 feet to form Shield Building penetration areas. Most radioactive fluid carrying pipes running from the containments to the Auxiliary Building pass through these pipe chase sectors which extend from approximately Az 270 to approximately Az 300 (See Figure 12.3-3). Another pipe chase runs along the fuel transfer canal and adjoins the A-5 and A-11 line pipe chases between the floors at Elevation 713.0 feet and Elevation 737.0 feet. A concrete partition in this pipe chase along the A-8 line, between units, inhibits the spread of contamination from one unit to the other should a pipe rupture occur.

Fuel Transfer Shielding During fuel transfer operations, the refueling canal and the region above the open reactor vessel are filled with borated water to Elevation 749.12 feet. The water level in the fuel transfer canal and spent fuel pit, which are in the Auxiliary Building, is also at Elevation 749.12 feet. The bottom of the refueling canal is at Elevation 709.23 feet in the fuel assembly tilting device area and at Elevation 713.87 feet elsewhere. A fuel assembly is transferred from the reactor vessel through the refueling canal toward the Auxiliary Building. It travels in a fuel transfer tube from the containment to the fuel transfer canal in the Auxiliary Building, and it is then moved into a storage location in the adjacent spent fuel pool. The reactor cavity filtration system assures water clarity in the reactor cavity during refueling.

After the fuel transfer has begun, the principal radioactive sources in the proximity of the fuel assembly transfer path are the following: (1) activity in the water which is a mixture of reactor coolant and water from the refueling water storage tank and (2) the fission product inventory in the fuel assembly being transferred. The activity in the water will not normally be above 0.01 Ci/cc of nontritium activity when a fuel assembly is moved from the vessel. If the activity is above this level, it will be reduced to this level with the Spent Fuel Pool Cooling System (SFPCS) equipment.

The minimum water shield above the active fuel region of a spent fuel assembly as it moves from the reactor vessel to the storage position in the spent fuel pool is 10 feet except when the assembly is in the fuel transfer tube. The design of the transfer equipment incorporates restraints to assure that this minimum water shield is maintained. The calculated DDE exposure rate to a person on the Spent Fuel Pit Bridge or the manipulator crane bridge resulting from a fuel assembly at its maximum elevation during fuel transfer activities is less than 2.5 mrem/hr. The transfer of spent 12.3-10 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 fuel assemblies does not generate any high radiation areas in accessible plant areas.

The minimum shielding between the fuel assembly and the emergency passageway is 3 feet 0 inch of heavy concrete (218.0 LB/ft3). The DDE dose rate in the passageway is less than 50 m rem/hr. The minimum shielding inside the primary containment between fuel assembly and personnel on the floor at Elevation 716.0 feet is 1 foot of water and over 5.5 feet of ordinary concrete. The corresponding maximum DDE exposure level is less than 5 m rem/hr. During fuel assembly transfer, the region in the annulus between the steel containment and the Shield Building is protected from the fuel assembly by concrete and water equivalent to more than 6 feet of concrete. A small access opening is provided through the shielding in the annulus to allow for inspection of the fuel transfer tube. This opening is normally filled with solid concrete blocks which are removed only when access for inspection purposes is required. A radiation streaming gap between the steel containment and the concrete on each side of it in the vicinity of fuel transfer tube is avoided by offsetting the concrete and attaching to each side of the steel containment a steel ring. Similarly, offsets in the Shield Building concrete and in the Auxiliary Building wall in the area of the transfer tube are used to avoid a direct streaming path between these two structures.

When the spent fuel assembly is outside the Shield Building, during passage through the Auxiliary Building wall and fuel transfer canal to the spent fuel pool, it is shielded by a minimum of 6 feet of concrete or by a minimum of 10 feet 6 inches of water. Spent fuel pool concrete walls which separate spent fuel assemblies in their storage locations from the Auxiliary Building access area at Elevation 692.0 feet are 7 feet thick.

Turbine Building and Service Building Activity in the Turbine Building occurs only in the event of steam generator primary-to-secondary leakage. Almost the entire Turbine Building is an access Type I area.

Also located in the Turbine Building are condensate demineralizers and associated regeneration equipment. This equipment is adequately shielded to maintain maximum dose rates in controlled access areas to 0.001 rem/hr. Cubicles in which condensate demineralizers and associated equipment are located are generally designed for Type III or IV access (see Figures 12.3-11 and 12.3-13).

There are several areas of low activity level in the Service Building, such as the protective clothing change room, health physics laboratory, and radiochemical laboratory filter room. These areas furnish necessary shielding, but their principal purpose is to minimize the spread of contamination.

Outside Areas Except for the following, all areas outside the plant buildings are either access Type I or II areas during normal operation including anticipated operational occurrences.

RADIATION PROTECTION DESIGN FEATURES 12.3-11

WATTS BAR WBNP-88 (1) For short periods of time when solid waste shipping is imminent, the casks will be outside. The number of casks allowed outside at any one time is controlled and depends on the exposure rates from each cask. The maximum exposure rate from each cask satisfies the provisions of 49 CFR 173.

Access to the outside region where these casks are located during the short preshipment periods is controlled. The type of control required depends on the designated access type, which in turn is established by the exposure rate.

(2) During solid waste and spent fuel shipment, the area immediately adjacent to the train will be an access Type IVa area.

(3) There are six outside tanks that contain radioactive liquids: Two refueling water storage tanks, two primary water storage tanks, and two condensate storage tanks. The activity in each is low level, and no shielding is required.

Maximum DDE exposure rates at the site boundary from these tanks are 4.3E-7 mrem/hr, 2.1E-9 mrem/hr, and 1.3E-9 mrem/hr, respectively, for a refueling water storage tank, a primary water storage tank, and a condensate storage tank. These dose rates are calculated using computer code QAD-P5Z[6].

Shielding For Accident Conditions Some shielding provided for normal operation also has a function during accident conditions. However, other shielding has a function during accident conditions only.

This accident shielding is required to serve two functions:

(1) it must restrict the exposure at the site boundary from activity in the containment to a small fraction of 10 CFR 100 limits and (2) it must attenuate exposure rates at interior and other onsite locations from activity in the containment to levels which will allow required access. Continuous occupancy of the main control room and the Technical Support Center is required during accident conditions. Infrequent access is required for the operational support center, radio-chem lab, and other plant areas identified below during accident conditions. Analyses have been done to ensure the following post accident activities can be accomplished with dose to workers remaining below 5 rem as required by NUREG-0737, II.B.2.

(1) Continuous main control room and Technical Support Center occupancy is required.

(2) Visits of several minutes duration into the shutdown board rooms to operate breakers and switches must be possible. For these visits which may occur at any time after the start of accident conditions, the operator will wear anti-contamination clothing and have breathing protection.

12.3-12 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 (3) Since a single crew cannot remain in the main control room for the duration of the accident, it must be possible to make the trip from the site boundary to the main control room sometime after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without receiving an excessive dose.

(4) Sampling of gaseous effluents (at the Shield Building vent monitor location) per the requirements of NUREG-0737.

(5) Sampling of the reactor coolant and the containment atmosphere per NUREG-0737 requirements. This sampling is accomplished in the post accident sampling facility (PASF).

(6) Realignment of component cooling water to the spent fuel pool cooling system.

(7) Mission to the Intake Pumping Station to place ERCW backwash screens in service.

(8) Survey of the Auxiliary Building for leaks.

(9) Survey of the main steam lines and the steam generator blowdown during a steam generator tube rupture accident.

(10) Control and verification functions in the switchyard, Diesel Generator Building, and Turbine Building.

The Shield Building is the principal structure that limits exposure at the site boundary and at site exterior locations from activity in the Containment. The Shield Building also, in concert with other shields, limits exposure levels at interior and other on-site locations. The accident shielding functions of the Shield Building are shared by the structures that shield its penetrations, such as the steam line penetrations, the personnel hatches, the equipment hatch, ventilation ducts, and the many smaller penetrations. Some of the structures that shield the Shield Building penetrations are Auxiliary Building external walls. These and other Auxiliary Building walls and the Auxiliary Building ceilings further attenuate radiation from sources within the Containment to improve accessibility during accident conditions.

The ESF equipment compartment shielding provides for emergency maintenance. (To make possible this maintenance, the equipment will be drained before the maintenance begins and the operator will wear anti-contamination clothing and have breathing protection). In the case of ESF equipment, such as the RHR pumps which also operate during normal operation, the shielding required for normal operation is controlling.

The main control room and the Technical Support Center (TSC) are shielded so that the integrated DDE dose from external sources (activity inside the primary containment, in the passing cloud and in surrounding rooms) obtained during occupancy following a loss-of-coolant accident would be a very small fraction of 5.0 rem. The major portion of the Total Effective Dose Equivalent (TEDE) dose can then RADIATION PROTECTION DESIGN FEATURES 12.3-13

WATTS BAR WBNP-88 come from the airborne activity within the main control room. (The dose from this airborne activity which is more difficult to limit than that from the external sources is discussed in Chapter 15 which considers integrated exposures in the main control room under accident conditions from all sources).

In the main control room's shielding design, sufficiently thick walls, ceiling, and floor are provided. In addition, special attention is given to the doorways. Shield doors are provided at the entrances from the Turbine Building to attenuate radiation from the radioactive cloud which is assumed to occupy the Turbine Building.

Analysis shows that shield doors at the small entrances from the main control room to the Auxiliary Building are not necessary.

Shielding Calculations Shielding required to reduce the exposure rates based on conservative source strengths in known source geometries to design objective values were determined with hand calculations and/or with the SDC computer code. A computer program is used to solve the equations for the whole body beta and gamma dose rates from airborne activity. The program also provides the whole body gamma exposure rate after attenuation by a shield. Both the hand calculations and the computer codes employ the point-to-point kernel integration method. The SDC code[1] integrates the basic exponential attenuation point kernel over the various geometries to provide the uncollided gamma-ray flux. Many of the integrations found in the Reactor Shielding Design Manual[2] are utilized. Exposure rates are obtained by multiplying the uncollided flux by the product of a flux-weighted buildup factor and a dose-conversion factor. The hand calculations generally employ the more conservative procedure of multiplication of the buildup factors for the different materials between source point and exposure point.

When reflected gamma rays are important contributors to exposure rates, as in the case of labyrinth design and in the case of some shield penetrations, the angularly and energy distributed source strengths at the reflection surface are calculated using albedo techniques.

Condensate Demineralizer Waste Evaporator Building Components of the Condensate Demineralizer Waste Evaporator are contained in a specially designed building adjacent to the Auxiliary Building near the on-site packaging area. Each component of the processing package is located in separately shielded compartments with the potentially more highly radioactive equipment further separated from equipment with less potential for radioactive contamination. Access to the building is designed for an Access Type III (Radiation Area) with radioactive components located in areas designed for Access Types IVa and IVb (High Radiation Area). The design DDE rate outside equipment cubicles is 0.001 rem/hr, for areas generally accessible on a routine basis. Layout of the building showing radiation protection design features is provided on Figure 12.3-7. See the note in Section 12.2.1.5 relative to the Unit 1 operation without the Condensate Demineralizer Waste Evaporator package.

12.3-14 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-88 12.3.3 Ventilation The plant ventilation systems are designed to assure that air will flow from areas of low potential airborne radioactivity to areas of higher airborne radioactivity. The ventilation systems will maintain concentrations of airborne radioactive material below 10% of the respective derived air concentrations (DAC) and annual limits on intake (ALI) in areas routinely occupied. Additionally, (in accordance with the provisions of 10 CFR 20 Paragraph 20.1701) the systems reduce concentrations of airborne radioactivity in areas not normally occupied, but where maintenance or in-service inspection has to be performed.

12.3.3.1 Airflow Control The Watts Bar Nuclear Plant ventilation systems are designed to supply air to the relatively cleaner plant areas and to exhaust air from areas of potentially higher airborne radioactivity levels. Major plant areas that could be subjected to radiation contamination, and their associated air exhaust flow rates, are shown in Table 12.3-3.

Air that is removed from these areas following an accident is passed through air cleanup units equipped with HEPA filters and charcoal adsorbers before being exhausted to the environment.

12.3.3.2 Typical System The following is an illustrative example of the air cleanup system design. This typical system is designed to provide fresh, clean air inflow to and removal of potentially contaminated air from the Auxiliary Building to assure personnel comfort and safety during normal plant operations.

The Auxiliary Building general ventilation supply subsystem has a major impact on the personnel protection features incorporated in the design of the ventilation system. To control airborne activity, the Auxiliary Building ventilation supply air is delivered to clean areas and areas of general personnel occupancy. This air is then routed to areas of progressively greater contamination potential by natural pressure gradients induced by the exhaust system. Air is supplied as follows:

Area El. 782 Control Rod Drive Equipment Rooms*

El. 757 Fuel Handling Area EGTS Room Blowdown Treatment Room Waste Packaging Area CDWE Building*

El. 737 Penetration Rooms RADIATION PROTECTION DESIGN FEATURES 12.3-15

WATTS BAR WBNP-92 Area General Areas*

Hot Instrument Shop El. 730 CDWE Building*

El. 713 General Areas*

Cask Loading Area Nitrogen Storage Area Post Accident Sampling Facility*

Penetration Rooms El. 692 General Areas*

Penetration Rooms Spent Resin Tank Room Cask Decon Tank Room El. 676 General Areas*

Air is exhausted from the Auxiliary Building rooms/areas as shown in Table 12.3-3 and in Section 9.4. Air supplied to the relatively clean areas, marked by an asterisk, is allowed to follow natural flow paths to air exhausts in areas of potentially greater contamination. Because of its potentially higher levels of radioactivity and requirements for personnel access, the remaining areas listed above are provided with both air supply and air exhaust. The exhaust from the Hot Instrument Shop is from a hooded area over the potentially higher radioactivity areas. Clean outside air is supplied to the Auxiliary Building air supply system through medium efficiency filter cells at each inlet plenum. The rated efficiency of each cell is 85%, based on the NBS atmospheric dust spot test. Each filter bank is provided with a static pressure differential indicating gauge. See Figure 12.3-16 for the general layout plan of the air intakes for this system. The exhaust air during normal operation is released to the outside through the Auxiliary Building vent.

12.3.3.3 Additional Radiation Controls The ventilation system is designed so that filters containing radioactivity will not create a radiation exposure hazard to personnel in normally occupied areas. Normally, waste filters containing radioactive contaminants will be removed from the filter housings, transported to the waste packaging area, and stored in appropriate shipping containers to await shipment to a disposal/storage site. Filters with especially high levels of contamination may be transported to and temporarily stored in the shielded filter storage area. Although the basic design of the air cleanup units was completed prior to the publication of Regulatory Guide 1.52, good general compliance with the 12.3-16 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-89 requirements of Section 4 of that document has been accomplished. See Section 6.5.1 for specific compliance with these requirements.

12.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation 12.3.4.1 Area Radiation Monitoring Instrumentation 12.3.4.1.1 Objectives and Design Basis The area monitoring system assists in compliance with 10 CFR 50, Appendix A, General Design Criteria 19, 63, and 64.

Monitors are provided throughout the plant to monitor exposure rates and to warn personnel if the measured dose rate exceeds selected levels. Monitors are placed as follows:

(1) In areas where personnel routinely work without continuous health physics surveillance if the area is or could become a radiation area during normal operation.

(2) In a few selected locations in the Auxiliary Building to provide knowledge of any increasing trends in general plant exposure rate levels. These monitors also provide warning of hazardous airborne noble gas concentrations.

(3) In specific areas where exposure rates are normally low but in which high exposure rates could occur under postulated anticipated operational occurrences or accident conditions.

(4) At locations outside the Shield Building at which detected exposure rates can provide a measure of airborne concentrations in the containment under postulated accident conditions.

(5) In the control room to indicate exposure rates during accident conditions.

(6) 6.Portable continuous air monitors, area monitors, and hand-held survey instruments are available to provide surveillance of areas not covered by fixed monitors, or to supplement fixed monitor surveillance.

12.3.4.1.2 Operation Characteristics Table 12.3-4 lists the physical location (by building elevation and coordinates), and detector range of each area monitor. The specific location may be found on Figures 12.3-4 through 12.3-12 using the coordinates given in Table 12.3-4. Other characteristics of the area radiation monitoring system are given in the following sections.

12.3.4.1.2.1 Area Monitor Detector The detectors for the Reactor Building upper and lower compartment post accident monitors and upper compartment personnel lock monitor are ion chambers. The other RADIATION PROTECTION DESIGN FEATURES 12.3-17

WATTS BAR WBNP-90 area monitors employ Geiger-Mueller type gamma detectors. Each detector has its own independent high-voltage power supply located on panel 0-M-12 or 1-M-30 in the main control room.

The Reactor Building upper and lower compartment post accident monitors are redundant high range monitors which are required to meet the requirements of RG 1.97 and NUREG-0737.

12.3.4.1.2.2 Main Control Room Ratemeter (0-M-12, 1-,2-M-30)

Ratemeters are of solid-state construction containing a solid-state, high-voltage power supply. Alarms are provided on the ratemeter chassis forhigh radiation and instrument malfunction. Visual and audible alarms are provided for high radiation and instrument malfunction in the main control room.

12.3.4.1.2.3 Local Indicator-Alarm Panel With the exception of the main control room and Reactor Building upper and lower compartment post accident monitors, each monitor has a locally mounted panel which contains an indicator, a visual and audible high radiation alarm, and a power-on light.

12.3.4.1.2.4 Multipoint Recorders (Main Control Room 0-M-12, 1-,2-M-31)

The area monitors are recorded on multipoint recorders on panels 0-M-12, or 1-,2-M-31 in the MCR.

12.3.4.1.2.5 Monitor Sensitivity and Range The ranges of the instrumentation provided are given in Table 12.3-4. The area monitors set points, adjustable over the entire range, are determined by the radiation control group based on operating background levels. The setpoints for the Reactor Building upper and lower compartment post accident monitors are determined by engineering analysis.

12.3.4.1.3 Area Monitor Calibration and Maintenance With the exception of the Reactor Building upper and lower compartment post accident monitors, periodic testing of each area monitor includes a channel calibration performed at least once per fuel cycle, and a channel operational test that is performed at least once per quarter. Testing of the Reactor Building upper and lower compartment post accident monitors is described by plant Technical Specifications.

The channel calibration is the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input from a radioactive calibration source. The channel calibration encompasses the entire channel, including the required sensor, alarm, interlock, display and trip functions. The channel calibration may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

The channel operational test (COT) is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the operability of required 12.3-18 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-90 alarm, interlock, display and trip functions. The COT includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

The built-in checksource function exposes the channel detector to a radioactive source for all channels except those employing an ion chamber detector. The checksource function simulates a detector signal at the channel electronics in channels employing an ion chamber detector. The built-in checksource feature is used to verify functional response of the detector and/or electronics during the performance of the channel operational test. This function is also used by operations personnel at other times, such as after maintenance, to rapidly determine channel operability.

Maintenance is performed, as necessary, if abnormalities are detected during any of the above checks. Unscheduled maintenance will be performed as required.

12.3.4.2 Airborne Particulate Radioactivity Monitoring 12.3.4.2.1 Design Basis The airborne radioactivity monitoring systems are one of the plant features provided to comply with 10 CFR 50, Appendix A, General Design Criteria 19, 63, and 64, and with paragraph 20.1501(a) of 10 CFR 20. Adequate systems are provided to comply with 10 CFR 20 paragraph 20.1201, 20.1204, and 20.1502. Each of the systems monitor an air space to which one or more of the following descriptions are applicable.

(1) Spaces in which there is, during normal operation, a potential for airborne concentrations at DAC levels which, when integrated over a normal 40 hr/wk and 50 wk/yr, would exceed the ALI of any isotope or mixture of isotopes and for which there are requirements for either (a) frequent (i.e., once per shift) visits, each of which is for a duration of at least several minutes, or (b) infrequent but routine visits of at least an hour's duration and for which monitoring systems can be practicably supplied in lieu of provision for safely taking and analyzing grab samples for airborne activity prior to personnel entry.

(2) General spaces (e.g., spaces outside shielded equipment rooms) of buildings that contain equipment which bears, in process fluids, potentially significant radioactivity. (Although the plant ventilation systems normally supply clean air upstream of the spaces containing potential leakage points, monitoring is provided to detect airborne activity in the event of malfunction of the ventilation systems).

(3) Spaces which have requirements for routine occupancy into which significant airborne activity may be introduced directly (e.g., physical barriers to its introduction do not exist). If an existing physical barrier consists of a ventilation system, consideration is given to the magnitudes of possible airborne concentrations should the ventilation system malfunction.

RADIATION PROTECTION DESIGN FEATURES 12.3-19

WATTS BAR WBNP-91 (4) Spaces in which habitable conditions must be guaranteed at all times, even during accident conditions.

12.3.4.2.2 Airborne Monitoring Channels Process and effluent radiation monitoring systems provide useful information about the airborne activity within the plant buildings. These systems, described in Section 11.4, are the following:

(1) Containment Building lower and upper compartment air monitors.

(2) Auxiliary Building ventilation monitor.

In addition to these process and effluent monitors, there are 11 channels for monitoring airborne particulate activity. These channels are listed in Table 12.3-5.

Monitoring of the Auxiliary Building airborne activity during normal conditions is accomplished with the seven airborne monitoring systems that monitor the spent fuel pool area, the Unit 1 sample room, the holdup valve gallery general area, the decontamination area, the safety injection pump general area and the two waste packaging areas. Monitoring of the Unit 1 control room and the Unit 2 hot sample room is accomplished using portable continuous air monitor (CAM) having a range of 0.1 to 100 DAC. The portable CAMs use a pumping system that automatically regulates sample flow through a filter that is periodically advanced. They are provided with local alarms that annunciate upon instrument malfunction or high radiation. The CAMs are calibrated annually using approved calibration sources. The portable CAMs are located and administrated by Site RADCON.

The Auxiliary Building ventilation monitor real time particulate and iodine channels are used to supplement auxiliary building monitors listed in Table 12.3-5. These channels detect airborne radioactivity from particulates and iodines in excess of 10 DAC-Hr from any area in the Auxiliary Building which may be normally occupied, taking into account dilution in the ventilation system.

The lower compartment Unit 1 instrument room monitor provides monitoring of airborne activity in the Reactor Building instrument room during normal operation. The Containment Building upper and lower compartment monitors (described in Section 11.4) can also be used to monitor for airborne activity in the Reactor Building under normal conditions and accident conditions provided containment isolation has not occurred, or conditions allow the lines to be reopened.

The locations of the permanently installed airborne monitors that are in the Auxiliary, Reactor and Control Buildings can be determined by their respective coordinates from Table 12.3-5 applied to Figures 12.3-4, 12.3-8, 12.3-10, 12.3-12, and 12.3-15.

12.3.4.2.3 Operational Characteristics Paragraphs 12.3.4.2.3 through 12.3.4.2.6 discuss the permanently installed airborne monitors. For operational characteristics of the Containment Building lower and upper 12.3-20 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-90 compartment air monitors and the Auxiliary Building ventilation monitors, refer to Section 11.4.

Each permanently installed particulate channel has a detector assembly containing a moving tape filter through which a continuous air sample is drawn. Particulate radioactivity collected on the moving tape filter is detected by a beta scintillation detector. Local indication, recording, alarms, and pump and flow controls are provided.

12.3.4.2.4 Component Descriptions The following component descriptions apply to the permanently installed airborne monitoring channels:

Detectors The particulate detector units employ beta scintillation detectors and built-in preamplifiers. The detectors are able to detect beta radiation of an energy level of 0.2 MeV and above.

Filter Transport System The particulate filter transport mechanism is operated in the continuous advance mode. A local visual filter tape tear alarm is provided. A filter tape tear alarm also initiates an instrument malfunction alarm in the main control room.

Local Ratemeter The log ratemeter is a solid-state device with a range of 101 to 107 cpm. It is equipped with a solid-state power supply with adjustable high voltage.

Local alarms are provided on the ratemeter chassis for high radiation and instrument malfunction. Visual and audible alarms are provided in the main control room on detection of a high radiation or instrument malfunction condition.

Multipoint Recorder (0-M-12)

The air particulate monitor activity signals are recorded on a common multipoint recorder located on panel 0-M-12 in the main control room.

Local Recorders A single pin local recorder is provided at each airborne monitor unit. Each recorder has direct-reading indicating scales calibrated from 101 to 107 cpm.

Pumping System The pumping system has automatic flow regulation with manual override. A flow indicator is also provided. Visual alarms for loss of flow are provided at the enclosure.

An instrument malfunction alarm is provided in the main control room on loss of flow.

RADIATION PROTECTION DESIGN FEATURES 12.3-21

WATTS BAR WBNP-90 12.3.4.2.5 Sensitivity, Range and Set Point The permanently installed particulate monitors located in the Auxiliary Building have a required range of 1.99E-8 to 7.96E-6 Ci/cc. For the Reactor Building instrument room particulate monitor, the required range is 1.08E-8 to 4.32E-6 Ci/cc. This required range is sufficient to detect 10 DAC hours of airborne radionuclides expected in the area (i.e., Co-60, Cs-137, etc.).

12.3.4.2.6 Calibration and Maintenance This section applies only to permanently installed monitors.

Periodic testing of each air particulate monitor includes a channel calibration performed at least once per fuel cycle, and a channel operational test that is performed at least once per quarter. Periodic maintenance is performed to replace filters, and ensure proper sampler operation. Testing of Containment Building lower and upper compartment air monitors are described in Chapter 11 and the Technical Specifications. Testing of the Auxiliary Building vent monitor is described in Chapter 11 and the Offsite Dose Calculation Manual.

The channel calibration is the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input from a radioactive calibration source. The channel calibration encompasses the entire channel, including the required sensor, alarm, interlock, display and trip functions. The channel calibration may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

The channel operational test (COT) is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the operability of required alarm, interlock, display and trip functions. The COT includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

The built-in checksource function exposes the channel detector to a radioactive source. The built-in checksource feature is used to verify functional response of the detector and/or electronics during the performance of the channel operational test.

This function is also used by operations personnel at other times, such as after maintenance, to rapidly determine channel operability.

Periodic maintenance to change filters and to ensure proper sampler operation is performed at a frequency based upon length of service time for filters and sample equipment as determined by operational experience and trending.

Maintenance is performed, as necessary, if abnormalities are detected during any of the above checks. Unscheduled maintenance is performed as required.

12.3-22 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-90 12.3.4.3 Deleted by Amendment 84.

12.3.4.4 Special Radiation Monitors The types of special radiation monitors are described in the following sections.

12.3.4.4.1 Portal Monitors The portal monitor is a radiation monitoring device for providing a visual and aural warning when the radiation contamination of an individual exceeds a preset level, especially on the heels and soles of the shoes and a general overall body scan. The portal monitors are located at the gatehouse.

12.3.4.4.2 Personnel Contamination Monitors The personnel contamination monitor is a radiation monitoring device designed to detect the presence and general location of beta-gamma contamination on the hands, shoes, and clothing of personnel. The instrument is designed to allow an individual to monitor himself.

12.3.4.4.3 Deleted by Amendment 84.

REFERENCES:

(1) SDC, A Shielding - Design Calculation Code for Fuel Handling Facilities (RSIC Code Package CCC-60).

(2) Reactor Shielding Design Manual, Theodore Rockwell III, D. Van Nostrund Company, Incorporated, New York, N. Y., 1956.

(3) ANSI N13.10-1974, Specification and Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents.

(4) WBNTSR-077 "Radiation Zones" (5) 10 CFR 20 January 1994.

(6) QAD-P5Z, "Source Shield Detector Problems", ID No. 262361.

RADIATION PROTECTION DESIGN FEATURES 12.3-23

WATTS BAR WBNP-84 Table 12.3-1 Deleted by Amendment 84 12.3-24 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-84 Table 12.3-2 Deleted by Amendment 84 RADIATION PROTECTION DESIGN FEATURES 12.3-25

WATTS BAR WBNP-84 Table 12.3-3 Ventilation Air Exhaust Points (Page 1 of 2)

Volume Location (ft3)

REACTOR BUILDING (General Purge) 1,053,800 Vicinity of Reactor Vessel Refueling Canal Instrument Room1 MAIN CONTROL ROOM (Level of Control Building) 260,000 AUXILIARY BUILDING - GENERAL 3,480,000 Pipe Shafts Elevation 737 Heat Exchanger Rooms Hot Instrument Shop General Spaces Spent Fuel Pit Skimmer Filter Room Elevation 713 Sample Room Waste Gas Compressor Rooms Decontamination Equipment Room Demineralizer and Filter Rooms Ion Exchange and Filter Rooms Sample Exhaust Hood Waste Gas Analyzer Room Valve Gallery General Spaces Boric Acid Filter Rooms Demineralizer Ion Exchange and Filter Valve Room Elevation 692 Gas Stripper Rooms Hold-Up Tank Rooms Waste Evaporator Rooms Gas Decay Tank Rooms Turbine Driven Auxiliary Feedwater Pump Rooms Charging Pump Rooms Safety Injection System Pump Rooms Refueling Purification Filter Rooms Sample Exhaust Hood General Areas Concentrates Filter Rooms Construction Tool Room 12.3-26 RADIATION PROTECTION DESIGN FEATURES

WATTS BAR WBNP-84 Table 12.3-3 Ventilation Air Exhaust Points (Page 2 of 2)

Volume Location (ft3)

AUXILIARY BUILDING - GENERAL (Continued)

Elevation 676 Gas Stripper Feed Pump Room Waste Hold-up Tank Room Floor Drain Collector Tank Room Waste Evaporator Feed Pump Room Sump Tank and Pump Room Containment Spray Pump Rooms Residual Heat Removal Pump Rooms Waste Evaporator Feed Filter Rooms Auxiliary Waste Evaporator Feed Filter Rooms FUEL HANDLING AND RADWASTE PACKAGING AREA 1,012,900 Fuel Handling Area Spent Fuel Pool Area Waste Package Area Fuel Transfer Canal Cask Loading Area Cask Decontamination Tank Room Nitrogen Storage Area EGTS Room Blowdown Treatment Room Spent Resin Tank Room Post Accident Sampling Facility Cask Decontamination Room Condensate Demineralizer Waste Evaporator Building RADIATION PROTECTION DESIGN FEATURES 12.3-27

12.3-28 WATTS BAR Table 12.3-4 Location of Plant Area Radiation Monitors (Page 1 of 2)

Location Building Monitor No. Building and Elevation Coordinates Area Range 1-RE-90-1 Auxiliary El. 757.0 A5-w Spent Fuel Pool Area 10-1 to 104 mR/hr 1-RE-90-2 Auxiliary El. 757.0 A5-w Personnel Air Lock 10-1 to 104 R/hr 0-RE-90-3 Auxiliary El. 729.0 A6-y Waste Packaging Area 10-1 to 104 mR/hr 0-RE-90-4 Auxiliary El. 713.0 A2-q Equipment Decon Area 10-1 to 104 mR/hr 0-RE-90-5 Auxiliary El. 737.0 A9-v Spent Fuel Pool Pump Area 10-1 to 104 mR/hr 1-RE-90-6 Auxiliary El. 737.0 A5-s Comp Clg Ht Exch Area 10-1 to 104 mR/hr 1-RE-90-7 Auxiliary El. 713.0 A5-w Sample Room 10-1 to 104 mR/hr 1-RE-90-8 Auxiliary El. 713.0 A4-t Aux FW Pumps Area 10-1 to 104 mR/hr 0-RE-90-9 Auxiliary El. 692.0 A5-w Waste Evap Cnds Tk Area 10-1 to 104 mR/hr 1-RE-90-10 Auxiliary El. 692.0 A4-t CVCS Board Area 10-1 to 104 mR/hr 0-RE-90-11 Auxiliary El. 676.0 A7-u Cntmt Spray & RHR Pump Area 10-1 to 104 mR/hr 1-RE-90-59 Reactor El. 756.63 Az 315E Cntmt Refueling Floor 10-1 to 104 mR/hr 1-RE-90-60 Reactor El. 756.63 Az 225E Cntmt Refueling Floor 10-1 to 104 mR/hr 1-RE-90-61 Reactor El. 713.0 T1-K Lower Compt Inst Rm. 10-1 to 104 mR/hr 0-RE-90-135 Control El. 757.0 C7-q Main Cntl Rm Rad Mon 10-1 to 104 mR/hr 0-RE-90-230 Turbine El. 685.0 T8-E Condensate Demin Area 10-1 to 104 mR/hr 0-RE-90-231 Turbine El. 685.0 T8-E Condensate Demin Area 10-1 to 104 mR/hr 10E to 108 R/hr RADIATION PROTECTION DESIGN FEATURES 1-RE-90-271 Reactor El. 806.0 Az 180E RB Upper Comp Post Accident 1-RE-90-272 Reactor El. 806.0 Az 180E RB Upper Comp Post Accident 10E to 108 R/hr 1-RE-90-273 Reactor El. 728.0 Az 170E RB Lower Comp Post Accident 10E to 108 R/hr 1-RE-90-274 Reactor El. 728.0 Az 170E RB Lower Comp Post Accident 10E to 108 R/hr 2-RE-90-1 Auxiliary El. 757.0 A11-W Spent Fuel Pool Area 10-1 to 104 mR/hr 2-RE-90-2(1) Auxiliary El. 757.0 A11-W Personnel Air Lock 10-1 to 104 R/hr 2-RE-90-6 Auxiliary El. 737.0 A11-s Comp Clg Ht Exch Area 10-1 to 104 mR/hr 2-RE-90-7 Auxiliary El. 713.0 A11-W Sample Room 10-1 to 104 mR/hr 2-RE-90-8 Auxiliary El. 713.0 A12-t Aux FW Pumps Area 10-1 to 104 mR/hr 2-RE-90-10 Auxiliary El. 692.0 A12-t CVCS Board Area 10-1 to 104 mR/hr 2-RE-90-59(1) Reactor El. 756.63 Az 315E Contmt Refueling Floor 10-1 to 104 mR/hr WBNP-91 2-RE-90-60(1) Reactor El. 756.63 Az 225E Contmt Refueling Floor 10-1 to 104 mR/hr

Table 12.3-4 Location of Plant Area Radiation Monitors RADIATION PROTECTION DESIGN FEATURES WATTS BAR (Page 2 of 2)

Location Building Monitor No. Building and Elevation Coordinates Area Range 2-RE-90-61(1) Reactor El. 736.0 Az 88E Lower Compt Inst. Rm. 10-1 to 104 mR/hr 2-RE-90-271(1) Reactor El. 806.0 Az 360E Upper Cont High Range 10E to 108 R/hr 2-RE-90-272(1) Reactor El. 806.0 Az 360E Upper Cont High Range 10E to 108 R/hr 2-RE-90-273(1) Reactor El. 728.0 Az 7E Lower Cont High Range 10E to 108 R/hr 2-RE-90-274(1) Reactor El. 728.0 Az 7E Lower Cont High Range 10E to 108 R/hr 1-RE-90-280 Auxiliary El. 729.0 A5-x Post Accident Sampling Room Unit 1 10E to 105 mR/hr 2-RE-90-280(1) Auxiliary El. 729.0 A11-x Post Accident Sampling Room Unit 2 10E to 105 mR/hr (1) Not required for Unit 1 operation.

12.3-29 WBNP-91

12.3-30 WATTS BAR Table 12.3-5 Airborne Particulate Activity Monitoring Channels Location Building Monitor No. Building and Elevation Coordinates Area Range 0-RE-90-12 Aux Bldg. EL. 757.0 A8-x Spent Fuel Pool Area 101 to 107 cpm 1-RE-90-14 Aux Bldg, EL. 713.0 A6-w Sample Room Unit 1 101 to 107 cpm 0-RE-90-15 Aux Bldg. EL. 713.0 A8-t Holdup Valve Gallery 101 to 107 cpm General Spaces 0-RE-90-16 Aux Bldg. EL. 713.0 A2-r Decontamination Area 101 to 107 cpm 0-RE-90-17 Aux Bldg. EL. 692.0 A8-u Safety Injection Pump 101 to 107 cpm General Spaces 1-RE-90-62 Reac Bldg EL. 716.0 Az 272° Lower Compartment Unit 1 101 to 107 cpm Instrument Room 2-RE-90-62(1) Reac Bldg EL. 716.0 Az 272° Lower Compartment Unit 2 101 to 107 cpm Instrument Room 0-RE-90-13 Aux Bldg. EL. 729.0 A8-y Shipping Bay 101 to 107 cpm 0-RE-90-138 Aux Bldg. EL. 729.0 A10-y Waste Packaging Room 101 to 107 cpm (1)

Not required for Unit 1 operation.

RADIATION PROTECTION DESIGN FEATURES WBNP-90

RADIATION PROTECTION DESIGN FEATURES WATTS BAR Table 12.3-6 Shield Wall Thickness (Page 1 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing Elev. 674' 0" A1 Concrete Fill 2' 0" 2' 3" 3'9" / 4' 3" 41N366-1 41N328-1 A2 Concrete Fill 2' 0" 4' 3" 2' 3" 41N328-1 41N366-1 Elev. 676' 0" A2 5' 0" 3' 0" 3' 0" 5' 0" 41N306-1 41N473-1 41N470-1 A3 5' 0" 4' 0" 3' 0" 3' 0" 41N306-1 41N473-1 41N470-1 A4 5' 0" 2' 0" 2' 6" 3' 0" 41N306-1 41N373-1 41N309-1 41N470-1 A4a 2' 0" 4' 0" 2' 6" 3' 0" 41N306-1 41N373-1 41N309-1 A5 1' 2" 3' 0" 1' 2" 1' 2" 41N309-1 41N307-4 A10 2' 0" 2' 0" 2' 3" 3' 9" 41N366-1 41N309-1 WBNP-84 12.3-31 WBNP-72

Table 12.3-6 Shield Wall Thickness 12.3-32 WATTS BAR (Page 2 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing All 2' 0" 2' 0" 2' 3" 3' 9" 41N366-1 41N309-1 A12 2' 0" 2' 0" 3' 9" 2' 3" 41N366-1 41N309-1 A13 2' 0" 2' 0" 3' 9" 2' 3" 41N366-1 41N309-1 Elev. 692' 0" A3 3' 6" 4' 0" 3' 6" 3' 0" 41N330-1 41N310-1 41N368-1 41N473-1 A4 1' 0" 1' 0" 3' 0" 1' 0" 41N368-1 41N310-1 A5 3' 6" 4' 0" 3' 0" 3' 6" 41N310-1 41N470-1 A9 3' 0" 2' 0" 3' 6" 2' 0" 41N368-1 RADIATION PROTECTION DESIGN FEATURES 41N330-1 A10 3' 0" 2' 0" 2' 0" 3' 6" 41N368-1 41N330-1 A11 2' 2" 2' 0" 2' 0" 2' 0" 41N368-1 A15 2' 3" 4' 0" 2' 6" 4' 0" 41N328-2 41N368-2 41N337-1 WBNP-84

Table 12.3-6 Shield Wall Thickness RADIATION PROTECTION DESIGN FEATURES WATTS BAR (Page 3 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing WBNP-72 Al6 4' 0" 2' 3" 2' 6" 4' 0" 41N328-2 41N368-2 A18 2' 0" 1' 6" 1' 6" 2' 0" 41N368-1 A21 2' 0" 2' 0" 2' 0" 2' 0" 41N368-1 A22 3' 0" 2' 0" 3' 6" 2' 0" 41N368-1 41N330-1 A23 3' 0" 2' 0" 2' 0" 3' 6" 41N368-1 41N330-1 A29 2' 3" 3' 6" 2' 3" 2' 3" 41N368-1 41N310-1 41N470-1 A30 2' 3" 3' 6" 2' 3" 3' 0" 41N368-1 41N330-1 41N470-1 A31 3' 0" 3' 6" 3' 0" 3' 0" 41N368-1 41N330-1 Elev. 713' 0" A6 3' 0" 4' 0" / 3' 0" 2' 3" / 4' 0" 1' 6" 41N370-1 41N315-2 41N344-3 A7 4' 0" 4' 0" 2' 3" 4' 0" 41N315-2 41N370-1 A9 4' 0" 2' 0" 4' 0" / 3' 0" 2' 0" 41N370-1 A10 2' 3" 2' 6" 2' 3" 2' 3" 41N370-1 12.3-33 WBNP-72 A11 2' 3" 2' 3" 2' 3" 2' 6" 41N370-1

Table 12.3-6 Shield Wall Thickness 12.3-34 WATTS BAR (Page 4 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing A12 2' 3" 2' 3" 2' 3" 2' 6" 41N370-1 A13 2' 3" 2' 3" 0' 6" / 1' 0" 2' 6" 41N370-1 41N373-1 A14 2' 3" 2' 3" 2' 6" 0' 6" / 1' 0" 41N370-1 41N373-1 A15 2' 3" 2' 3" 2' 6" 2' 3" 41N370-1 A16 2' 3" 2' 3" 2' 6" 2' 3" 41N370-1 A17 2' 3" 2' 6" 2' 3" 2' 3" 41N370-1 A18 4' 0" 2' 0" 2' 0" 4' 0" / 3' 0" 41N370-1 A19 3' 0" 4' 0" / 3' 0" 1' 6" 2' 3" / 4' 0" 41N370-1 41N315-1 41N344-3 A20 4' 0" 4' 0" 4' 0" 2' 3" 41N370-1 41N315-1 A22 1' 9" 1' 9" 1' 9" 1' 9" 41N370-1 RADIATION PROTECTION DESIGN FEATURES A23 1' 6" / 1' 0" 2' 6" / 1' 0" 1' 6" / 1' 0" 1' 0" 41N370-1 A24 1' 9" 2' 6" 2' 1" 2' 6" 41N370-1 A25 2' 6" 3' 0" 2' 6" 2' 6" 41N370-1 41N315-2 A26 2' 6" 3' 0" 2' 6" 2' 6" 41N370-1 41N315-2 A28 6' 0" 2' 3" 2' 3" / 2' 6" 3' 0" 41N358-1 41N315-2 A29 6' 0" 2' 3" 3' 0" 2' 3" / 2' 6" 41N358-1 WBNP-72 41N315-2

Table 12.3-6 Shield Wall Thickness RADIATION PROTECTION DESIGN FEATURES WATTS BAR (Page 5 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing Elev. 713' 0" DEMIN PITS P1 1' 0" 3' 0" 1' 0" 2' 6" 41N370-1 P2 2' 6" 3' 0" 2' 6" 2' 6" 41N370-1 P3 2' 3" 3' 0" 2' 6" 2' 6" 41N370-1 P4 2' 6" 3' 0" 2' 6" 2' 6" 41N370-1 P5 2' 6" 3' 0" 2' 6" 2' 6" 41N370-1 P6 2' 6" 3' 0" 2' 6" 2' 0" 41N370-1 P7 1' 0" 3' 0" 2' 0" 1' 0" 41N370-1 P8 1' 0" 3' 0" 1' 0" 1' 0" 41N370-1 P9 4' 0" 3' 3" 3' 0" 3' 0" 41N370-1 P10 4' 0" 2' 6" 3' 0" 2' 6" 41N370-1 P11 4' 0" 2' 6" 2' 6" 2' 6" 41N370-1 P12 4' 0" 2' 6" 2' 6" 2' 6" 41N370-1 P13 4' 0" 4' 0" 4' 0" 4' 0" 41N370-1 P14 4' 0" 4' 0" 4' 0" 4' 0" 41N370-1 P15 3' 3" 4' 0" 4' 0" 3' 3" 41N370-1 P16 4' 0" 2' 6" 2' 6" 2' 6" 41N370-1 P17 4' 0" 2' 6" 2' 6" 2' 6" 41N370-1 P18 4' 0" 2' 6" 2' 6" 3' 0" 41N370-1 WBNP-72 P19 4' 0" 3' 3" 3' 0" 3' 0" 41N370-1 12.3-35 P20 3' 3" 4' 0" 3' 3" 4' 0" 41N370-1

Table 12.3-6 Shield Wall Thickness 12.3-36 WATTS BAR (Page 6 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing P21 4' 0" 4' 0" 4' 0" 4' 0" 41N370-1 P22 4' 0" 4' 0" 4' 0" 4' 0" 41N370-1 RADIATION PROTECTION DESIGN FEATURES WBNP-72

Table 12.3-6 Shield Wall Thickness RADIATION PROTECTION DESIGN FEATURES WATTS BAR (Page 7 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing ELEV. 713' 0" BATS FILTER 1' 6" 4' 0" 1' 6" 1' 6" 41N370-6 ELEV. 729' 0" A3 2' 9" 2' 0" / 1' 6" 2' 9" 2' 9" 41N388-1 A4 2' 6" 4' 6" / 3' 6" 2' 9" 2' 6" 41N388-1 A5 4' 6" / 3' 6" 1' 6" 2' 9" 2' 9" 41N337-1 41N388-1 ELEV. 737' 0" A5 3' 0" 3' 0" 2' 3" / 2' 6" 3' 0" 41N318-2 A7 2' 3" 3' 0" 2' 3" 2' 6" 41N372-1 A8 2' 3" 3' 0" 2' 6" 2' 3" / 2' 6" 41N372-1 A9 3' 0" 3' 0" 3' 0" 2' 6" 41N318-1 A15 1' 6" 1' 6" 1' 6" 1' 6" 41N372-2 A16 1' 6" 1' 6" 1" 6" 1' 6" 41N372-2 ELEV. 730' 6" DE3 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 DE4 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 DE5 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 DE6 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 12.3-37 WBNP-84

Table 12.3-6 Shield Wall Thickness 12.3-38 WATTS BAR (Page 8 of 8)

North Wall South Wall East Wall West Wall Room Number and Elevation Thickness Thickness Thickness Thickness Ref. Drawing WBNP-72 ELEV. 750' 6" DE2 2' 0" 2' 0" 2' 0' 2' 0" 41W391-7 DE3 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 DE4 2' 0" 2' 0" 2' 0" 2' 0" 41W391-7 ELEV. 685' 0" UNIT 1 TURB. BLD. CON DEMIN 1' 3" 1' 3" 1' 3" 8' 6" 41N233-3 41W238-2 ELEV. 685' 0" UNIT 2 TURB. BLD. CON DEMIN 1' 3" 1' 3" 8' 6" 1' 3" 41W238-2 41N233-3 ELEV. 685' 0" UNIT 1/2 TURB. BLD. CON 1' 3" 1' 3" 1' 3" 1' 3" 41W233-3 DEMIN NEUT TANK RADIATION PROTECTION DESIGN FEATURES WBNP-72

WATTS BAR WBNP-89 Table 12.3-7 Special Radiation Monitors Monitor Area Portal Monitors Gate Houses Personnel Aux.

Contamination Building Monitors Entrance RADIATION PROTECTION DESIGN FEATURES 12.3-39

WATTS BAR WBNP-89 THIS PAGE INTENTIONALLY BLANK 12.3-40 RADIATION PROTECTION DESIGN FEATURES

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-1 Plans - Roof - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-41

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-2 Plan El. 772.0 and Above - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-42

WATTS BAR RADIATION PROTECTION DESIGN FEATURES Security-Related Information - Withheld Under 10CFR2.390 WBNP-72 Figure 12.3-3 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 772.0, 782.0 and 786.0 - Radiation Zone Map 12.3-43

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-90 Figure 12.3-4 Powerhouse, Auxiliary, Reactor and Control Building - Plan El. 755.0 and 757.0 - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-44

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-5 Powerhouse, Units 1 and 2 - Plan El. 757.0 and 755.0 - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-45

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-6 Powerhouse, Units 1 and 2 - Plan El. 737.0 and 729.0 - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-46

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-7 Auxiliary Building, Units 1 and 2 - Radiation Zone Map, Condensate Demineralizer RADIATION PROTECTION DESIGN FEATURES 12.3-47

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-90 Figure 12.3-8 Powerhouse, Auxiliary, Reactor and Control Buildings - Radiation Zone Map - Plan El. 729.0 and 737.0 RADIATION PROTECTION DESIGN FEATURES 12.3-48

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-9 Powerhouse, Units 1 and 2 - Plan El. 713.0 and 708.0 RADIATION PROTECTION DESIGN FEATURES 12.3-49

WATTS BAR RADIATION PROTECTION DESIGN FEATURES Security-Related Information - Withheld Under 10CFR2.390 WBNP-90 Figure 12.3-10 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 708.0 and 713.0 - Radiation Zone Map 12.3-50

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-11 Powerhouse, Units 1 and 2 - Plan El. 685.5 RADIATION PROTECTION DESIGN FEATURES 12.3-51

WATTS BAR RADIATION PROTECTION DESIGN FEATURES Security-Related Information - Withheld Under 10CFR2.390 WBNP-90 Figure 12.3-12 Powerhouse, Auxiliary, Reactor and Control Buildings - Plan El. 676.0 and 692.0 - Radiation Zone Map 12.3-52

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-13 Powerhouse, Units 1 and 2 - Turbine Building - Plan El. 670.0 Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-53

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-14 Powerhouse, Units 1 and 2 - Reactor Building - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-54

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-15 Powerhouse, Units 1 and 2 - Reactor Building - Radiation Zone Map RADIATION PROTECTION DESIGN FEATURES 12.3-55

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-16 Powerhouse, Auxiliary Building, Mechanical, Heating, Ventilating and Air Conditioning RADIATION PROTECTION DESIGN FEATURES 12.3-56

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-72 Figure 12.3-17 Fuel Transfer Tube Shielding Details RADIATION PROTECTION DESIGN FEATURES 12.3-57

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-88 Figure 12.3-18 Powerhouse Service Building RADIATION PROTECTION DESIGN FEATURES 12.3-58

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-88 Figure 12.3-19 Service Building - Miscellaneous Details, Hot Shop Facilities RADIATION PROTECTION DESIGN FEATURES 12.3-59

WATTS BAR WBNP-88 THIS PAGE INTENTIONALLY BLANK RADIATION PROTECTION DESIGN FEATURES 12.3-60

WATTS BAR WBNP-92 12.4 DOSE ASSESSMENT Annual Exposures Estimates of yearly exposures to plant personnel are made from the information used to prepare the annual Regulatory Guide 1.16 Report.Estimates include the exposure to these persons during one refueling and during minor maintenance that may be performed during the year. The estimates follow.

Total Dose Activity Total Persons* (man-rem/yr)

Reactor operations 950 12,000

& surveillance Routine 1700 45,000 Maintenance Inservice 400 5,000 inspections Special 1000 50,000 maintenance Waste Processing 100 2,000 Refueling 500 35,000 Total 4650 149,000

• Individual persons are counted in each activity in which they are estimated to participate. One individual may be counted as six if working in all of the different activities. The average measurable dose per plant worker is expected to be between 0.050 and 0.500 Rem/yr.

The Regulatory Guide 1.16 Report is submitted annually to the Nuclear Regulatory Commission and contains actual dose accrual in the subject categories for the previous calendar year.

A requirement imposed upon all features of the WBN design is that radiation doses to plant personnel during normal operation, including anticipated operational occurrences, can be maintained as low as reasonably achievable (ALARA). In satisfying this requirement, the shielding design of the WBN incorporates (1) effective locations for many shield wall penetrations, (2) operation of valves from behind shield walls, (3) provisions to drain equipment from behind shield walls prior to maintenance, and (4) arrangements for shielding of spent filters during the removal process.

DOSE ASSESSMENT 12.4-1

WATTS BAR WBNP-92 Since the original design of WBN, several dose minimization initiatives have been implemented that will provide additional assurance that worker doses will be ALARA.

Some of the significant changes include upgrades to steam generators, i.e., channel-head polishing, tube-sheet stenciling, tube stress relieving, and expanding work area platforms. Additionally, the elimination of the RTD bypass loops and upper-head injection systems will aid in the minimization of source terms within the plant. A major effort has also been undertaken to seal, where practicable, penetrations and pipe sleeves with high density sealant material to minimize radiation streaming and scattering to normally accessible areas.

Operating procedures are designed to ensure that the plant staff receive the minimum radiation dose allowed by the plant design.

The combination of shielding design and operation of the plant in accordance with written operating procedures gives WBN a lower man-rem total. Design improvements should decrease the need for staff augmentation in order to keep personnel doses within 10 CFR 20 limits.

Estimate of Personnel Internal Exposure The anticipated maximum internal exposure to any employee is expected to be far below the NRC limits set out in 10 CFR 20. This is accomplished by controlling ventilation air in the buildings such that the supplied clean air is exhausted from clean areas to the more contaminated atmospheres. Therefore, those areas normally occupied by personnel are supplied with fresh outside air and would contain very small quantities of radioactive material.

The need for monitoring internal exposures will be documented in a prospective evaluation following the guidance of Regulatory Guide 8.7. For work in contaminated atmospheres, RADCON will evaluate the need for respiratory protection equipment using the guidance of an established Total Effective Dose Equivalent (TEDE) ALARA policy (further discussion of this evaluation is in Section 12.5.2).

Boundary of the Restricted Area Estimated annual doses at the boundary of the restricted area (located inside the exclusion area defined in 2.1.2) are given in Tables 12.4-1 and 12.4-2. The total dose rates are the sum of the adult inhalation dose rate and the gamma dose rate from the plume and ground contamination and the gamma dose rate from outdoor storage tanks shown in Figure 12.4-1. The gaseous effluents are releases from the Auxiliary Building, the Containment Building, and the Turbine Building. Dose rate contributions from shielded sources inside the plant building are negligible. The model used for estimating the plume dose rates is described in the Offsite Dose Calculation Manual for WBN. The dose rates from the two makeup water storage tanks and the two refueling water storage tanks were estimated with the computer code QAD-P5Z. As indicated in Table 12.4-2, the highest total dose rate at the boundary of the restricted area is 105 mrem/yr, based on a continuous 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />s/yr occupancy. Use of a more realistic occupancy, reflective of the transient traffic expected for this location, would result in a much lower dose estimate. It is, therefore, considered highly unlikely that a 12.4-2 DOSE ASSESSMENT

WATTS BAR WBNP-84 member of the public would receive > 100 mrem/yr at or beyond the restricted area boundary.

Inside Restricted Area Estimated dose rates from the gaseous plume and tanks have also been calculated at locations within the restricted area. Dose points inside the restricted area have been selected as being possible points at which the maximum dose rate might occur and locations construction workers might occupy. Again, dose rate contribution, from shielded sources inside the plant buildings are negligible. Table 12.4-2 lists each point of interest. The dose rates listed are for two units in operation. Figure 12.4-1 shows the locations of the dose points inside the restricted area.

REFERENCES None.

DOSE ASSESSMENT 12.4-3

WATTS BAR WBNP-84 Table 12.4-1 Restricted Area Boundary Dose Rates For 2000 Hr/Year Exposure Total Compass Distance Inhalation Plume Ground Effluent Dose Sector (meters) Dose Dose Exposure (mrem)

N 169 4.2E-01 3.1E+01 1.0E+00 32.9 NNE 174 7.2E-01 4.2E+01 2.0E+00 44.6 NE 203 4.6E-01 2.5E+01 6.9E-01 26.3 ENE 240 4.4E-01 2.3E+01 4.7E-01 24.0 E 234 5.0E-01 2.6E+01 5.4E-01 27.5 ESE 240 4.5E-01 2.4E+01 4.9E-01 24.7 SE 282 4.7E-01 2.4E+01 5.2E-01 25.0 SSE 436 1.4E-01 6.5E+00 2.2E-01 6.9 S 431 8.9E-02 4.5E+00 2.3E-01 4.8 SSW 364 1.2E-01 6.1E+00 3.7E-01 6.6 SW 217 3.0E-01 1.7E+01 5.7E-01 17.4 WSW 217 4.7E-01 2.5E+01 5.9E-01 26.3 W 167 3.0E-01 1.7E+01 5.7E-01 17.4 WNW 168 1.6E-01 9.1E+00 1.7E-01 9.4 NW 195 2.0E-01 1.1E+01 2.2E-01 11.4 NNW 173 3.2E-01 1.8E+01 4.1E-01 18.3 12.4-4 DOSE ASSESSMENT

WATTS BAR WBNP-84 Table 12.4-2 Onsite Dose Rates (Mrem/Yr)

Gaseous Location Effluents U1 PWST U2 PWST U1 RWST U2 RWST CSTs Total RA 1 1.74E+01 2.66E-01 - 0.88E+02 - - 1.05E+02 RA 2 3.29E+01 4.06E-02 4.06E-02 9.41E+00 9.41E+00 - 5.18E+01 RA 3 2.75E+01 - 8.90E-02 - 2.93E+01 2.66E-02 5.69E+01 P1 0.99E+02 1.77E-01 1.77E-01 4.67E+01 4.67E+01 - 1.93E+02 P2 1.27E+02 - 8.28E+00 - 5.19E+02 3.30E-02 6.54E+02 P3 1.57E+02 - 3.36E+01 - 8.81E+03 7.16E-02 9.00E+03 P4 1.10E+02 - 3.66E-01 - 3.19E+02 6.40E-01 4.30E+02 P5 2.28E+01 - 1.79E-02 - 7.99E+00 7.16E-02 3.09E+01 P6 2.41E+01 7.20E-02 - 3.67E+01 - - 6.09E+01 P7 1.83E+01 4.86E-02 - 2.37E+01 - - 4.20E+01 P8 0.81E+02 - - 8.81E+03 - - 8.89E+03 SB 2.81E+00 9.26E-06 9.29E-06 2.67E-03 2.67E-03 9.33E-06 2.82E+00 Location Descriptions RA 1 - West restricted area boundary RA 2 - North restricted area boundary RA 3 - East restricted area boundary P1 - 80' North of waste package area "A" P2 - 45' NE of U2 primary water storage tank (PWST)

P3 - Between U2 refueling water storage tank (RWST) and PWST P4 - 90' NE of U2 condensate storage tank (CST)

P5 - North end of switchyard P6 - West corner of main office building P7 - SE corner of main parking lot P8 - Between U1 RWST and PWST SB - Land site boundary (1250 meters SE)

DOSE ASSESSMENT 12.4-5

WATTS BAR WBNP-84 THIS PAGE INTENTIONALLY BLANK 12.4-6 DOSE ASSESSMENT

Security-Related Information - Withheld Under 10CFR2.390 WATTS BAR WBNP-84 Figure 12.4-1 Dose Points Inside Restricted Area Dose Assessment 12.4-7

WATTS BAR WBNP-84 THIS PAGE INTENTIONALLY BLANK Dose Assessment 12.4-8

WATTS BAR WBNP-92 12.5 RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5.1 Organization The radiological control program consists of four elements that are directed toward essential support to TVA's nuclear power program.

(1) Radiological impact assessments.

(2) Radiation protection planning and radiological safety evaluation, including preliminary safety analysis reports, final safety analysis reports, and radiological emergency plans.

(3) Radiological environmental monitoring.

(4) Radiological control activities The RADCON Section is under the supervision of the Plant Manager.

The RADCON Section is responsible for the radiological control activities at the plant.

It applies radiation standards and procedures; reviews proposed methods of plant operation; participates in development of plant documents; and assists in the plant training program, providing specialized training in radiation protection. It provides coverage for all operations involving radiation or radioactive materials including maintenance, fuel handling, waste disposal, and decontamination. It is responsible for personnel and inplant radiation monitoring, and maintains continuing records of personnel exposures, plant radiation, and contamination levels.

The RADCON superintendent is the onsite supervisor of the RADCON Section and is responsible for direction of an adequate program of radiological health surveillance for all plant operations involving potential radiation hazards. He keeps the plant manager informed at all times of radiation hazards and conditions related to potential exposure, contamination of plant and equipment, or contamination of site and environs. His duties include training and supervising RADCON technicians; planning and scheduling monitoring and surveillance services, scheduling technicians to ensure around-the-clock shift coverage as required; maintaining current data files on radiation and contamination levels, personnel exposures, and work restrictions; and ensuring that operations are carried out within the provisions of developed radiological control standards and procedures. He critiques plant operations and reviews suggestions from employees to identify areas in which exposures can be reduced. As an alternate member of the Plant Operations Review Committee, he reviews and consents on operating procedures. He provides assistance and advice to the Site Emergency Director during radiological emergencies.

As a minimum, the guidance of Regulatory Guides 8.2, 8.8, 8.10, and 1.8 has been followed in developing the RADCON program and the personnel qualifications.

The minimum qualification requirements for the RADCON Superintendent are stated in Section 13.1.3.

RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5-1

WATTS BAR WBNP-92 The minimum requirements for RADCON (health physics) technicians responsible for a shift are appropriate technical training and two years of applied health physics experience dealing with radiological problems similar to those encountered in a nuclear power station. Applicable experience may be granted as equivalency for the technical training.

Further information on the training and qualifications of RADCON personnel may be found in Chapter 13.

12.5.2 Equipment, Instrumentation, and Facilities The RADCON facilities consist of office space; short-term record storage; and a service center. The service center is equipped with instrumentation, supplies, cabinets, and storage space. Service center drains are piped to the Liquid Radwaste System for processing. The service center is located between the Auxiliary Building and service shop areas. The technicians use the service center as their base of operations and communications. Portable and laboratory radiation monitoring instruments, and other RADCON supplies including signs, personnel decontamination supplies, air sampling equipment, etc., are kept in this area.

Adjacent to the RADCON service center is a protective clothing/dress out area and a personnel decontamination room equipped with a shower. Radiation monitoring instruments for detection of very low levels of radioactive contamination are readily available.

The portable and laboratory equipment located in the service center will allow the RADCON personnel to measure dose rates and contamination levels throughout the plant in all routine and emergency situations. The portable RADCON survey instrumentation and the fixed RADCON and chemistry laboratory counting equipment are equivalent to the instrumentation described in Regulatory Guide 8.8, Position C.4.

Each portable survey instrument is calibrated and checked periodically with standard radioactive sources in accordance with instrument specific calibration and maintenance procedures. Accurate records on the performance of each instrument during each calibration are maintained. Each laboratory counting system is checked at regular intervals with standard radioactive sources for proper counting efficiencies, background count rates, and operating parameters.

TVA provides protective clothing for use in radiological areas. Clothing required for a particular instance is prescribed by RADCON based upon the actual or potential radiological conditions. Protective clothing is cleaned, surveyed for contamination, checked for physical condition, and returned to service if acceptable. Additional protective clothing stock is available from the plant warehouse as required. Protective clothing available for use includes:

(1) Coveralls (2) Lab coats 12.5-2 RADIOLOGICAL CONTROL (RADCON) PROGRAM

WATTS BAR WBNP-92 (3) Gloves (4) Head covers (5) Foot covers Tape may be provided so that openings in clothing can be sealed, if necessary.

Respiratory protection devices are available from the RADCON service center. The RADCON unit is responsible for the maintenance of the devices, although other groups may perform the actual work. The need for, and type of, respiratory protection equipment to be issued for specific tasks/activities is determined by RADCON evaluations. Maintaining TEDE ALARA and minimization of the total risk from all expected hazards is the goal of the evaluations. Considerations made in the performance of these evaluations should include:

(1) Process controls (e.g.,system flushing, venting, isolation)

(2) Engineering controls (e.g., containment devices, ventilation, remote handling tools)

(3) Radiological hazards (4) Industrial Safety hazards (5) Effects of respirators on worker efficiency and total dose (6) Environmental conditions (7) Need for precise communications and/or visual perception (8) Physiological and/or psychological affects of respirators (9) Job duration (e.g., access controls, stay times)

(10) Worker acceptance and input Available respiratory protection devices include:

(1) Full face mask with high efficiency filters (2) Full face mask with constant or pressure demand air flow. A manifold unit is used that contains mist filters, a regulator, and relief valve.

(3) Powered air purifying respirators (PAPRs) with high efficiency filters (4) Constant air flow hoods and/or suits (5) Self contained pressure demand breathing apparatus (bottled air type).

RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5-3

WATTS BAR WBNP-92 Prospective monitoring determinations for internal and external dose monitoring are performed for individuals or group of individuals entering the restricted area.

Personnel monitoring, for dose from sources external to the body, is conducted using appropriate dosimeters as required by 10CFR20. TVA maintains accreditation as a processing laboratory for dosimeters, as described in American Standards Institute (ANSI) N13.11-1983, "Personnel Dosimeter - Criteria for Performance". This accreditation is under the National Voluntary Laboratory Accreditation Program conducted by the National Institute of Standards and Technology. Dosimeters may be processed onsite by WBN, an accredited subfacility, or by another processing laboratory within the scope of TVA's accreditation. Dose information for whole body (total effective dose equivalent), external exposure of the skin, lens of the eye, and extremities is recorded in a dose tracking system and retained in a permanent historical database for generating required reports. Real time control is generally implemented using information from direct reading dosimeters. Official doses of record are taken from dosimeters. However, doses are calculated when dosimeter results are not available or do not accurately represent actual dose received.

Personnel monitoring and confirmatory monitoring for dose from intakes of radioactive material is conducted using DAC-HR tracking and bioassays, including whole body counting. Monitoring is performed for each person required to be monitored by 10CFR20. The whole body counter is calibrated with standard radioisotopes in configurations that approximate the human body. It is able to detect expected gamma emitting radionuclides per ANSI-N13.30, September 1989, Table-1, "Acceptable Minimum Detectable Activities."

12.5.3 Procedures Routine radiological surveys to detect radiation, radioactive contamination, and airborne radioactivity are performed throughout the plant on periodic schedules.

Survey frequencies are determined by the RADCON Superintendent based upon the actual or potential radiological conditions. Schedules for completion of routine surveys are issued to the technicians. As plant conditions change, the schedule will be updated. Radiological surveys may be performed whenever personnel enter potential or actual radiological areas and there is any doubt as to the existing conditions.

Retention of survey records follows the requirements of 10 CFR 20.2103 and Regulatory Guide 1.88.

Section 12.1.1 defines the TVA overall ALARA program. Inplant procedures involving radiological conditions are written such that keeping exposures ALARA is a major consideration.

Entry into Radiation Areas as defined by 10 CFR 20.1003 is administratively controlled.

Radiation Areas are posted per 10 CFR 20.1902. Entry to these areas requires the issuance of a Radiation Work Permit (RWP). The RWP sets out entry requirements and other precautions. In addition, any entry into Radiation Areas requires possession of an appropriate dosimeter.

Access controls to prevent unplanned exposures in high radiation areas are implemented in accordance with Technical Specifications 5.11, High Radiation Area.

12.5-4 RADIOLOGICAL CONTROL (RADCON) PROGRAM

WATTS BAR WBNP-92 In addition to the access control requirements for high radiation areas, the following control measures are implemented to control access to very high radiation areas in which radiation levels could be encountered at 500 rads or more in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 1 meter from a radiation source or any surface through which the radiation penetrates:

(1) Conspicuously posted with a sign(s) stating GRAVE DANGER - VERY HIGH RADIATION AREA (2) Area is locked. Each lock shall have a unique core. The keys shall be administratively controlled by the RADCON Superintendent.

(3) Plant manager's (or designee) approval required for entry.

(4) RADCON personnel shall be in accompaniment of the person(s) making the entry. RADCON shall assess the radiation exposure conditions at the time of the entry.

Areas where transferable radioactive contamination is present in levels greater than 1000 dpm/100 cm2 beta-gamma or 20 dpm/100 cm2 alpha are posted as "Contaminated Areas." Entry into a Contaminated Area requires a RWP which specifies protective clothing and measures dependent upon the conditions. Whenever practical, contaminated equipment will be packaged to prevent contamination spread and tagged with radioactive material tags when removed from a Contaminated Area.

All materials and equipment leaving Contaminated Areas will be monitored and released only if there is no contamination present in excess of established limits. All items which have been in a radiologically controlled area and which have the potential for becoming contaminated are monitored prior to being released from the area.

Potential airborne radioactivity concentrations are kept to a minimum by process and engineering controls. Airborne radioactivity conditions are evaluated by using strategically located continuous air monitors, as well as routine and special grab sampling.

Regulatory Guide 8.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program," is used as guidance in implementing a bioassay program.

Planned Special Exposures (PSE) may be authorized. In the event WBN uses a PSE, the PSE will be conducted in accordance with guidance from Regulatory Guide 8.35, "Planned Special Exposures".

Occupational exposure limits for minors, declared pregnant women, and for radiation dose to the embryo/fetus are established following the guidance of Regulatory Guide 8.34, "Monitoring Criteria and Methods to Calculate Occupational Radiation Doses",

and Regulatory Guide 8.36, "Radiation Dose to the Embryo/Fetus".

RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5-5

WATTS BAR WBNP-92 All personnel entering the RCA unescorted will have completed a radiological orientation course. This course consists of introductory subjects, monitoring techniques and equipment, protective procedures and equipment, and the radiological emergency plan. The presentation methods, length of the particular course, material emphasized, and participation in demonstrations are based upon the needs of the individual.

The storage and handling of byproduct materials and special nuclear materials is detailed in procedures.

REFERENCES None.

12.5-6 RADIOLOGICAL CONTROL (RADCON) PROGRAM

WATTS BAR WBNP-92 Appendix 12A Radiation Protection Features for the Tritium Production Program Beginning with Cycle 5, WBN will irradiate Tritium Producing Burnable Absorber Rods (TPBARs) in Unit 1 only to fulfill an Interagency Agreement with the U.S. Department of Energy (DOE).

TVA has performed an evaluation of the radiation sources for the tritium production program and determined that the core source term for the maximum irradiation level of 2,304 TPBARs is bounded by the existing source term of record for WBN. Additionally, the transition cycles to arrive at the maximum number of TPBARs have also been analyzed and are bounded by the existing source term of record.

Tritium is a radioactive isotope of hydrogen with a half-life of 12.3 years, which undergoes beta decay, with a maximum energy level of 18.6 KeV. The average energy level is 5.7 KeV. This low energy limits the maximum range of a tritium beta to about 6 millimeters in air and 0.0042 millimeters in soft tissue. Therefore, the primary radiological significance of exposure to tritium is in the form of internal exposure and the only potential hazard comes when personnel are exposed to open processes that have been wetted with tritiated liquids. Therefore, the design features of the plant that deal with contamination and airborne radioactive control, such as drain and ventilation systems are of potential concern. TVAs evaluation of these systems concluded that there is negligible impact to these systems by the tritium production program.

TVA has evaluated the current Operational Radiation Protection Program described in the preceding sections and determined that there will be no major impact due to the tritium production program. The program modifications are adjustments or changes in scope, rather than major program revisions. Additional monitoring instrumentation and sample equipment to allow better assessment of plant airborne tritium activity will be procured. Plant specific procedures addressing these actions will be developed before TPBAR irradiation. Tritium program enhancements include:

Tritium Bioassay Program - the tritium bioassay program will follow the guidance of U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 8.9 Acceptable Concepts, Models, Equations and Assumptions for a Bioassay Program.

Tritium Monitoring - tritium monitoring will consist of Air Monitoring, Air Sampling, Surface Monitoring, Liquid Monitoring and Liquid Scintillation Counting.

Radiological Environmental Monitoring Program - changes to the REMP to accommodate tritium production have been identified. These changes will consist of:

• Atmospheric moisture - selected atmospheric sampling stations will be modified to include the collection of atmospheric moisture.

• Surface water - perform tritium analysis on samples collected every four weeks from the downstream and upstream sampling locations.

RADIOLOGICAL CONTROL (RADCON) PROGRAM 12.5-7

WATTS BAR WBNP-92

• Public water - perform tritium analysis on samples collected every four weeks from downsteam public water systems.

• Ground water - perform tritium analysis on samples collected every four weeks from the site monitoring wells. Add monthly grab sampling at locations for the nearest (within 5 mile radius) offsite users of ground water as the source of drinking water.

With the program changes listed above, the WBN Radiation Protection Program will continue to provide assurance of the health and safety of plant employees and the public during the tritium production program.

Watts Barr FSAR Section 12.0 Radiation Protection 12.5-8 RADIOLOGICAL CONTROL (RADCON) PROGRAM